The comparison of various approach to evaluation erosion risks and design control erosion measures

Science.gov (United States)

Kapicka, Jiri

2015-04-01

In the present is in the Czech Republic one methodology how to compute and compare erosion risks. This methodology contain also method to design erosion control measures. The base of this methodology is Universal Soil Loss Equation (USLE) and their result long-term average annual rate of erosion (G). This methodology is used for landscape planners. Data and statistics from database of erosion events in the Czech Republic shows that many troubles and damages are from local episodes of erosion events. An extent of these events and theirs impact are conditional to local precipitation events, current plant phase and soil conditions. These erosion events can do troubles and damages on agriculture land, municipally property and hydro components and even in a location is from point of view long-term average annual rate of erosion in good conditions. Other way how to compute and compare erosion risks is episodes approach. In this paper is presented the compare of various approach to compute erosion risks. The comparison was computed to locality from database of erosion events on agricultural land in the Czech Republic where have been records two erosion events. The study area is a simple agriculture land without any barriers that can have high influence to water flow and soil sediment transport. The computation of erosion risks (for all methodology) was based on laboratory analysis of soil samples which was sampled on study area. Results of the methodology USLE, MUSLE and results from mathematical model Erosion 3D have been compared. Variances of the results in space distribution of the places with highest soil erosion where compared and discussed. Other part presents variances of design control erosion measures where their design was done on based different methodology. The results shows variance of computed erosion risks which was done by different methodology. These variances can start discussion about different approach how compute and evaluate erosion risks in areas

[Research progress on wind erosion control with polyacrylamide (PAM).

Science.gov (United States)

Li, Yuan Yuan; Wang, Zhan Li

2016-03-01

Soil wind erosion is one of the main reasons for soil degradation in the northwest region of China. Polyacrylamide (PAM), as an efficient soil amendment, has gained extensive attention in recent years since it is effective in improving the structure of surface soil due to its special physical and chemical properties. This paper introduced the physical and chemical properties of PAM, reviewed the effects of PAM on soil wind erosion amount and threshold wind velocity, as well as the effect differences of PAM in soil wind erosion control under conditions of various methods and doses. Its effect was proved by comparing with other materials in detail. Furthermore, we analyzed the mecha-nism of wind erosion control with PAM according to its influence on soil physical characteristics. Comprehensive analysis showed that, although some problems existed in wind erosion control with (PAM), PAM as a sand fixation agent, can not only enhance the capacity of the soil resis-tance to wind erosion, but also improve soil physical properties to form better soil conditions. Besides, we proposed that combination of PAM and plant growth would increase the survival rate of plants greatly, control soil wind erosion in wind-erosive areas, and improve the quality of the ecological environment construction. Thus, PAM has practically important significance and wide application prospect in controlling soil wind erosion.

U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

International Nuclear Information System (INIS)

1995-01-01

Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li 2 O) and lithium zirconate (Li 2 ZrO 3 ) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

Workshop on cold-blanket research

International Nuclear Information System (INIS)

1977-05-01

The objective of the workshop was to identify and discuss cold-plasma blanket systems. In order to minimize the bombardment of the walls by hot neutrals the plasma should be impermeable. This requires a density edge-thickness product of nΔ > 10 15 cm -2 . An impermeable cold plasma-gas blanket surrounding a hot plasma core reduces the plasma wall/limiter interaction. Accumulation of impurities in this blanket can be expected. Fuelling from a blanket may be possible as shown by experimental results, though not fully explained by classical transport of neutrals. Refuelling of a reacting plasma had to be ensured by inward diffusion. Experimental studies of a cold impermeable plasma have been done on the tokamak-like Ringboog device. Simulation calculations for the next generation of large tokamaks using a particular transport model, indicate that the plasma edge profile can be controlled to reduce the production of sputtered impurities to an acceptable level. Impurity control requires a small fraction of the radial space to accomodate the cold-plasma layer. The problem of exhaust is, however, more complicated. If the cold-blanket scheme works as predicted in the model calculations, then α-particles generated by fusion will be transported to the cold outside layer. The Communities' experimental programme of research has been discussed in terms of the tokamaks which are available and planned. Two options present themselves for the continuation of cold-blanket research

Can control of soil erosion mitigate water pollution by sediments?

Science.gov (United States)

Rickson, R J

2014-01-15

The detrimental impact of sediment and associated pollutants on water quality is widely acknowledged, with many watercourses in the UK failing to meet the standard of 'good ecological status'. Catchment sediment budgets show that hill slope erosion processes can be significant sources of waterborne sediment, with rates of erosion likely to increase given predicted future weather patterns. However, linking on-site erosion rates with off-site impacts is complicated because of the limited data on soil erosion rates in the UK and the dynamic nature of the source-pathway-receptor continuum over space and time. Even so, soil erosion control measures are designed to reduce sediment production (source) and mobilisation/transport (pathway) on hill slopes, with consequent mitigation of pollution incidents in watercourses (receptors). The purpose of this paper is to review the scientific evidence of the effectiveness of erosion control measures used in the UK to reduce sediment loads of hill slope origin in watercourses. Although over 73 soil erosion mitigation measures have been identified from the literature, empirical data on erosion control effectiveness are limited. Baseline comparisons for the 18 measures where data do exist reveal erosion control effectiveness is highly variable over time and between study locations. Given the limitations of the evidence base in terms of geographical coverage and duration of monitoring, performance of the different measures cannot be extrapolated to other areas. This uncertainty in effectiveness has implications for implementing erosion/sediment risk reduction policies, where quantified targets are stipulated, as is the case in the EU Freshwater Fish and draft Soil Framework Directives. Also, demonstrating technical effectiveness of erosion control measures alone will not encourage uptake by land managers: quantifying the costs and benefits of adopting erosion mitigation is equally important, but these are uncertain and difficult to

U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

Energy Technology Data Exchange (ETDEWEB)

1995-01-01

Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

Performance of silvered Teflon (trademark) thermal control blankets on spacecraft

Science.gov (United States)

Pippin, Gary; Stuckey, Wayne; Hemminger, Carol

1993-01-01

Silverized Teflon (Ag/FEP) is a widely used passive thermal control material for space applications. The material has a very low alpha/e ratio (less than 0.1) for low operating temperatures and is fabricated with various FEP thicknesses (as the Teflon thickness increases, the emittance increases). It is low outgassing and, because of its flexibility, can be applied around complex, curved shapes. Ag/FEP has achieved multiyear lifetimes under a variety of exposure conditions. This has been demonstrated by the Long Duration Exposure Facility (LDEF), Solar Max, Spacecraft Charging at High Altitudes (SCATHA), and other flight experiments. Ag/FEP material has been held in place on spacecraft by a variety of methods: mechanical clamping, direct adhesive bonding of tapes and sheets, and by Velcro(TM) tape adhesively bonded to back surfaces. On LDEF, for example, 5-mil blankets held by Velcro(TM) and clamping were used for thermal control over 3- by 4-ft areas on each of 17 trays. Adhesively bonded 2- and 5-mil sheets were used on other LDEF experiments, both for thermal control and as tape to hold other thermal control blankets in place. Performance data over extended time periods are available from a number of flights. The observed effects on optical properties, mechanical properties, and surface chemistry will be summarized in this paper. This leads to a discussion of performance life estimates and other design lessons for Ag/FEP thermal control material.

Emission Facilities - Erosion & Sediment Control Facilities

Data.gov (United States)

NSGIC Education | GIS Inventory — An Erosion and Sediment Control Facility is a DEP primary facility type related to the Water Pollution Control program. The following sub-facility types related to...

Erosion control works and the intensity of soil erosion in the upper part of the river Toplica drainage basin

International Nuclear Information System (INIS)

Kostadinov, S; Dragovic, N; Zlatic, M; Todosijevic, M

2008-01-01

Aiming at the protection of the future storage 'Selova' against erosion and sediment, and also to protect the settlements and roads in the drainage basin against torrential floods, erosion control works in the upper part of the river Toplica basin, upstream of the storage 'Selova', started in 1947. The works included building-technical works (check dams) and biological works (afforestation and grassing of bare lands and other erosion risk areas). Within the period 1947-2006, the following erosion control works were executed: afforestation of bare lands on the slopes 2,257.00 ha, grassing of bare lands 1,520.00 ha, and altogether 54 dams were constructed in the river Toplica tributaries. This caused the decrease of sediment transport in the main flow of the river Toplica. This paper, based on the field research conducted in two time periods: 1988 and in the period 2004-2007, presents the state of erosion in the basin before erosion control works; type and scope of erosion control works and their effect on the intensity of erosion in the river Toplica basin upstream of the future storage 'Selova'.

Forest road erosion control using multiobjective optimization

Science.gov (United States)

Matthew Thompson; John Sessions; Kevin Boston; Arne Skaugset; David Tomberlin

2010-01-01

Forest roads are associated with accelerated erosion and can be a major source of sediment delivery to streams, which can degrade aquatic habitat. Controlling road-related erosion therefore remains an important issue for forest stewardship. Managers are faced with the task to develop efficient road management strategies to achieve conflicting environmental and economic...

Airphoto analysis of erosion control practices

Science.gov (United States)

Morgan, K. M.; Morris-Jones, D. R.; Lee, G. B.; Kiefer, R. W.

1980-01-01

The Universal Soil Loss Equation (USLE) is a widely accepted tool for erosion prediction and conservation planning. In this study, airphoto analysis of color and color infrared 70 mm photography at a scale of 1:60,000 was used to determine the erosion control practice factor in the USLE. Information about contour tillage, contour strip cropping, and grass waterways was obtained from aerial photography for Pheasant Branch Creek watershed in Dane County, Wisconsin.

Categorization of erosion control matting for slope applications.

Science.gov (United States)

2013-12-25

Erosion control is an important aspect of any Georgia Department of Transportation (GDOT) construction project, with the extreme negative impacts of high sediment loads in natural waterways having been well documented. Selection of a proper erosion c...

Wind erosion control of soils using polymeric materials

Directory of Open Access Journals (Sweden)

Mohammad Movahedan

2012-07-01

Full Text Available Wind erosion of soils is one of the most important problems in environment and agriculture which could affects several fields. Agricultural lands, water reservoires, irrigation canals, drains and etc. may be affected by wind erosion and suspended particles. As a result wind erosion control needs attention in arid and semi-arid regions. In recent years, some polymeric materials have been used for improvement of structural stability, increasing aggregate stability and soil stabilization, though kind of polymer, quantity of polymer, field efficiency and durability and environmental impacts are some important parameters which should be taken into consideration. In this study, a Polyvinil Acetate-based polymer was used to treat different soils. Then polymer-added soil samples were investigated experimentally in a wind tunnel to verify the effecte of polymer on wind erosion control of the soils and the results were compared with water treated soil samples. The results of wind tunnel experiments with a maximum 26 m/s wind velocity showed that there was a significat difference between the erosion of polymer treated and water treated soil samples. Application of 25g/m2 polymer to Aeolian sands reduced the erosion of Aeolian sands samples to zero related to water treated samples. For silty and calyey soils treated by polymer, the wind erosion reduced minimum 90% in relation to water treated samples.

Can we manipulate root system architecture to control soil erosion?

Science.gov (United States)

Ola, A.; Dodd, I. C.; Quinton, J. N.

2015-09-01

Soil erosion is a major threat to soil functioning. The use of vegetation to control erosion has long been a topic for research. Much of this research has focused on the above-ground properties of plants, demonstrating the important role that canopy structure and cover plays in the reduction of water erosion processes. Less attention has been paid to plant roots. Plant roots are a crucial yet under-researched factor for reducing water erosion through their ability to alter soil properties, such as aggregate stability, hydraulic function and shear strength. However, there have been few attempts to specifically manipulate plant root system properties to reduce soil erosion. Therefore, this review aims to explore the effects that plant roots have on soil erosion and hydrological processes, and how plant root architecture might be manipulated to enhance its erosion control properties. We demonstrate the importance of root system architecture for the control of soil erosion. We also show that some plant species respond to nutrient-enriched patches by increasing lateral root proliferation. The erosional response to root proliferation will depend upon its location: at the soil surface dense mats of roots may reduce soil erodibility but block soil pores thereby limiting infiltration, enhancing runoff. Additionally, in nutrient-deprived regions, root hair development may be stimulated and larger amounts of root exudates released, thereby improving aggregate stability and decreasing erodibility. Utilizing nutrient placement at specific depths may represent a potentially new, easily implemented, management strategy on nutrient-poor agricultural land or constructed slopes to control erosion, and further research in this area is needed.

Urban Runoff: Model Ordinances for Erosion and Sediment Control

Science.gov (United States)

The model ordinance in this section borrows language from the erosion and sediment control ordinance features that might help prevent erosion and sedimentation and protect natural resources more fully.

Structural analysis under the Blanket Comparison and Selection Study

International Nuclear Information System (INIS)

Majumdar, S.

1985-01-01

Structural design procedures followed in the Blanket Comparison and Selection Study are briefly reviewed. The American Society of Mechanical Engineers Boilers and Pressure Vessels Code, Section III, Code Case N47 has been used as a design guide. Its relevance to fusion reactor applications, however, is open to question and needs to be evaluated in the future. The primary structural problem encountered in tokamak blanket designs is the high thermal stress due to surface heat flux, with fatigue being an additional concern for pulsed systems. The conflicting requirements of long erosion life and high surface heat flux capability imply that some form of stress relief in the first-wall region will be necessary. Simplified stress and fatigue crack growth analyses are presented to show that the use of orthogonally grooved first wall may be a potential solution for mitigating the thermal stress problem. A comparison of three structural alloys on the basis of both grooved and nongrooved first-wall designs is also presented. Other structural problems encountered in tokamak designs include stresses due to plasma disruptions, and magnetohydrodynamic (MHD) pressure drop in liquid-metal-cooled systems. In particular, it is shown that the maximum stress in the side wall of a uniform duct generated by MHD pressure drop cannot be reduced by increasing the wall thickness or by decreasing the span. In contract to tokamak blankets, tandem mirror blankets are far less severely stressed because of a much lower surface heat flux, coolant pressure, and also because of their axisymmetric geometry. Both blankets, however, will require detailed structural dynamics analysis to verify their ability to withstand seismic loadings if the heavy 17Li-83Pb is used as a coolant

New concepts for controlled fusion reactor blanket design

International Nuclear Information System (INIS)

Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

1975-01-01

Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

NET test blanket design and remote maintenance

International Nuclear Information System (INIS)

Holloway, C.; Hubert, P.

1991-01-01

The NET machine has three horizontal ports reserved for testing tritium breeding blanket designs during the physics phase and possibly five during the technology phase. The design of the ports and test blankets are modular to accept a range of blanket options, provide radiation shielding and allow routine replacement. Radiation levels during replacement or maintenance require that all operations must be carried out remotely. The paper describes the problems overcome in providing a port design which includes attachment to the vacuum vessel with double vacuum seals, an integrated cooled first wall and support guides for the test blanket module. The method selected to remotely replace the test module whilst controlling the spread of contamination is also adressed. The paper concludes that the provisions of a test blanket facility based on the NET machine design is feasible. (orig.)

Tectonic controls of Holocene erosion in a glaciated orogen

OpenAIRE

Adams, Byron A.; Ehlers, Todd A.

2018-01-01

Recent work has highlighted a strong, worldwide, glacial impact of orogen erosion rates over the last 2 Ma. While it may be assumed that glaciers increased erosion rates when active, the degree to which past glaciations influence Holocene erosion rates through the adjustment of topography is not known. In this study, we investigate the influence of long-term tectonic and post-glacial topographic controls on erosion in a glaciated orogen, the Olympic Mountains, USA. We present 14 new 10Be and ...

Upgrading the data acquisition and control systems of the European Breeding Blanket Test Facility

International Nuclear Information System (INIS)

Mannori, Simone; Sermenghi, Valerio; Utili, Marco; Malavasi, Andrea; Gianotti, Daniel

2013-01-01

Highlights: • Data Acquisition and Control Systems (DACS) upgrading of experimental plant for full size thermo hydraulic testing of nuclear subsystems. • DACS development using integrated hardware/software platform with graphical programming (LabVIEW). • Development of simplified models for real-time simulation. • Rapid prototyping with real time simulation of the complete plant. • Using the code developed for the real time simulator for the real plant DACS. -- Abstract: The EBBTF (European Breeding Blanket Test Facility) experimental plant is a key component for the development of the breeding blankets (TBMs test blanket modules, HCLL helium cooled lithium lead and HCPB helium cooled pebble bed types) used by ITER. EBBTF is an experimental plant which provides the double breeding/cooling loops (liquid metal and gas) required for HCLL testing. EBBTF is composed of four subsystems (TBM, IELLLO integrated European lead lithium loop, HE-FUS3 helium fusion loop, version 3 and helium compressor build by ATEKO) with dedicated control systems realized with hardware/software combinations covering 15 years (1995–2010) time span. At the end of 2010 we began to upgrade the HE-FUS3 data acquisition control systems (DACS) replacing the obsolete PLC Siemens S5 with National Instruments Compact FieldPoint and LabVIEW. The control room has been completely reorganized using high resolution monitors and workstations linked with standard Ethernet interfaces. The data acquisition, control, safety and SCADA software has been completely developed in ENEA using LabVIEW. In this paper we are going to discuss the technical difficulties and the solutions that we have used to accomplish the upgrade

Applicability of tungsten/EUROFER blanket module for the DEMO first wall

International Nuclear Information System (INIS)

Igitkhanov, Yu.; Bazylev, B.; Landman, I.; Boccaccini, L.

2013-01-01

In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m 2 . It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface

Applicability of tungsten/EUROFER blanket module for the DEMO first wall

Energy Technology Data Exchange (ETDEWEB)

Igitkhanov, Yu., E-mail: juri.igitkhanov@lhm.fzk.de [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Bazylev, B.; Landman, I. [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Boccaccini, L. [Karlsruhe Institute of Technology, INR, Karlsruhe (Germany)

2013-07-15

In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m{sup 2}. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

Applicability of tungsten/EUROFER blanket module for the DEMO first wall

Science.gov (United States)

Igitkhanov, Yu.; Bazylev, B.; Landman, I.; Boccaccini, L.

2013-07-01

In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ˜14 MW/m2. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

Energy Technology Data Exchange (ETDEWEB)

Merrill, B.J., E-mail: Brad.Merrill@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Wong, C.P.C. [General Atomics, San Diego, CA 92186-5608 (United States); Cadwallader, L.C. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Abdou, M.; Morley, N.B. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)

2014-10-15

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the {sup 210}Po and {sup 203}Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.

Design analyses of self-cooled liquid metal blankets

International Nuclear Information System (INIS)

Gohar, Y.

1986-12-01

A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations

48 CFR 436.574 - Control of erosion, sedimentation, and pollution.

Science.gov (United States)

2010-10-01

..., sedimentation, and pollution. 436.574 Section 436.574 Federal Acquisition Regulations System DEPARTMENT OF... 436.574 Control of erosion, sedimentation, and pollution. The contracting officer shall insert the clause at 452.236-74, Control of Erosion, Sedimentation and Pollution, if there is a need for applying...

An assessment of the base blanket for ITER

International Nuclear Information System (INIS)

Raffray, A.R.; Abdou, M.A.; Ying, A.

1991-01-01

Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored

Control of erosive tooth wear: possibilities and rationale

Directory of Open Access Journals (Sweden)

Mônica Campos Serra

2009-06-01

Full Text Available Dental erosion is a type of wear caused by non bacterial acids or chelation. There is evidence of a significant increase in the prevalence of dental wear in the deciduous and permanent teeth as a consequence of the frequent intake of acidic foods and drinks, or due to gastric acid which may reach the oral cavity following reflux or vomiting episodes. The presence of acids is a prerequisite for dental erosion, but the erosive wear is complex and depends on the interaction of biological, chemical and behavioral factors. Even though erosion may be defined or described as an isolated process, in clinical situations other wear phenomena are expected to occur concomitantly, such as abrasive wear (which occurs, e.g, due to tooth brushing or mastication. In order to control dental loss due to erosive wear it is crucial to take into account its multifactorial nature, which predisposes some individuals to the condition.

Materials for breeding blankets

International Nuclear Information System (INIS)

Mattas, R.F.; Billone, M.C.

1995-09-01

There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified

Materials for breeding blankets

International Nuclear Information System (INIS)

Mattas, R.F.; Billone, M.C.

1996-01-01

There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as primary blanket materials, which have the greatest influence in determining the overall design and performance, and secondary blanket materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified. (orig.)

ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

International Nuclear Information System (INIS)

WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

2002-01-01

OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

Objectives and status of EUROfusion DEMO blanket studies

Energy Technology Data Exchange (ETDEWEB)

Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

2016-11-01

Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

European DEMO BOT solid breeder blanket

International Nuclear Information System (INIS)

Dalle Donne, M.

1994-11-01

The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

Advanced high performance solid wall blanket concepts

International Nuclear Information System (INIS)

Wong, C.P.C.; Malang, S.; Nishio, S.; Raffray, R.; Sagara, A.

2002-01-01

First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

ITER blanket designs

International Nuclear Information System (INIS)

Gohar, Y.; Parker, R.; Rebut, P.H.

1995-01-01

The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

Application Of GIS Software For Erosion Control In The Watershed Scale

Directory of Open Access Journals (Sweden)

C. Setyawan

2017-01-01

Full Text Available Land degradation in form of soil erosion due to uncontrolled farming is occurred in many watersheds of Indonesia particularly in Java Island. Soil erosion is decreasing watershed function as a rainwater harvesting area. Good conservation practices need to be applied to prevent more degradation. This study aims to investigate the effectiveness of land conservation practice for erosion control through land use modeling in the watershed scale. The modeling was applied in the Sempor watershed Indonesia. Three scenarios of land use were used for modeling. Soil erosion measurement and land use modeling were performed by using Universal Soil Loss Equation USLE method and Geographic Information System GIS software ArcGIS 10.1. Land use modeling was conducted by increasing permanent vegetation coverage from existing condition 4 to 10 20 and 30. The result showed that the modeling can reduce heavy class erosion about 15-37 of total area. GIS provides a good tool for erosion control modeling in the watershed scale.

ITER convertible blanket evaluation

International Nuclear Information System (INIS)

Wong, C.P.C.; Cheng, E.

1995-01-01

Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

48 CFR 452.236-74 - Control of Erosion, Sedimentation, and Pollution.

Science.gov (United States)

2010-10-01

..., Sedimentation, and Pollution. 452.236-74 Section 452.236-74 Federal Acquisition Regulations System DEPARTMENT OF....236-74 Control of Erosion, Sedimentation, and Pollution. As prescribed in 436.574, insert the following clause: Control of Erosion, Sedimentation, and Pollution (NOV 1996) (a) Operations shall be...

Fusion-reactor blanket-material safety-compatibility studies

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

1982-11-01

Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

International Nuclear Information System (INIS)

Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse.

1988-03-01

We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

Blanket testing in NET

International Nuclear Information System (INIS)

Chazalon, M.; Daenner, W.; Libin, B.

1989-01-01

The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

Soil erosion and its control in Chile - An overview

International Nuclear Information System (INIS)

Ellies, A.

2000-01-01

Accelerate erosion in Chile is a consequence from land use that degrade soil such as compaction, loss of organic matter and soil structure. The erosion is favored by the very hilly landscape of the country that increases erosivity index and the high erodibility given by an elevated annual rate of rainfall with irregular distribution. Several experiences have demonstrated that adequate crop management and crop rotations can minimize erosion. The most effective control is achieved conserving and improving soil structure with management systems that include regular use of soil-improving crops, return of crop residues and tillage practices, thus avoiding unnecessary breakdown soil or compacted soil structure. Conservation tillage increased organic matter levels improving stabile soil structure, aeration and infiltration. (author) [es

Beach erosion control study at Pass Christian. [using remote sensors and satellite observation

Science.gov (United States)

1978-01-01

The methods of measuring the existence of erosion and the effects of sand stabilization control systems are described. The mechanics of sand movement, the nature of sand erosion, and the use of satellite data to measure these factors and their surrogates are discussed using the locational and control aspects of aeolian and litoral erosion zones along the sand beach of the Mississippi coast. The aeolian erosion is highlighted due to the redeposition of the sand which causes high cleanup costs, property damage, and safety and health hazards. The areas of differential erosion and the patterns of beach sand movement are illustrated and the use of remote sensing methods to identify the areas of erosion are evaluated.

Fusion reactor blanket-main design aspects

International Nuclear Information System (INIS)

Strebkov, Yu.; Sidorov, A.; Danilov, I.

1994-01-01

The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

Limitations on blanket performance

International Nuclear Information System (INIS)

Malang, S.

1999-01-01

The limitations on the performance of breeding blankets in a fusion power plant are evaluated. The breeding blankets will be key components of a plant and their limitations with regard to power density, thermal efficiency and lifetime could determine to a large degree the attractiveness of a power plant. The performance of two rather well known blanket concepts under development in the frame of the European Blanket Programme is assessed and their limitations are compared with more advanced (and more speculative) concepts. An important issue is the question of which material (structure, breeder, multiplier, coatings) will limit the performance and what improvement would be possible with a 'better' structural material. This evaluation is based on the premise that the performance of the power plant will be limited by the blankets (including first wall) and not by other components, e.g. divertors, or the plasma itself. However, the justness of this premise remains to be seen. It is shown that the different blanket concepts cover a large range of allowable power densities and achievable thermal efficiencies, and it is concluded that there is a high incentive to go for better performance in spite of possibly higher blanket cost. However, such high performance blankets are usually based on materials and technologies not yet developed and there is a rather high risk that the development could fail. Therefore, it is explained that a part of the development effort should be devoted to concepts where the materials and technologies are more or less in hand in order to ensure that blankets for a DEMO reactor can be developed and tested in a given time frame. (orig.)

Fusion blanket design and optimization techniques

International Nuclear Information System (INIS)

Gohar, Y.

2005-01-01

In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

ITER shielding blanket

Energy Technology Data Exchange (ETDEWEB)

Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

1995-03-01

A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

Fusion fuel blanket technology

International Nuclear Information System (INIS)

Hastings, I.J.; Gierszewski, P.

1987-05-01

The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

Development of blanket remote maintenance system

International Nuclear Information System (INIS)

Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou

1998-01-01

ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

Development of blanket remote maintenance system

Energy Technology Data Exchange (ETDEWEB)

Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-04-01

ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

Tritium breeding blanket

International Nuclear Information System (INIS)

Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

1991-01-01

The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

Novel blanket design for ICTR's

International Nuclear Information System (INIS)

Abdel-Khalik, S.I.; Conn, R.W.; Wolfer, W.G.; Larsen, E.N.; Sviatoslavsky, I.N.

1978-01-01

A novel blanket design for ICTRs is described. This blanket is used in SOLASE, the conceptual laser fusion reactor of the University of Wisconsin. The blanket to be described offers numerous advantages, including low cost, low weight, low induced radioactivity levels, the potential for hands-on maintenance, modular construction, low pressure, ability to decouple first wall and blanket coolant temperatures, adequate breeding, low tritium inventory and leakage, and sufficiently long life

Mirror reactor blankets

International Nuclear Information System (INIS)

Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

1976-01-01

The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

The success of headwater rehabilitation towards gully erosion control

Science.gov (United States)

Frankl, Amaury; Poesen, Jean; Nyssen, Jan

2017-04-01

The ill-management of headwaters has frequently shown to have adverse effects on both humans and the environment. Historical examples often refer to altered hydrological conditions and stream incision resulting from deforestation. Agricultural expansion and intensification - often accompanied with land reforms in the 20th century - also showed to severely impact the fluvial environment, with stream incision and gully erosion hazards increasingly affecting many headwater areas around the world. To counter this, many regions have adopted improved management schemes aiming at restoring the physical, biological and hydrological integrity of the soil- and landscape. In terms of hydrogeomorpology, the objective was to minimize dynamics to a lower level so that runoff, sediment and pollutant transfers do not cause danger to human life, environmental/natural resources deterioration or economic stress. Therefore, much attention was given to the rehabilitation and re-naturalization of headwater streams and gullies, which are the conduits of these transfers. This is done in both indirect and direct ways, i.e. reducing the delivery of runoff and sediment to the gullies and interventions in the incised channels. Although much has been published on gully erosion development and control, few studies assess the success of gully rehabilitation on the mid- to long term or confront results against the gully life-cycle. The latter refers to the rate law in fluvial geomorphology, whereby gully morphological changes (increases in length, area, volume) are initially rapid, followed by a much slower development towards a new equilibrium state. Here, we present a review of headwater rehabilitation measures and their success towards gully erosion control. By confronting this to the life-cycle of a gully, we also want to shed light on our understanding of when and where gully erosion control needs to be applied; making land management more efficient and effective. Keywords: land

Method of operating water cooled reactor with blanket

International Nuclear Information System (INIS)

Suzuki, Katsuo.

1988-01-01

Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

Wind Erosion Processes and Control Techniques in the Sahelian Zone of Niger

NARCIS (Netherlands)

Sterk, G.; Stroosnijder, L.; Raats, P.A.C.

1999-01-01

Wind Erosion Processes and Control Techniques in the Sahelian Zone of Niger G. Sterk, L. Stroosnijder, and P.A.C. Raats Abstract The objective of this paper is to present the main results and conclusions from three years of field research on wind erosion processes and control techniques in the

Trade-off study of liquid-metal self-cooled blankets

International Nuclear Information System (INIS)

Gohar, Y.

1986-01-01

A trade-off study of liquid-metal self-cooled blankets was carried out to define the performance of these blankets with respect to the main functions in a fusion reactor, and to determine the potential to operate at the maximum possible values of the performance parameters. The main purpose is to improve the reactor economics by maximizing the blanket energy multiplication factor, reduce the capital cost of the reactor, and satisfy the design requirements. The main parameters during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the 6 Li enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, the impact of different reactor design choices on the performance parameters was analyzed. The effect of the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, the coolant choice for the nonbreeding inboard blanket, and the neutron source distribution were part of the trade-off study. In addition, tritium breeding benchmark calculations were performed to study the impact of the use of different transport codes and nuclear data libraries. The importance and the negative effect of high TBR on the energy multiplication motivated the benchmark calculations

Application of the system of water erosion control measures in growths of special cultivations

Directory of Open Access Journals (Sweden)

Vítězslav Hálek

2004-01-01

Full Text Available The aim of the study is to select an optimal variant of the system of water erosion control measures. The water erosion issue was observed and evaluated in 15 blocks of special cultivations-vineyards and orchards. These blocks are situated in the managed area of the join-stock company PATRIA Kobylí. At first the average long-term loss of soil with the influence of water erosion is calculated. The universal Wischmeier-Smith equation is used for this purpose. If the calculated loss of soil exceeds the permissible value, the erosion control measures have to be suggested. The optimal variant has been selected on the bases of the evaluation of several kinds of measures in each block. This variant follows first of all the erosion control efficiency, but also demands on production as well as slope accessibility for mechanization, expensiveness and some negative sides of suggested measures. The suggested system of water erosion control measures contributes to increasing of soil fertility and production ability with the respect to landscape management and environmental protection.

An evaluation of fast reactor blankets

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1974-01-01

A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

Blanket for thermonuclear device

International Nuclear Information System (INIS)

Ozawa, Yoshihiro; Uda, Tatsuhiko; Maki, Koichi.

1993-01-01

The present invention provides a blanket of a thermonuclear device which produces tritium fuels consumed in plasmas while converting neutrons generated in the plasmas into heat energy. That is, zirconium is coated to at least one of neutron breeder pebbles and breeder pebbles, to suppress reaction between them by being in direct contact with each other at a high temperature. Further, fins are attached to a cooling pipe at a pitch smaller than the diameter of both of the pebbles, to prevent direct contact at whole surface of the pebbles and the cooling pipe, which would lower a temperature excessively. The length of the fin is controlled to control the thickness of a helium gas gap. With such constitution, direct contact of neutron breeder pebbles and the breeder pebble which are to be filled and mixed, and tend to react at a high temperature, can be prevented. The temperature of the breeding blanket is reliably prevented from lowering below a tritium emitting temperature. The structure is simplified and the production is facilitated. (I.S.)

Status of fusion reactor blanket design

International Nuclear Information System (INIS)

Smith, D.L.; Sze, D.K.

1986-02-01

The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie

1998-01-01

Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

Modeling of technical soil-erosion control measures and its impact on soil erosion off-site effects within urban areas

Science.gov (United States)

Dostal, Tomas; Devaty, Jan

2013-04-01

The paper presents results of surface runoff, soil erosion and sediment transport modeling using Erosion 3D software - physically based mathematical simulation model, event oriented, fully distributed. Various methods to simulate technical soil-erosion conservation measures were tested, using alternative digital elevation models of different precision and resolution. Ditches and baulks were simulated by three different approaches, (i) by change of the land-cover parameters to increase infiltration and decrease flow velocity, (ii) by change of the land-cover parameters to completely infiltrate the surface runoff and (iii) by adjusting the height of the digital elevation model by "burning in" the channels of the ditches. Results show advantages and disadvantages of each approach and conclude suitable methods for combinations of particular digital elevation model and purpose of the simulations. Further on a set of simulations was carried out to model situations before and after technical soil-erosion conservation measures application within a small catchment of 4 km2. These simulations were focused on quantitative and qualitative assessment of technical soil-erosion control measures impact on soil erosion off-site effects within urban areas located downstream of intensively used agricultural fields. The scenarios were built upon a raster digital elevation model with spatial resolution of 3 meters derived from LiDAR 5G vector point elevation data. Use of this high-resolution elevation model allowed simulating the technical soil-erosion control measures by direct terrain elevation adjustment. Also the structures within the settlements were emulated by direct change in the elevation of the terrain model. The buildings were lifted up to simulate complicated flow behavior of the surface runoff within urban areas, using approach of Arévalo (Arévalo, 2011) but focusing on the use of commonly available data without extensive detailed editing. Application of the technical

Exploring climatic controls on blanket bog litter decomposition across an altitudinal gradient

Science.gov (United States)

Bell, Michael; Ritson, Jonathan P.; Clark, Joanna M.; Verhoef, Anne; Brazier, Richard E.

2016-04-01

The hydrological and ecological functioning of blanket bogs is strongly coupled, involving multiple ecohydrological feedbacks which can affect carbon cycling. Cool and wet conditions inhibit decomposition, and favour the growth of Sphagnum mosses which produce highly recalcitrant litter. A small but persistent imbalance between production and decomposition has led to blanket bogs in the UK accumulating large amounts of carbon. Additionally, healthy bogs provide a suite of other ecosystems services including water regulation and drinking water provision. However, there is concern that climate change could increase rates of litter decomposition and disrupt this carbon sink. Furthermore, it has been argued that the response of these ecosystems in the warmer south west and west of the UK may provide an early analogue for later changes in the more extensive northern peatlands. In order to investigate the effects of climate change on blanket bog litter decomposition, we set-up a litter bag experiment across an altitudinal gradient spanning 200 m of elevation (including a transition from moorland to healthy blanket bog) on Dartmoor, an area of hitherto unstudied, climatically marginal blanket bog in the south west of the UK. At seven sites, water table depth and soil and surface temperature were recorded continuously. Litter bags filled with the litter of three vegetation species dominant on Dartmoor were incubated just below the bog surface and retrieved over a period of 12 months. We found significant differences in the rate of decomposition between species. At all sites, decomposition progressed in the order Calluna vulgaris (dwarf shrub) > Molinia caerulea (graminoid) > Sphagnum (bryophyte). However, while soil temperature did decrease along the altitudinal gradient, being warmer in the lower altitudes, a hypothesised accompanying decrease in decomposition rates did not occur. This could be explained by greater N deposition at the higher elevation sites (estimated

Use of Low-Cost Methods of Soil Erosion Control In Kisii District, South Western kenya

International Nuclear Information System (INIS)

Nzabi, A.W; Makini, F; Onyango, M; Mureithi, J.G

1999-01-01

Kisii District has a topography of undulating hills and is prone to severe soil erosion. The average rainfall is 1900 mm and occurs in biomodal pattern. During a participatory appraisal survey in 1995, farmers indicated that soil erosion in the area had contributed to decline in soil fertility resulting in low crop yields. To address this problem, an on-farm trial was conducted in 1996 at Nyamonyo village to test the effectiveness of four low cost methods of controlling soil erosion. These included maize stover trash line, sweet potatoes,Penicum maximum var. Makarikari grass strip and vetiveria zizanioides (Vertiver) grass strip. A treatment without soil erosion control measure was included. The trial was planted in three farms which acted as replicates. The treatments were planted in runoff plots measuring 4 x 2 m in which had a maize crop were laid down in a randomized complete block design. Surface runoff and eroded soils were collected in 50-l buckets. The experimental site had a slope ranging from 16 to 35%. Preliminary results indicated that maize stover trash line and sweet potato strips were more effective in controlling soil erosion than the grass strips. As the season progressed the grass strips became increasingly more effective in erosion control. The trail is still continuing but results indicate that for short term soil erosion control, maize stover trash lines and sweet potatoes are more effective while Makarikari and Vertiver grass strips are promising as long term soil erosion control measure

Wind erosion processes and control

Science.gov (United States)

Wind erosion continues to threaten the sustainability of our nations' soil, air, and water resources. To effectively apply conservation systems to prevent wind driven soil loss, an understanding of the fundamental processes of wind erosion is necessary so that land managers can better recognize the ...

(D,T) Driven thorium hybrid blankets

International Nuclear Information System (INIS)

Al-Kusayer, T.A.; Khan, S.; Sahin, S.

1983-01-01

Recently, a project has started, with the aim to establish the neutronic performance and the basic design of an experimental fusionfission (hybrid) reactor facility, called AYMAN, in cylinderical geometry. The fusion reactor will have to be simulated by a (D,T) neutron generator. Fissile and fertile fuel will have to surround the neutron generator as a cylinderical blanket to simulate the boundary conditions of the hybrid blanket in a proper way. This geometry is consistent with Tandem Mirror Hybrid Blanket design and with most of the ICF blanket designs. A similar experimental installation will become operational around 1984 at the Swiss Federal Institute of Technology in Lausanne, Switzerland known under the project LOTUS. Due to the limited dimensions of the experimental cavity of the LOTUS-hybrid reactor, the LOTUS blankets have to be designed in plane geometry. Also, the bulky form of the Haefely neutron generator of the LOTUS facility obliges one to design a blanket in the plane geometry. This results in a vacuum left boundary conditions for the LOTUS blanket. The importance of a reflecting left boundary condition on the overall neutronic performance of a hybrid blanket has been analyzed in previous work in detail

Concepts for fusion fuel production blankets

International Nuclear Information System (INIS)

Gierszewski, P.

1986-06-01

The fusion blanket surrounds the burning hydrogen core of the fusion reactor. It is in this blanket that most of the energy released by the DT fusion reaction is converted into useable product, and where tritium fuel is produced to enable further operation of the reactor. Blankets will involve new materials, conditions and processes. Several recent fusion blanket concepts are presented to illustrate the range of ideas

Breeding blanket for Demo

International Nuclear Information System (INIS)

Proust, E.; Giancarli, L.

1992-01-01

This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme

Overview of EU activities on DEMO liquid metal breeder blanket

International Nuclear Information System (INIS)

Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

1994-01-01

The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

Heating an aquaculture pond with a solar pool blanket

Energy Technology Data Exchange (ETDEWEB)

Wisely, B; Holliday, J E; MacDonald, R E

1982-01-01

A floating solar blanket of laminated bubble plastic was used to heat a 0.11 ha seawater pond of 1.3 m depth. The covered pond maintained daily temperatures 6 to 9/sup 0/C above two controls. Local air temperatures averaged 14 to 19/sup 0/C. Oysters, prawns, seasquirts, and fish in the covered pond all survived. After three weeks, the blanket separated. This was the result of pond temperatures exceeding 30/sup 0/C, the maximum manufacturer's specification. Floating blankets fabricated to higher specifications would be useful for maintaining above-ambient temperatures in small ponds or tanks in temporary situations during cold winter months and might have a more permanent use.

Robotic weld overlay coatings for erosion control

Science.gov (United States)

The erosion of materials by the impact of solid particles has received increasing attention during the past twenty years. Recently, research has been initiated with the event of advanced coal conversion processes in which erosion plays an important role. The resulting damage, termed Solid Particle Erosion (SPE), is of concern primarily because of the significantly increased operating costs which result in material failures. Reduced power plant efficiency due to solid particle erosion of boiler tubes and waterfalls has led to various methods to combat SPE. One method is to apply coatings to the components subjected to erosive environments. Protective weld overlay coatings are particularly advantageous in terms of coating quality. The weld overlay coatings are essentially immune to spallation due to a strong metallurgical bond with the substrate material. By using powder mixtures, multiple alloys can be mixed in order to achieve the best performance in an erosive environment. However, a review of the literature revealed a lack of information on weld overlay coating performance in erosive environments which makes the selection of weld overlay alloys a difficult task. The objective of this project is to determine the effects of weld overlay coating composition and microstructure on erosion resistance. These results will lead to a better understanding of erosion mitigation in CFB's.

Blankets for thermonuclear device

International Nuclear Information System (INIS)

Maki, Koichi; Fukumoto, Hideshi.

1986-01-01

Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

Minimum thickness blanket-shield for fusion reactors

International Nuclear Information System (INIS)

Karni, Y.; Greenspan, E.

1989-01-01

A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

Cloud forest restoration for erosion control in a Kichwa community of the Ecuadorian central Andes Mountains

Science.gov (United States)

Backus, L.; Giordanengo, J.; Sacatoro, I.

2013-12-01

The Denver Professional Chapter of Engineers Without Borders (EWB) has begun conducting erosion control projects in the Kichwa communities of Malingua Pamba in the Andes Mountains south of Quito, Ecuador. In many high elevation areas in this region, erosion of volcanic soils on steep hillsides (i.e., food crops. Following a 2011 investigation of over 75 erosion sites, the multidisciplinary Erosion Control team traveled to Malingua Pamba in October 2012 to conduct final design and project implementation at 5 sites. In partnership with the local communities, we installed woody cloud forest species, grass (sig-sig) contour hedges, erosion matting, and rock structures (toe walls, plunge pools, bank armoring, cross vanes, contour infiltration ditches, etc.) to reduce incision rates and risk of slump failures, facilitate aggradation, and hasten revegetation. In keeping with the EWB goal of project sustainability, we used primarily locally available resources. High school students of the community grew 5000 native trees and some naturalized shrubs in a nursery started by the school principal, hand weavers produced jute erosion mats, and rocks were provided by a nearby quarry. Where possible, local rock was harvested from landslide areas and other local erosion features. Based on follow up reports and photographs from the community and EWB travelers, the approach of using locally available materials installed by the community is successful; plants are growing well and erosion control structures have remained in place throughout the November to April rainy season. The community has continued planting native vegetation at several additional erosion sites. Formal monitoring will be conducted in October 2013, followed by analysis of data to determine if induced meandering and other low-maintenance erosion control techniques are working as planned. For comparison of techniques, we will consider installing check dams in comparable gullies. The October 2013 project will also

Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

International Nuclear Information System (INIS)

Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

2006-01-01

In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

Disruption problematics in segmented blanket concepts

International Nuclear Information System (INIS)

Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

1994-01-01

In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

Analyses of Hubble Space Telescope Aluminized-Teflon Multilayer Insulation Blankets Retrieved After 19 Years of Space Exposure

Science.gov (United States)

de Groh, Kim K.; Perry, Bruce A.; Mohammed, Jelila S.; Banks, Bruce

2015-01-01

Since its launch in April 1990, the Hubble Space Telescope (HST) has made many important observations from its vantage point in low Earth orbit (LEO). However, as seen during five servicing missions, the outer layer of multilayer insulation (MLI) has become increasingly embrittled and has cracked in many areas. In May 2009, during the 5th servicing mission (called SM4), two MLI blankets were replaced with new insulation and the space-exposed MLI blankets were retrieved for degradation analyses by teams at NASA Glenn Research Center (GRC) and NASA Goddard Space Flight Center (GSFC). The retrieved MLI blankets were from Equipment Bay 8, which received direct sunlight, and Equipment Bay 5, which received grazing sunlight. Each blanket was divided into several regions based on environmental exposure and/or physical appearance. The aluminized-Teflon (DuPont, Wilmington, DE) fluorinated ethylene propylene (Al-FEP) outer layers of the retrieved MLI blankets have been analyzed for changes in optical, physical, and mechanical properties, along with chemical and morphological changes. Pristine and as-retrieved samples (materials) were heat treated to help understand degradation mechanisms. When compared to pristine material, the analyses have shown how the Al-FEP was severely affected by the space environment. Most notably, the Al-FEP was highly embrittled, fracturing like glass at strains of 1 to 8 percent. Across all measured properties, more significant degradation was observed for Bay 8 material as compared to Bay 5 material. This paper reviews the tensile and bend-test properties, density, thickness, solar absorptance, thermal emittance, x-ray photoelectron spectroscopy (XPS) and energy dispersive spectroscopy (EDS) elemental composition measurements, surface and crack morphologies, and atomic oxygen erosion yields of the Al-FEP outer layer of the retrieved HST blankets after 19 years of space exposure.

ARIES-IV Nested Shell Blanket Design

International Nuclear Information System (INIS)

Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

1993-11-01

The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design

North Fork Feather River Erosion Control Program

International Nuclear Information System (INIS)

Harrison, L.

1991-01-01

PG and E, an investor owned gas and electric utility serving northern and central California, has been engaged since 1984 in the development and implementation of a regional erosion control program for the 954 square mile northern Sierra Nevada watersheed of the East Branch of the North Fork Feather River in Plumas County, California. PG and E entered into an agreement with 13 governmental agencies and a number of private landowners using Coordinated Resource Management and Planning: to cooperatively develop, fund and implement the program. The group has completed several field projects and has a number of additional projects in various stages of development. This paper reports that the program provides multiple environmental and economic benefits including reduction of soil erosion and sedimentation, improved fisheries, enhancement of riparian habitat, increased land values, improved recreation opportunities, and preservation of watershed resources

Main maintenance operations for Test Blanket Systems in ITER TBM port cells

Energy Technology Data Exchange (ETDEWEB)

Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Friconneau, J.-P.; Giancarli, L.M.; Gotewal, K.K.; Iseli, M.; Kim, B.Y.; Levesy, B.; Martins, J.-P.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Siarras, A. [Sogetti, Parc de la Duranne, 13857 Aix-en-Provence (France); Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.

LMFBR blanket physics project progress report No. 4

International Nuclear Information System (INIS)

Driscoll, M.J.; Lanning, D.D.; Kaplan, I.; Supple, A.T.

1973-01-01

During the period covered by the report, July 1, 1972, through June 30, 1973, work was devoted to completion of experimental measurements and data analysis on Blanket Mockup No. 3, a graphite-reflected blanket, and to initiation of experimental work on Blanket Mockup No. 4, a steel-reflected assembly designed to mock up a demonstration plant blanket. Work was also carried out on the analysis of a number of advanced blanket concepts, including the use of high-albedo reflectors, the use of thorium in place of uranium in the blanket region, and the ''parfait'' or completely internal blanket concept. Finally, methods development work was initiated to develop the capability for making gamma heating measurements in the blanket mockups. (U.S.)

Literature review of models for estimating soil erosion and deposition from wind stresses on uranium-mill-tailings covers

International Nuclear Information System (INIS)

Bander, T.J.

1982-11-01

Pacific Northwest Laboratory (PNL) is investigating the use of a rock armoring blanket (riprap) to mitigate wind and water erosion of an earthen radon-suppression cover applied to uranium-mill tailings. The mechanics of wind erosion, as well as of soil deposition, are discussed in this report. Several wind erosion models are reviewed to determine if they can be used to estimate the erosion of soil from a mill-tailings cover. One model, developed by W.S. Chepil, contains the most-important factors that describe variables that influence wind erosion. Particular features of other models are also discussed, as well as the application of Chepil's model to a particular tailings pile. For this particular tailings pile, the estimated erosion was almost one inch per year for an unprotected tailings soil surface. Wide variability in the deposition velocity and lack of adequate deposition models preclude reliable estimates of the rate at which airborne particles are deposited

Literature review of models for estimating soil erosion and deposition from wind stresses on uranium-mill-tailings covers

Energy Technology Data Exchange (ETDEWEB)

Bander, T.J.

1982-11-01

Pacific Northwest Laboratory (PNL) is investigating the use of a rock armoring blanket (riprap) to mitigate wind and water erosion of an earthen radon-suppression cover applied to uranium-mill tailings. The mechanics of wind erosion, as well as of soil deposition, are discussed in this report. Several wind erosion models are reviewed to determine if they can be used to estimate the erosion of soil from a mill-tailings cover. One model, developed by W.S. Chepil, contains the most-important factors that describe variables that influence wind erosion. Particular features of other models are also discussed, as well as the application of Chepil's model to a particular tailings pile. For this particular tailings pile, the estimated erosion was almost one inch per year for an unprotected tailings soil surface. Wide variability in the deposition velocity and lack of adequate deposition models preclude reliable estimates of the rate at which airborne particles are deposited.

Annual report of the CTR Blanket Engineering research facility in 1994

International Nuclear Information System (INIS)

1995-09-01

This is an annual report of the studies on Controlled Thermo-nuclear Reactor(CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1994. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

Annual report of the CTR blanket engineering research facility in 1993

International Nuclear Information System (INIS)

1994-08-01

This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1993. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

Wind tunnel experimental study on the effect of PAM on soil wind erosion control.

Science.gov (United States)

He, Ji-Jun; Cai, Qiang-Guo; Tang, Ze-Jun

2008-10-01

In recent years, high-molecular-weight anionic polyacrylamide (PAM) have been widely tested on a variety of soils, primarily in water erosion control. However, little information is available regarding the effectiveness of PAM on preventing soil loss from wind erosion. The research adopted room wind tunnel experiment, two kinds of soils were used which were from the agro-pastoral area of Inner Mongolia, the northwest of China, the clay content of soils were 22.0 and 13.7%, respectively. For these tests, all the treatments were performed under the condition of wind velocity of 14 m s(-1) and a blown angle of 8.75%, according to the actual situation of experimented area. The study results indicated that using PAM on the soil surface could enhance the capability of avoiding the wind erosion, at the same time, the effect of controlling wind soil erosion with 4 g m(-2) PAM was better than 2 g m(-2) PAM's. Economically, the 2 g m(-2) PAM used in soil surface can control wind erosion effectively in this region. The prophase PAM accumulated in soil could not improve the capability of avoiding the wind erosion, owing to the degradation of PAM in the soil and the continual tillage year after year. The texture of soil is a main factor influencing the capability of soil avoiding wind erosion. Soil with higher clay content has the higher capability of preventing soil from wind erosion than one with the opposite one under the together action of PAM and water.

Soil erosion and sediment control laws. A review of state laws and their natural resource data requirements

Science.gov (United States)

Klein, S. B.

1980-01-01

Twenty states, the District of Columbia, and the Virgin Islands enacted erosion and sediment control legislation during the past decade to provide for the implementation or the strengthening of statewide erosion and sediment control plans for rural and/or urban lands. That legislation and the state programs developed to implement these laws are quoted and reviewed. The natural resource data requirements of each program are also extracted. The legislation includes amendments to conservation district laws, water quality laws, and erosion and sediment control laws. Laws which provides for legislative review of administrative regulations and LANDSAT applications and/or information systems that were involved in implementing or gathering data for a specific soil erosion and sediment control program are summarized as well as principal concerns affecting erosion and sediment control laws.

Self-cooled liquid-metal blanket concept

International Nuclear Information System (INIS)

Malang, S.; Arheidt, K.; Barleon, L.

1988-01-01

A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

Blanket comparison and selection study. Volume II

International Nuclear Information System (INIS)

1983-10-01

This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies

profitability of soil erosion control technologies in eastern uganda

African Journals Online (AJOL)

Prof. Adipala Ekwamu

The lack of farmer awareness of costs and benefits associated with the use of sustainable land management (SLM) .... land under soil erosion control technologies, cost of labour and ..... and promotion of quality protein maize hybrids in Ghana.

Fusion blanket inherent safety assessment

International Nuclear Information System (INIS)

Sze, D.K.; Jung, J.; Cheng, E.T.

1986-01-01

Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

Blanket materials for DT fusion reactors

International Nuclear Information System (INIS)

Smith, D.L.

1981-01-01

This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

Blanket maintenance by remote means using the cassette blanket approach

International Nuclear Information System (INIS)

Werner, R.W.

1978-01-01

Induced radioactivity in the blanket and other parts of a fusion reactor close to the plasma zone will dictate remote assembly, disassembly, and maintenance procedures. Time will be of the essence in these procedures. They must be practicable and certain. This paper discusses the reduction of a complicated Tokamak reactor to a simpler assembly via the use of a vacuum building in which to house the reactor and the introduction in this new model of cassette blanket modules. The cassettes significantly simplify remote handling

Annual report of the CTR Blanket Engineering research facility in 1992

International Nuclear Information System (INIS)

1993-08-01

This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1992. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

Annual report of the CTR Blanket Engineering research facility in 1996

International Nuclear Information System (INIS)

1998-02-01

This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1996. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

Control of Sound Radiation and Reflection With Advanced Smart Foam Blankets

National Research Council Canada - National Science Library

Fuller, Chris

2003-01-01

.... The past few years of the project have demonstrated the high potential of using smart foam blankets for efficiently reducing the interior sound levels in the payloads of launch vehicles in the low...

Design requirement on HYPER blanket fuel assembly

International Nuclear Information System (INIS)

Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

2000-07-01

This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

Estimating soil erosion risk and evaluating erosion control measures for soil conservation planning at Koga watershed in the highlands of Ethiopia

Science.gov (United States)

Molla, Tegegne; Sisheber, Biniam

2017-01-01

Soil erosion is one of the major factors affecting sustainability of agricultural production in Ethiopia. The objective of this paper is to estimate soil erosion using the universal soil loss equation (RUSLE) model and to evaluate soil conservation practices in a data-scarce watershed region. For this purpose, soil data, rainfall, erosion control practices, satellite images and topographic maps were collected to determine the RUSLE factors. In addition, measurements of randomly selected soil and water conservation structures were done at three sub-watersheds (Asanat, Debreyakob and Rim). This study was conducted in Koga watershed at upper part of the Blue Nile basin which is affected by high soil erosion rates. The area is characterized by undulating topography caused by intensive agricultural practices with poor soil conservation practices. The soil loss rates were determined and conservation strategies have been evaluated under different slope classes and land uses. The results showed that the watershed is affected by high soil erosion rates (on average 42 t ha-1 yr-1), greater than the maximum tolerable soil loss (18 t ha-1 yr-1). The highest soil loss (456 t ha-1 yr-1) estimated from the upper watershed occurred on cultivated lands of steep slopes. As a result, soil erosion is mainly aggravated by land-use conflicts and topographic factors and the rugged topographic land forms of the area. The study also demonstrated that the contribution of existing soil conservation structures to erosion control is very small due to incorrect design and poor management. About 35 % out of the existing structures can reduce soil loss significantly since they were constructed correctly. Most of the existing structures were demolished due to the sediment overload, vulnerability to livestock damage and intense rainfall. Therefore, appropriate and standardized soil and water conservation measures for different erosion-prone land uses and land forms need to be implemented in Koga

Field studies of erosion-control technologies for arid shallow land-burial sites at Los Alamos

International Nuclear Information System (INIS)

Nyhan, J.W.; Abeele, W.V.; DePoorter, G.L.; Hakonson, T.E.; Perkins, B.A.; Foster, G.R.

1983-01-01

The field research program involving corrective measures technologies for arid shallow land-burial sites is described. Research performed for a portion of this task, the identification, evaluation, and modeling of erosion control technologies, is presented in detail. In a joint study with USDA-ARS, soil erosion and infiltration of water into a simulated trench cap with various surface treatments was measured and compared with data from undisturbed soil surfaces with natural plant cover. The distribution of soil particles in the runoff was measured for inclusion in CREAMS (a field scale model for Chemicals, Runoff and Erosion from Agricultural Management Systems). Neutron moisture gauge data collected beneath the erosion plots are presented to show the seasonal effects of the erosion control technologies on the subsurface component of water balance. 12 references, 4 figures, 4 tables

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Bioengineering Technology to Control River Soil Erosion using Vetiver (Vetiveria Zizaniodes)

Science.gov (United States)

Sriwati, M.; Pallu, S.; Selintung, M.; Lopa, R.

2018-04-01

Erosion is the action of surface processes (such as water flow or wind) that removes soil, rock or dissolved material from one location on the earth’s crust, and then transport it away to another location. Bioengineering is an attempt to maximise the use of vegetation components along riverbanks to cope with landslides and erosion of river cliffs and another riverbank damage. This study aims to analyze the bioengineering of Vetiver as a surface layer for soil erosion control using slope of 100, 200, and 300. This study is conducted with 3 variations of rain intensity (I), at 103 mm/hour, 107 mm/hour, and 130 mm/hour by using rainfall simulator tool. In addition, the USLE (Universal Soil Loss Equation) method is used in order to measure the rate of soil erosion. In this study, there are few USLE model parameters were used such as rainfall erosivity factor, soil erodibility factor, length-loss slope and stepness factor, cover management factor, and support practise factor. The results demonstrated that average of reduction of erosion rate using Vetiver, under 3 various rainfalls, namely rainfall intensity 103 mm/hr had reduced 84.971%, rainfall intensity 107 mm/hr had reduced 86.583 %, rainfall intensity 130 mm/hr had reduced 65.851%.

Erosion-corrosion

International Nuclear Information System (INIS)

Aghili, B.

1999-05-01

A literature study on erosion-corrosion of pipings in the nuclear industry was performed. Occurred incidents are reviewed, and the mechanism driving the erosion-corrosion is described. Factors that influence the effect in negative or positive direction are treated, as well as programs for control and inspection. Finally, examples of failures from databases on erosion-corrosion are given in an attachment

Low technology high tritium breeding blanket concept

International Nuclear Information System (INIS)

Gohar, Y.; Baker, C.C.; Smith, D.L.

1987-10-01

The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

Tritium transport analysis for CFETR WCSB blanket

Energy Technology Data Exchange (ETDEWEB)

Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

2017-01-15

Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

Blanket comparison and selection study. Volume I

International Nuclear Information System (INIS)

1983-10-01

The objectives of the Blanket Comparison and Selection Study (BCSS) can be stated as follows: (1) Define a small number (approx. 3) of blanket design concepts that should be the focus of the blanket R and D program. A design concept is defined by the selection of all materials (e.g., breeder, coolant, structure and multiplier) and other major characteristics that significantly influence the R and D requirements. (2) Identify and prioritize the critical issues for the leading blanket concepts. (3) Provide the technical input necessary to develop a blanket R and D program plan. Guidelines for prioritizing the R and D requirements include: (a) critical feasibility issues for the leading blanket concepts will receive the highest priority, and (b) for equally important feasibility issues, higher R and D priority will be given to those that require minimum cost and short time

Trade-off study of liquid metal self-cooled blankets

International Nuclear Information System (INIS)

Gohar, Y.

1986-01-01

A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of this study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. The primary results of the study are as follows: a) the lithium-lead blanket achieves a higher TBR with a smaller blanket thickness relative to the lithium blanket; b) the lithium blanket generates more energy per fusion neutron relative to the lithium-lead blanket; c) among the possible reflector materials, the carbon reflector produces the highest TBR; d) the high-Z reflector materials (Mo, Cu, W, or steel) generate more energy per fusion neutron and produce smaller TBRs relative to the carbon reflector; e) lithium-6 enrichment is required for the lithium-lead blanket to reduce the total blanket thickness; and f) the energy deposition per fusion neutron reaches a saturation as the blanket thickness, the fraction of the high-Z material in the reflector, or the reflector zone thickness increases (this allows one to design the blanket for a specific TBR without reducing the energy production)

The blanket interface to TSTA

International Nuclear Information System (INIS)

Clemmer, R.G.; Finn, P.A.; Grimm, T.L.; Sze, D.K.; Anderson, J.L.; Bartlit, J.R.; Naruse, Y.; Yoshida, H.

1988-01-01

The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

Convertible shielding to ceramic breeding blanket

International Nuclear Information System (INIS)

Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

1995-05-01

Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

Classification Using Markov Blanket for Feature Selection

DEFF Research Database (Denmark)

Zeng, Yifeng; Luo, Jian

2009-01-01

Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

Development of a virtual reality simulator for the ITER blanket remote handling system

International Nuclear Information System (INIS)

Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi; Tesini, Alessandro

2008-01-01

The authors developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robotic simulation software, ENVISION. The simulator is connected to the control system of the manipulator, which was developed as part of the blanket maintenance system during the Engineering Design Activity (EDA), and can reconstruct the positions of the manipulator and blanket module using position data transmitted from motors through a LAN. In addition, it can provide virtual visual information (e.g., about the interface structures behind the blanket module) by making the module transparent on the screen. It can also be used for confirming a maintenance sequence before the actual operation. The simulator will be modified further, with addition of other necessary functions, and will finally serve as a prototype of the actual simulator for the blanket remote handling system, which will be procured as part of an in-kind contribution

Test Blanket Working Group's recent activities

International Nuclear Information System (INIS)

Vetter, J.E.

2001-01-01

The ITER Test Blanket Working Group (TBWG) has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out until July 1998, which formed part of the ITER Final Design Report (FDR), completed in 1998. Whilst the machine parameters for ITER-FEAT have been significantly revised compared to the FDR, testing of breeding blanket modules remains a main objective of the test programme and the development of a reactor-relevant breeding blanket to ensure tritium fuel self-sufficiency is recognized a key issue for fusion. Design work and R and D on breeding blanket concepts, including co-operation with the other Contacting Parties of the ITER-EDA for testing these concepts in ITER, are included in the work plans of the Parties

Achievements of element technology development for breeding blanket

International Nuclear Information System (INIS)

Enoeda, Mikio

2005-03-01

Japan Atomic Energy Research Institute (JAERI) has been performing the development of breeding blanket for fusion power plant, as a leading institute of the development of solid breeder blankets, according to the long-term R and D program of the blanket development established by the Fusion Council of Japan in 1999. This report is an overview of development plan, achievements of element technology development and future prospect and plan of the development of the solid breeding blanket in JAERI. In this report, the mission of the blanket development activity in JAERI, key issues and roadmap of the blanket development have been clarified. Then, achievements of the element technology development were summarized and showed that the development has progressed to enter the engineering testing phase. The specific development target and plan were clarified with bright prospect. Realization of the engineering test phase R and D and completion of ITER test blanket module testing program, with universities/NIFS cooperation, are most important steps in the development of breeding blanket of fusion power demonstration plant. (author)

Estimating surface soil erosion losses and mapping erosion risk for Yusufeli micro-catchment (Artvin

Directory of Open Access Journals (Sweden)

Mustafa Tüfekçioğlu

2016-10-01

Full Text Available Sheet erosion, one of the most important types of water erosion, takes place on the top soil as tiny soil layer movement that affects lake and stream ecosystem. This type of erosion is very important because the productive soil layer on the top soil can be lost in a very short period of time. The goal of this study was to quantify the amount of surface (sheet and rill soil erosion, and to identify areas under high erosion risk within the study area at Yusufeli province in Artvin by using RUSLE erosion methodology. As a result of the study it was found that the average annual potential soil loss by surface erosion was 3.6 ton ha-1yr-1. Additionally, the maps produced and conclusions reached by the study revealed that the areas of high erosion risk were identified spatially and measures to control erosion on some of these high risk areas can be possible with appropriate erosion control techniques.

Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

International Nuclear Information System (INIS)

Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

2016-01-01

Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

Energy Technology Data Exchange (ETDEWEB)

Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

2016-11-01

Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

High-Z material erosion and its control in DIII-D carbon divertor

Directory of Open Access Journals (Sweden)

R. Ding

2017-08-01

Full Text Available As High-Z materials will likely be used as plasma-facing components (PFCs in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH4 injection also provides information on radial transport due to E ×B drifts and cross field diffusion. Finally, D2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.

Design study of blanket structure for tokamak experimental fusion reactor

International Nuclear Information System (INIS)

1979-11-01

Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

ITER breeding blanket module design and analysis

Energy Technology Data Exchange (ETDEWEB)

Kuroda, Toshimasa; Enoeda, Mikio; Kikuchi, Shigeto [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1998-11-01

The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

Remote handling of the blanket segments: Testing of 1/3 scale mock-ups on the ROBERTINO facility

International Nuclear Information System (INIS)

Maisonnier, D.; Amelotti, F.; Chiasera, A.

1994-01-01

The remotized replacement of the blanket segments inside the Vacuum Vessel of a fusion reactor is one of the critical tasks for reactor components design, operational procedures, and safety. This open-quotes hostile environmentclose quotes task must be accomplished by a specific Blanket Handling Device, with a grasping device acting as open-quotes end-effectorclose quotes, because of intervention complexity, of components dimensions and weights, and of consequences of possible accidents during the blanket segments handling operations. Therefore, specific support experimental studies in this field appear to be necessary in order to: select appropriate blanket handling devices and procedures; assess the design of all components involved in the handling operations; perform checks in all field related to the robotized handling control (kinematics and dynamics of the grasping device trajectory planning and motion control, sensing and intelligence of the blanket handling devices, etc.); improve reliability and safety for the replacement sequences; give a realistic estimation of the time duration of the replacement duration. During the test phase, handling operations were carried out on the blanket mock-ups by means of different gripping devices. The operations were driven in the control room by means of the Motion command computer and the real time sensing data display allowed operations' control. The results were analyzed by charting the sensors' data

Wind erosion control with scattered vegetation in the Sahelian zone of Burkina Faso

NARCIS (Netherlands)

Leenders, J.K.

2006-01-01

The Sahelian zone ofAfricais the region that is globally most subjected to land degradation, with wind erosion being the most important soil degradation process. By using control measures, the negative effects of wind erosion can be reduced. At present, adoption of

APT target-blanket fabrication development

Energy Technology Data Exchange (ETDEWEB)

Fisher, D.L.

1997-06-13

Concepts for producing tritium in an accelerator were translated into hardware for engineering studies of tritium generation, heat transfer, and effects of proton-neutron flux on materials. Small-scale target- blanket assemblies were fabricated and material samples prepared for these performance tests. Blanket assemblies utilize composite aluminum-lead modules, the two primary materials of the blanket. Several approaches are being investigated to produce large-scale assemblies, developing fabrication and assembly methods for their commercial manufacture. Small-scale target-blanket assemblies, designed and fabricated at the Savannah River Site, were place in Los Alamos Neutron Science Center (LANSCE) for irradiation. They were subjected to neutron flux for nine months during 1996-97. Coincident with this test was the development of production methods for large- scale modules. Increasing module size presented challenges that required new methods to be developed for fabrication and assembly. After development, these methods were demonstrated by fabricating and assembling two production-scale modules.

Proposal for the award of a blanket order contract for the supply of microprocessor-based protection and control devices for the CERN HV distribution network

CERN Document Server

2004-01-01

This document concerns the award of a blanket contract for the supply of microprocessor-based protection and control devices for the CERN HV distribution network. The Finance Committee is invited to agree to the negotiation of a blanket order contract with SCHNEIDER ELECTRIC (PT), the lowest technically acceptable bidder after realignment, for the supply of microprocessor-based protection and control devices for the CERN HV distribution network for a total amount of 1 900 000 euros (2 924 128 Swiss francs), subject to revision for inflation after 1 January 2007. The rate of exchange used is that stipulated in the tender

On blanket concepts of the Helias reactor

International Nuclear Information System (INIS)

Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

1999-07-01

The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

Preliminary Analysis for K-DEMO Water Cooled Breeding Blanket Using MARS-KS

International Nuclear Information System (INIS)

Lee, Jeong-Hun; Kim, Geon-Woo; Park, Goon-Cherl; Cho, Hyoung-Kyu; Im, Kihak

2014-01-01

In the present study, thermal-hydraulic analyses for the blanket concept are being conducted using the Multidimensional Analysis of Reactor Safety (MARSKS) code, which has been used for the safety analysis of a pressurized water reactor. The purposes of the analyses are to verify the applicability of the code for the proposed blanket system, to investigate the departure of nucleate boiling (DNB) occurrence during the normal and transient conditions, and to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. In this paper, the thermal analysis results of the proposed blanket design using the MARS-KS code are presented for the normal operation and an accident condition of a reduced coolant flow rate. Afterwards, the plan for the whole blanket system analysis using MARSKS is introduced and the result of the first trial for the multiple blanket module analysis is summarized. In the present study, thermal-hydraulic analyses for the blanket concept were conducted using the MARS-KS code for a single blanket module. By comparing the MARS calculation results with the CFD analysis results, it was found that MARS-KS can be applied for the blanket thermal analysis with less number of computational meshes. Moreover, due to its capability on the two-phase flow analysis, it can be used for the transient or accident simulation where a phase change may be resulted in. In the future, the MARS-KS code will be applied for the anticipated transient and design based accident analyses. The investigation of the DNB occurrence during the normal and transient conditions will be of special interest of the analysis using it. After that, a methodology to simulate the entire blanket system was proposed by using the DLL version of MARS-KS. A supervisor program, which controls the multiple DLL files, was developed for the common header modelling. The program explicitly determines the flow rates of each module which can equalize

Magnetoconvection in HCLL blankets

International Nuclear Information System (INIS)

Mistrangelo, C.; Buehler, L.

2014-01-01

In the present work we consider magneto-convective flows in one of the proposed European liquid metal blankets that will be tested in the experimental fusion reactor ITER. Here the PbLi alloy is used as breeder material and helium as coolant. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma and in case of non-uniform thermal conditions. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating and cooling of walls, since only a weak forced ow is foreseen for tritium extraction in external ancillary systems. Buoyancy can therefore become very important and modify the velocity distribution and related heat transfer performance of the blanket. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. According to the last design review two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. Velocity and temperature distributions are discussed for various volumetric heat sources and magnetic field strengths.

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.E.; Cheng, E.T.

1985-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

Native Roadside Vegetation that Enhances Soil Erosion Control in Boreal Scandinavia

Directory of Open Access Journals (Sweden)

Annika K. Jägerbrand

2014-07-01

Full Text Available This study focused on identifying vegetation characteristics associated with erosion control at nine roadside sites in mid-West Sweden. A number of vegetation characteristics such as cover, diversity, plant functional type, biomass and plant community structure were included. Significant difference in cover between eroded and non-eroded sub-sites was found in evergreen shrubs, total cover, and total above ground biomass. Thus, our results support the use of shrubs in order to stabilize vegetation and minimize erosion along roadsides. However, shrubs are disfavored by several natural and human imposed factors. This could have several impacts on the long-term management of roadsides in boreal regions. By both choosing and applying active management that supports native evergreen shrubs in boreal regions, several positive effects could be achieved along roadsides, such as lower erosion rate and secured long-term vegetation cover. This could also lead to lower costs for roadside maintenance as lower erosion rates would require less frequent stabilizing treatments and mowing could be kept to a minimum in order not to disfavor shrubs.

Stress analysis of the tokamak engineering test breeder blanket

International Nuclear Information System (INIS)

Huang Zhongqi

1992-01-01

The design features of the hybrid reactor blanket and main parameters are presented. The stress analysis is performed by using computer codes SAP5p and SAP6 with the three kinds of blanket module loadings, i.e, the pressure of coolant, the blanket weight and the thermal loading. Numerical calculation results indicate that the stresses of the blanket are smaller than the allowable ones of the material, the blanket design is therefore reasonable

Remote handling demonstration of ITER blanket module replacement

International Nuclear Information System (INIS)

Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

2001-01-01

In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R and D program, a rail-mounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R and D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance. (author)

Preliminary study on lithium-salt aqueous solution blanket

International Nuclear Information System (INIS)

Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

1992-06-01

Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

Nuclear characteristics of D-D fusion reactor blankets

International Nuclear Information System (INIS)

Nakashima, Hideki; Ohta, Masao

1978-01-01

Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

A field experiment on the controls of sediment transport on bedrock erosion

Science.gov (United States)

Beer, A. R.; Turowski, J. M.; Fritschi, B.; Rieke-Zapp, D.; Campana, L.; Lavé, J.

2012-12-01

The earth`s surface is naturally shaped by interactions of physical and chemical processes. In mountainous regions with steep topography river incision fundamentally controls the geomorphic evolution of the whole landscape. There, erosion of exposed bedrock sections by fluvial sediment transport is an important mechanism forming mountain river channels. The links between bedload transport and bedrock erosion has been firmly established using laboratory experiments. However, there are only few field datasets linking discharge, sediment transport, impact energy and erosion that can be used for process understanding and model evaluation. To fill this gap, a new measuring setup has been commissioned to raise an appropriate simultaneous dataset of hydraulics, sediment transport and bedrock erosion at high temporal and spatial resolution. Two natural stone slabs were installed flush with the streambed of the Erlenbach, a gauged stream in the Swiss Pre-Alps. They are mounted upon force sensors recording vertical pressure und downstream shear caused by passing sediment particles. The sediment transport rates can be assessed using geophone plates and an automated moving basket system taking short-term sediment samples. These devices are located directly downstream of the stone slabs. Bedrock erosion rates are measured continuously with erosion sensors at sub-millimeter accuracy at three points on each slab. In addition, the whole slab topography is surveyed with photogrammetry and a structured-light 3D scanner after individual flood events. Since the installation in 2011, slab bedrock erosion has been observed during several transport events. We discuss the relation between hydraulics, bedload transport, resulting pressure forces on the stone slabs and erosion rates. The aim of the study is the derivation of an empirical process law for fluvial bedrock erosion driven by moving sediment particles.

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.L.; Cheng, E.T.

1986-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

Tritium inventory and permeation in liquid breeder blankets

International Nuclear Information System (INIS)

Reiter, F.

1990-01-01

This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

Protection from erosion following wildfire

Science.gov (United States)

Peter R. Robichaud; William J. Elliot

2006-01-01

Erosion in the first year after a wildfire can be up to three orders of magnitude greater than the erosion from undisturbed forests. To mitigate potential postfire erosion, various erosion control treatments are applied on highly erodible areas with downstream resources in need of protection. Because postfire erosion rates generally decline by an order of magnitude for...

Measurement of erosion: Is it possible?

NARCIS (Netherlands)

Stroosnijder, L.

2005-01-01

Reasons for erosion measurements are: (1) to determine the environmental impact of erosion and conservation practices, (2) scientific erosion research; (3) development and evaluation of erosion control technology; (4) development of erosion prediction technology and (5) allocation of conservation

Design and development of ceramic breeder demo blanket

International Nuclear Information System (INIS)

Enoeda, M.; Sato, S.; Hatano, T.

2001-01-01

Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

Conceptual design of Blanket Remote Handling System for CFETR

Energy Technology Data Exchange (ETDEWEB)

Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

2015-11-15

Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

Conceptual design of Blanket Remote Handling System for CFETR

International Nuclear Information System (INIS)

Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

2015-01-01

Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

Economic evaluation of the Blanket Comparison and Selection Study

International Nuclear Information System (INIS)

Waganer, L.M.

1985-01-01

The economic impact of employing the highly ranked blankets in the Blanket Comparison and Selection Study (BCSS) was evaluated in the context of both a tokamak and a tandem mirror power reactor (TMR). The economic evaluation criterion was determined to be the cost of electricity. The influencing factors that were considered are the direct cost of the blankets and related systems; the annual cost of blanket replacement; and the performance of the blanket, heat transfer, and energy conversion systems. The technical and cost bases for comparison were those of the STARFIRE and Mirror Advanced Reactor Study conceptual design power plants. The economic evaluation results indicated that the nitrate-salt-cooled blanket concept is an economically attractive concept for either reactor type. The water-cooled, solid breeder blanket is attractive for the tokamak and somewhat less attractive for the TMR. The helium-cooled, liquidlithium breeder blanket is the least economically desirable of higher ranked concepts. The remaining self-cooled liquid-metal and the helium-cooled blanket concepts represent moderately attractive concepts from an economic standpoint. These results are not in concert with those found in the other BCSS evaluation areas (engineering feasibility, safety, and research and development (R and D) requirements). The blankets faring well economically had generally lower cost components, lower pumping power requirements, and good power production capability. On the other hand, helium- and lithium-cooled systems were preferred from the standpoints of safety, engineering feasibility, and R and D requirements

Evaluation of the serum zinc level in erosive and non-erosive oral lichen planus.

Science.gov (United States)

Gholizadeh, N; Mehdipour, M; Najafi, Sh; Bahramian, A; Garjani, Sh; Khoeini Poorfar, H

2014-06-01

Lichen planus is a chronic inflammatory immunologic-based disease involving skin and mucosa. This disease is generally divided into two categories: erosive and non-erosive. Many etiologic factors are deliberated regarding the disease; however, the disorders of immune system and the role of cytotoxic T-lymphocytes and monocytes are more highlighted. Zinc is an imperative element for the growth of epithelium and its deficiency induces the cytotoxic activity of T-helper2 cells, which seems to be associated with lichen planus. This study was aimed to evaluate the levels of serum zinc in erosive and non-erosive oral lichen planus (OLP) and to compare it with the healthy control group to find out any feasible inference. A total of 22 patients with erosive oral lichen planus, 22 patients with non erosive OLP and 44 healthy individuals as the control group were recruited in this descriptive-comparative study. All the participants were selected from the referees to the department of oral medicine, school of dentistry, Tabriz University of Medical Sciences. Serum zinc level was examined for all the individuals with liquid-stat kit (Beckman Instruments Inc.; Carlsbad, CA). Data were analyzed by adopting the ANOVA and Tukey tests, using SPSS 16 statistical software. The mean age of patients with erosive and non-erosive LP was 41.7 and 41.3 years, respectively. The mean age of the healthy control group was 34.4 years .The mean serum zinc levels in the erosive and non erosive lichen planus groups and control groups were 8.3 (1.15), 11.15 (0.92) and 15.74 (1.75) μg/dl respectively. The difference was statistically significant (poral lichen planus. This finding may probably indicate the promising role of zinc in development of oral lichen planus.

Crater Mound Formation by Wind Erosion on Mars

Science.gov (United States)

Steele, L. J.; Kite, E. S.; Michaels, T. I.

2018-01-01

Most of Mars' ancient sedimentary rocks by volume are in wind-eroded sedimentary mounds within impact craters and canyons, but the connections between mound form and wind erosion are unclear. We perform mesoscale simulations of different crater and mound morphologies to understand the formation of sedimentary mounds. As crater depth increases, slope winds produce increased erosion near the base of the crater wall, forming mounds. Peak erosion rates occur when the crater depth is ˜2 km. Mound evolution depends on the size of the host crater. In smaller craters mounds preferentially erode at the top, becoming more squat, while in larger craters mounds become steeper sided. This agrees with observations where smaller craters tend to have proportionally shorter mounds and larger craters have mounds encircled by moats. If a large-scale sedimentary layer blankets a crater, then as the layer recedes across the crater it will erode more toward the edges of the crater, resulting in a crescent-shaped moat. When a 160 km diameter mound-hosting crater is subject to a prevailing wind, the surface wind stress is stronger on the leeward side than on the windward side. This results in the center of the mound appearing to "march upwind" over time and forming a "bat-wing" shape, as is observed for Mount Sharp in Gale crater.

Cleveland Dam East Abutment : seepage control project

Energy Technology Data Exchange (ETDEWEB)

Huber, F.; Siu, D. [Greater Vancouver Regional District, Burnaby, BC (Canada); Ahlfield, S.; Singh, N. [Klohn Crippen Consultants Ltd., Vancouver, BC (Canada)

2004-09-01

North Vancouver's 91 meter high Cleveland Dam was built in the 1950s in a deep bedrock canyon to provide a reservoir for potable water to 18 municipalities. Flow in the concrete gravity dam is controlled by a gated spillway, 2 mid-level outlets and intakes and 2 low-level outlets. This paper describes the seepage control measures that were taken at the time of construction as well as the additional measures that were taken post construction to control piezometric levels, seepage and piping and slope instability in the East Abutment. At the time of construction, a till blanket was used to cover the upstream reservoir slope for 200 meters upstream of the dam. A single line grout curtain was used through the overburden from ground surface to bedrock for a distance of 166 meters from the dam to the East Abutment. Since construction, the safety of the dam has been compromised through changes in piezometric pressure, seepage and soil loss. Klohn Crippen Consultants designed a unique seepage control measure to address the instability risk. The project involved excavating 300,000 cubic meters of soil to form a stable slope and construction bench. A vertical wall was constructed to block seepage. The existing seepage control blanket was also extended by 260 meters. The social, environmental and technical issues that were encountered during the rehabilitation project are also discussed. The blanket extension construction has met design requirements and the abutment materials that are most susceptible to internal erosion have been covered by non-erodible blanket materials such as plastic and roller-compacted concrete (RCC). The project was completed on schedule and within budget and has greatly improved the long-term stability of the dam and public safety. 2 refs., 8 figs.

Erosion and erosion-corrosion

International Nuclear Information System (INIS)

Isomoto, Yoshinori

2008-01-01

It is very difficult to interpret the technical term of erosion-corrosion' which is sometimes encountered in piping systems of power plants, because of complicated mechanisms and several confusing definitions of erosion-corrosion phenomena. 'FAC (flow accelerated corrosion)' is recently introduced as wall thinning of materials in power plant systems, as a representative of 'erosion-corrosion'. FAC is, however, not necessarily well understood and compared with erosion-corrosion. This paper describes firstly the origin, definition and fundamental understandings of erosion and erosion-corrosion, in order to reconsider and reconfirm the phenomena of erosion, erosion-corrosion and FAC. Next, typical mapping of erosion, corrosion, erosion-corrosion and FAC are introduced in flow velocity and environmental corrosiveness axes. The concept of damage rate in erosion-corrosion is finally discussed, connecting dissolution rate, mass transfer of metal ions in a metal oxide film and film growth. (author)

Fusion blankets for high efficiency power cycles

International Nuclear Information System (INIS)

Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Usher, J.L.

1980-04-01

Definitions are given of 10 generic blanket types and the specific blanket chosen to be analyzed in detail from each of the 10 types. Dimensions, compositions, energy depositions and breeding ratios (where applicable) are presented for each of the 10 designs. Ultimately, based largely on neutronics and thermal hyraulics results, breeding an nonbreeding blanket options are selected for further design analysis and integration with a suitable power conversion subsystem

Electromagnetic analysis of ITER shield blanket under VDE

International Nuclear Information System (INIS)

Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

2010-01-01

Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

Neutronic design for the TFTR lithium blanket module

International Nuclear Information System (INIS)

Cheng, E.T.; Engholm, B.A.; Su, S.D.

1981-01-01

The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

LMFBR Blanket Physics Project progress report No. 2

International Nuclear Information System (INIS)

Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

1971-01-01

This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

Liquid metal cooled blanket concept for NET

International Nuclear Information System (INIS)

Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

1986-01-01

A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

Beryllium R&D for blanket application

Science.gov (United States)

Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

1998-10-01

The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

Heat transfer problems in gas-cooled solid blankets

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1976-01-01

In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed

Investigation of aqueous slurries as fusion reactor blankets

International Nuclear Information System (INIS)

Schuller, M.J.

1985-01-01

Numerical and experimental studies were carried out to assess the feasibility of using an aqueous slurry, with lithium in its solid component, to meet the tritium breeding, cooling, and shielding requirements of a controlled thermonuclear reactor (CTR). The numerical studies were designed to demonstrate the theoretical ability of a conceptual slurry blanket to breed adequate tritium to sustain the CTR. The experimental studies were designed to show that the tritium retention characteristics of likely solid components for the slurry were conducive to adequate tritium recovery without the need for isotopic separation. The numerical portion of this work consisted in part of using ANISN, a one-dimensional finite difference neutron transport code, to model the neutronic performance of the slurry blanket concept. The parameters governing tritium production and retention in a slurry were computed and used to modify the results of the ANISN computer runs. The numerical work demonstrated that the slurry blanket was only marginally capable of breeding sufficient tritium without the aid of a neutron multiplying region. The experimental portion of this work consisted of several neutron irradiation experiments, which were designed to determine the retention abilities of LiF particles

Changes in the hydrological status of the basin due to the application of erosion control works

Directory of Open Access Journals (Sweden)

Radonjić Jasmina

2016-01-01

Full Text Available Protection of land with vegetation is the primary factor in the fight against water erosion with necessary application of biotechnical, technical, administrative and planning measures. One of the first basins to be treated with works for the protection against erosion and torrent control is the Gradasnica River basin. The basic parameters to display the changes of the hydrological status of the land are the state of erosion, the change of erosion-coefficient, annual sediment yield, specific annual sediment discharge through the hydrographic network, the value of the runoff curve number and value of the maximal discharge. Works on protection from erosion and regulations of torrents have influenced the decrease in erosion coefficient values from strong erosion (Z=0.99 to the value of weak erosion (Z=0.40, as well as the reduction of the maximum discharge value from Qmax(1956=108,12m3/s to the value of Qmax(2014=87.2 m3/s.

Testing the control of mineral supply rates on chemical erosion in the Klamath Mountains

Science.gov (United States)

West, N.; Ferrier, K.

2017-12-01

The relationship between rates of chemical erosion and mineral supply is central to many problems in Earth science, including the role of tectonics in the global carbon cycle, nutrient supply to soils and streams via soil production, and lithologic controls on landscape evolution. We aim to test the relationship between mineral supply rates and chemical erosion in the forested uplands of the Klamath mountains, along a latitudinal transect of granodioritic plutons that spans an expected gradient in mineral supply rates associated with the geodynamic response to the migration of the Mendocino Triple Junction. We present 10Be-derived erosion rates and Zr-derived chemical depletion factors, as well as bulk soil and rock geochemistry on 10 ridgetops along the transect to test hypotheses about supply-limited and kinetically-limited chemical erosion. Previous studies in this area, comparing basin-averaged erosion rates and modeled uplift rates, suggest this region may be adjusted to an approximate steady state. Our preliminary results suggest that chemical erosion at these sites is influenced by both mineral supply rates and dissolution kinetics.

Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

International Nuclear Information System (INIS)

Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

1979-09-01

The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO 2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li 2 O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

Reducing beryllium content in mixed bed solid-type breeder blankets

Energy Technology Data Exchange (ETDEWEB)

Shimwell, J., E-mail: mail@jshimwell.com [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Lilley, S.; Morgan, L.; Packer, L.; Kovari, M.; Zheng, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); McMillan, J. [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom)

2016-11-01

Highlights: • The ratio of breeder ceramic to neutron multiplier of breeder blankets was varied linearly with depth. • Blankets with varying composition were found to perform better than uniform composition breeder blankets. • It was also possible to reduce the amount of beryllium required by the blanket. - Abstract: Beryllium (Be) is a precious resource with many high value uses, the low energy threshold (n,2n) reaction makes Be an excellent neutron multiplier for use in fusion breeder blankets. Estimates of Be requirements and available resources suggest that this could represent a major supply difficulty for solid-type blanket concepts. Reducing the quantity of Be required by breeder blankets would help to alleviate the problem to some extent. In addition, it is important that the reduction in the Be quantity does not diminish the blanket's performance in key aspects such as the tritium breeding ratio (TBR), energy multiplication and peak nuclear heating. Mixed pebble bed designs allow for the multiplier fraction to be varied throughout the blanket. This neutronics study used MCNP 6 to investigate linear variations of the multiplier fraction in relation to blanket depth, in order to better utilise the important multiplying Be(n,2n) and breeding reactions. Blankets with a uniform multiplier fraction showed little scope for reduction in Be mass. Blankets with varying multiplier fractions were able to simultaneously use 10% less Be, increase the energy amplification by 1%, reduce the peak heating by 7% and maintaining a sufficient TBR when compared to the performance achievable using a uniform composition.

Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

International Nuclear Information System (INIS)

Norajitra, P.

1995-06-01

Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li 4 SiO 4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.) [de

MIT LMFBR blanket research project. Final summary report

International Nuclear Information System (INIS)

Driscoll, M.J.

1983-08-01

This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

Manufacturing aspects in the design of the breeder unit for Helium Cooled Pebble Bed blankets

International Nuclear Information System (INIS)

Rey, J.; Ihli, T.; Filsinger, D.; Polixa, C.

2007-01-01

The breeding blanket programme has been in the focus of European fusion research for more than a decade. Recently, it has been driven by the EU Power Plant Conceptual Study (PPCS), investigating the potential of fusion energy in a future economic environment. On the way to the first commercial nuclear fusion reactor (DEMO) new studies for reactor in-vessel components have been initiated. One central focus is the design and manufacturing of the blankets that have to ensure the breeding process to maintain the fuel cycle and are also responsible for the extraction of the main part of the reactor heat for power generation. Two kinds are established: One is the Helium Cooled Pebble Bed (HCPB) and the other the Helium Cooled Liquid Lead (HCLL) blanket. Both designs employ three different cooling plate assemblies. The outer, cooled U-shaped shell, namely the First Wall (FW), with two caps builds the blanket box. The structural strength of the blanket box is realized by integrating Stiffening Grids (SG) that separate the equally spaced Breeder Unit (BU) and allow the box, in case of faulted conditions, to withstand an internal pressure of 8 MPa. The cooled SG constitute the side walls of the BU and are also cooled. The BU consists of a dedicated Cooling Plate (CP) assembly. In present studies about the fabrication of Cooling Plates two kinds of diffusion welding processes are focused on. One is based on a Hot Isostatic Gas Process (HIP). The second is a uni-axial Diffusion Welding Process (DWP). In both cases the bond between the two halves of the cooling plate structure is reached by controlled pressure and heat cycles. Approaching larger, realistic scaled components the uncertainty of ensuring uniform process parameters across the bonding zone increases the risk of defect sources and, therefore, makes it difficult to guarantee the required bonding penetration. This study presents an alternative manufacturing strategy. The premises for this strategy are the reduction of

Composition of enamel pellicle from dental erosion patients.

Science.gov (United States)

Carpenter, G; Cotroneo, E; Moazzez, R; Rojas-Serrano, M; Donaldson, N; Austin, R; Zaidel, L; Bartlett, D; Proctor, G

2014-01-01

Oral health is dependent upon a thin mobile film of saliva on soft and hard tissues. Salivary proteins adhere to teeth to form the acquired enamel pellicle which is believed to protect teeth from acid erosion. This study investigated whether patients suffering diet-induced dental erosion had altered enamel pellicles. Thirty patients suffering erosion were compared to healthy age-matched controls. Subjects wore a maxillary splint holding hydroxyapatite and human enamel blocks for 1 h. The acquired enamel pellicle was removed from the blocks and compared to the natural incisor pellicle. Basic Erosive Wear Examination scores confirmed that dental erosion was present in erosion patients and absent from healthy age-matched controls. Erosion patients had half the amount of proteins (BCA assay) within the acquired pellicle forming on splint blocks compared to normal controls (p erosion patients (p erosion patients and healthy controls. In summary, the formation of new acquired pellicles on surfaces was reduced in erosion patients, which may explain their greater susceptibility to acid erosion of teeth. © 2014 S. Karger AG, Basel.

Thermomechanical analysis of the DFLL test blanket module for ITER

International Nuclear Information System (INIS)

Chen Hongli; Wu Yican; Bai Yunqing

2006-01-01

The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

RAMI analysis for DEMO HCPB blanket concept cooling system

Energy Technology Data Exchange (ETDEWEB)

Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

2015-10-15

Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

Autonomous watersheds: Reducing flooding and stream erosion through real-time control

Science.gov (United States)

Kerkez, B.; Wong, B. P.

2017-12-01

We introduce an analytical toolchain, based on dynamical system theory and feedback control, to determine how many control points (valves, gates, pumps, etc.) are needed to transform urban watersheds from static to adaptive. Advances and distributed sensing and control stand to fundamentally change how we manage urban watersheds. In lieu of new and costly infrastructure, the real-time control of stormwater systems will reduce flooding, mitigate stream erosion, and improve the treatment of polluted runoff. We discuss the how open source technologies, in the form of wireless sensor nodes and remotely-controllable valves (open-storm.org), have been deployed to build "smart" stormwater systems in the Midwestern US. Unlike "static" infrastructure, which cannot readily adapt to changing inputs and land uses, these distributed control assets allow entire watersheds to be reconfigured on a storm-by-storm basis. Our results show how the control of even just a few valves within urban catchments (1-10km^2) allows for the real-time "shaping" of hydrographs, which reduces downstream erosion and flooding. We also introduce an equivalence framework that can be used by decision-makers to objectively compare investments into "smart" system to more traditional solutions, such as gray and green stormwater infrastructure.

Erosion control and protection from torrential floods in Serbia-spatial aspects

Directory of Open Access Journals (Sweden)

Ristić Ratko

2011-01-01

Full Text Available Torrential floods represent the most frequent phenomenon within the category of “natural risks” in Serbia. The representative examples are the torrential floods on the experimental watersheds of the rivers Manastirica (June 1996 and Kamišna (May 2007. Hystorical maximal discharges (Qmaxh were reconstructed by use of ″hydraulics flood traces″ method. Computations of maximal discharges (Qmaxc, under hydrological conditions after the restoration of the watersheds, were performed by use of a synthetic unit hydrograph theory and Soil Conservation Service methodology. Area sediment yields and intensity of erosion processes were estimated on the basis of the “Erosion Potential Method”. The actual state of erosion processes is represented by the coefficients of erosion Z=0.475 (Manastirica and Z=0.470 (Kamišna. Restoration works have been planned with a view to decreasing yields of erosive material, increasing water infiltration capacity and reducing flood runoff. The planned state of erosion processes is represented by the coefficients of erosion Z=0.343 (Manastirica and Z=0.385 (Kamišna. The effects of hydrological changes were estimated by the comparison of historical maximal discharges and computed maximal discharges (under the conditions after the planned restoration. The realisation of restoration works will help decrease annual yields of erosive material from Wа=24357 m3 to Wа=16198.0 m3 (Manastirica and from Wа=19974 m3 to Wа=14434 m3 (Kamišna. The values of historical maximal discharges (QmaxhMan=154.9 m3•s-1; QmaxhKam=76.3 m3•s-1 were significantly decreased after the restoration (QmaxcMan=84.5 m3 •s-1; QmaxcKam=43.7 m3•s-1, indicating the improvement of hydrological conditions, as a direct consequence of erosion and torrent control works. Integrated management involves biotechnical works on the watershed, technical works on the hydrographic network within a precisely defined administrative and spatial framework in

Proposal for the award of a blanket contract for the supply, installation and maintenance of the LHC access control system

CERN Document Server

2004-01-01

This document concerns the award of a blanket contract for the supply, installation and maintenance of the LHC access control system. Following a market survey carried out among 134 firms in fifteen Member States, a call for tenders (IT-3026/TS/LHC) was sent on 22 January 2004 to eight firms and eight consortia in six Member States. By the closing date, CERN had received nine tenders from two firms and seven consortia in five Member States. The Finance Committee is invited to agree to the negotiation of a blanket contract with the consortium CEGELEC CENTRE EST (FR) - CEGELEC (NL), the lowest technically compliant bidder, for the supply, installation and maintenance of the LHC access control system for a total amount not exceeding 4 600 000 euros (7 141 000 Swiss francs), subject to revision for inflation from 1 January 2007. The rate of exchange used is that stipulated in the tender. The firm has indicated the following distribution by country of the contract value covered by this adjudication proposal: FR - ...

Blanket options for high-efficiency fusion power

International Nuclear Information System (INIS)

Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

1980-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

Fusion blankets for high-efficiency power cycles

International Nuclear Information System (INIS)

Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

1980-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

Fusion blanket for high-efficiency power cycles

International Nuclear Information System (INIS)

Usher, J.L.; Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Taussig, R.

1980-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by Ar) utilizing Li 2 O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

Fusion blankets for high-efficiency power cycles

International Nuclear Information System (INIS)

Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

1981-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 deg C) of conventional structural materials such as stainless steels. In this project 'two-zone' blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 deg C leading to an overall efficiency estimate of 55 to 60% for this reference case. (author)

Methods to enhance blanket power density

International Nuclear Information System (INIS)

Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

1982-06-01

The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

Neutronic optimization of solid breeder blankets for STARFIRE design

International Nuclear Information System (INIS)

Gohar, Y.; Abdou, M.A.

1980-01-01

Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

Development of advanced blanket materials for solid breeder blanket of fusion reactor

International Nuclear Information System (INIS)

Ishitsuka, E.

2002-01-01

Advanced solid breeding blanket design in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high dose of neutron irradiation. Therefore, the development of such advanced blanket materials is indispensable. In this paper, the cooperation activities among JAERI, universities and industries in Japan on the development of these advanced materials are reported. Advanced tritium breeding material to prevent the grain growth in high temperature had to be developed because the tritium release behavior degraded by the grain growth. As one of such materials, TiO 2 -doped Li 2 TiO 3 has been studied, and TiO 2 -doped Li 2 TiO 3 pebbles was successfully fabricated. For the advanced neutron multiplier, the beryllium intermetallic compounds that have high melting point and good chemical stability have been studied. Some characterization of Be 12 Ti was studied. The pebble fabrication study for Be 12 Ti was also performed and Be 12 Ti pebbles were successfully fabricated. From these activities, the bright prospect to realize the DEMO blanket by the application of TiO 2 -doped Li 2 TiO 3 and beryllium intermetallic compounds was obtained. (author)

Review: BNL Tokamak graphite blanket design concepts

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1976-01-01

The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

Vibration damage testing of thermal barrier fibrous blanket insulation

International Nuclear Information System (INIS)

Black, W.E.; Betts, W.S.

1984-01-01

GA Technologies is engaged in a long-term, multiphase program to determine the vibration characteristics of thermal barrier components leading to qualification of assemblies for High Temperature Gas-Cooled Reactor (HTGR) service. The phase of primary emphasis described herein is the third in a series of acoustic tests and uses as background the more elemental tests preceding it. Two sizes of thermal barrier coverplates with one fibrous blanket insulation type were tested in an acoustic environment at sound pressure levels up to 160 dB. Three tests were conducted using sinusoidal and random noise for up to 200 h duration at room temperature. Frequent inspections were made to determine the progression of degradation using definition of stages from a prior test program. Initially the insulation surface adjacent to the metallic seal sheets (noise side) assumed a chafed or polished appearance. This was followed by flattening of the as-received pillowed surface. This stage was followed by a depression being formed in the vicinity of the free edge of the coverplate. Next, loose powder from within the blanket and from fiber erosion accumulated in the depression. Prior experience showed that the next stage of deterioration exhibited a consolidation of the powder to form a local crust. In this test series, this last stage generally failed to materialize. Instead, surface holes generated by solid ceramic particulates (commonly referred to as 'shot') constituted the stage following powdering. With the exception of some manufacturing-induced anomalies, the final stage, namely, gross fiber breakup, did not occur. It is this last stage that must be prevented for the thermal barrier to maintain its integrity. (orig./GL)

Environmental considerations for alternative fusion reactor blankets

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Young, J.R.

1975-01-01

Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

A Li-particulate blanket concept for ITER

International Nuclear Information System (INIS)

Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

1989-01-01

The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

Neutron dosimetry for the TFTR Lithium-Blanket-Module program

International Nuclear Information System (INIS)

Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

1981-01-01

The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

International Nuclear Information System (INIS)

Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

1980-09-01

A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

Energy Technology Data Exchange (ETDEWEB)

Someya, Y.; Tobita, K.; Utoh, H.; Hoshino, K.; Asakura, N.; Nakamura, M.; Tanigawa, H.; Mikio, E.; Tanigawa, H.; Nakamichi, M.; Hoshino, T., E-mail: someya.yoji@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

2012-09-15

Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

Imploding-liner reactor nucleonic studies: the LINUS blanket

International Nuclear Information System (INIS)

Dudziak, D.J.

1977-09-01

Scoping nucleonic studies have been performed for a small imploding-liner fusion reactor concept. Tritium breeding ratio and time-dependent energy deposition rates were the primary parameters of interest in the study. Alloys of Pb and LiPb were considered for the liquid liner (blanket), and tritium breeding was found to be more than adequate with blankets less than 1 m thick. However, neutron leakages into the solid cylinder block surrounding the liquid liner are generally quite high, so considerable effort was concentrated on minimizing these values. Time-dependent calculations reveal that 89% of the energy is deposited in the blanket within 2 μs. Thus, LINUS's blanket should remain intact for the requisite neutron and gamma-ray lifetimes

Radiolysis and corrosion aspects of the aqueous self-cooled blanket concept

International Nuclear Information System (INIS)

Bruggeman, A.; Snykers, M.; Bogaerts, W.F.; Waeben, R.; Embrechts, M.J.; Steiner, D.

1989-01-01

Corrosion and radiolysis aspects of the Aqueous Self-Cooled Blanket concept, proposed as a potential shielding breeding blanket for near term fusion devices and fusion reactors, have been investigated. On the basis of preliminary results for selected aqueous solutions of lithium compounds, no particular corrosion problems have been revealed for the low-temperature concept envisaged for NET and radiolysis effects might be controlled by appropriate countermeasures. For the reactor-relevant high-temperature concept particular attention has to be paid to intergranular stress-corrosion and to the synergistic radiolysis-corrosion effects. Further information is needed from tests performed in relevant operational conditions. (orig.)

Agriculture and stream water quality: A biological evaluation of erosion control practices

Science.gov (United States)

Lenat, David R.

1984-07-01

Agricultural runoff affects many streams in North Carolina. However, there is is little information about either its effect on stream biota or any potential mitigation by erosion control practices. In this study, benthic macroinvertebrates were sampled in three different geographic areas of North Carolina, comparing control watersheds with well-managed and poorly managed watersheds. Agricultural streams were characterized by lower taxa richness (especially for intolerant groups) and low stability. These effects were most evident at the poorly managed sites. Sedimentation was the apparent major problem, but some changes at agricultural sites implied water quality problems. The groups most intolerant of agricultural runoff were Ephemeroptera, Plecoptera and Trichoptera. Tolerant species were usually filter-feeders or algal grazers, suggesting a modification of the food web by addition of particulate organic matter and nutrients. This study clearly indicates that agricultural runoff can severely impact stream biota. However, this impact can be greatly mitigated by currently recommended erosion control practices.

Beryllium R and D for blanket application

Energy Technology Data Exchange (ETDEWEB)

Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-10-01

The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

Beryllium R and D for blanket application

International Nuclear Information System (INIS)

Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

1998-01-01

The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

Mechanical and thermal design of hybrid blankets

International Nuclear Information System (INIS)

Schultz, K.R.

1978-01-01

The thermal and mechanical aspects of hybrid reactor blanket design considerations are discussed. This paper is intended as a companion to that of J. D. Lee of Lawrence Livermore Laboratory on the nuclear aspects of hybrid reactor blanket design. The major design characteristics of hybrid reactor blankets are discussed with emphasis on the areas of difference between hybrid reactors and standard fusion or fission reactors. Specific examples are used to illustrate the design tradeoffs and choices that must be made in hybrid reactor design. These examples are drawn from the work on the Mirror Hybrid Reactor

Blanket safety by GEMSAFE methodology

International Nuclear Information System (INIS)

Sawada, Tetsuo; Saito, Masaki

2001-01-01

General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

INTOR first wall/blanket/shield activity

International Nuclear Information System (INIS)

Gohar, Y.; Billone, M.C.; Cha, Y.S.; Finn, P.A.; Hassanein, A.M.; Liu, Y.Y.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.

1986-01-01

The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory

ITER driver blanket, European Community design

International Nuclear Information System (INIS)

Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

1993-01-01

Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

Use of hold-gro erosion control fabric in the establishment of plant species on coal mine soil.

Science.gov (United States)

Day, A D; Ludeke, K L

1986-09-01

Experiments were conducted on the Black Mesa Coal Mine, Kayenta, Arizona in 1977 and 1978 to study the effectiveness of Hold-Gro Erosion Control Fabric (a product from the Gulf States Paper Corporation, Tuscaloosa, Alabama) in the establishment of plants on coal mine soil following the surface mining of coal. Four plant species were planted: (1) spring barley (Horduem vulgare L.), an annual grass (2) crested wheatgrass (Agropyron cristatum L.), a perennial grass (3) alfalfa (lucerne) (Medicago sativa L.), a perennial legume and (4) fourwing saltbush (Atriplex canescens Pursh.), a perennial shrub. Seeds of each plant species were planted in reclaimed coal mine soil in the spring of the year by both broadcast seeding (conventional culture) and the incorporation of seeds in Hold-Gro Erosion Control Fabric. Average numbers of seedlings established and percent ground cover for all species studied were higher in areas where conventional culture was used than they were in areas where seeds were incorporated in Hold-Gro Erosion Control Fabric. The incorporation of seeds in Hold-Gro Erosion Control Fabric in the establishment of plant species on coal mine soil was not an effective cultural practice in the southwestern United States.

Use of Hold-Gro Erosion Control Fabric in the establishment of plant species on coal mine soil

Energy Technology Data Exchange (ETDEWEB)

Day, A.D.; Ludeke, K.L.

1986-09-01

Experiments were conducted on the Black Mesa Coal Mine, Kayenta, Arizona in 1977 and 1978 to study the effectiveness of Hold-Gro Erosion Control Fabric (a product from the Gulf States Paper Corporation, Tuscaloosa, Alabama) in the establishment of plants on coal mine soil following the surface mining of coal. Four plant species were planted: spring barley (Horduem vulgare L.), an annual grass; crested wheatgrass (Agropyron cristatum L.), a perennial grass; alfalfa (lucerne) (Medicago sativa L.), a perennial legume; and fourwing saltbush (Atriplex canescens Pursh.), a perennial shrub. Seeds of each plant species were planted in reclaimed coal mine soil in the spring of the year by both broadcast seeding (conventional culture) and the incorporation of seeds in Hold-Gro Erosion Control Fabric. Average numbers of seedlings established and percent ground cover for all species studied were higher in areas where conventional culture was used than they were in areas where seeds were incorporated in Hold-Gro Erosion Control Fabric. The incorporation of seeds in Hold-Gro Erosion Control Fabric in the establishment of plant species on coal mine soil was not an effective cultural practice in the southwestern United States. 11 refs.

Summary of the target-blanket breakout group

Energy Technology Data Exchange (ETDEWEB)

Capiello, M.; Bell, C. [Los Alamos National Laboratory, NM (United States); Barthold, W.

1995-10-01

This breakout group discussed a number of topics and issues pertaining to target and blanket concepts for accelerator-driven systems. This major component area is one marked by a broad spectrum of technical approaches. It is therefore less defined than other major component areas such as the accelerator and is at an earlier stage of technical needs and task specification. The working group did reach a number of general conclusions and recommendations that are summarized. The Conference and the Target/Blanket Breakout Group provided a first opportunity for people working on a variety of missions and concepts to get together and exchange information. A number of subcritical systems applicable for a spectrum of missions were proposed at the Conference and discussed in the Breakout Group. Missions included plutonium disposition, energy production, waste destruction, isotope production, and neutron scattering. The Target/Blanket Breakout Group also defined areas where parameters and data should be addressed as target/blanket design activities become more detailed and sophisticated.

Fusion blanket testing in MFTF-α + T

International Nuclear Information System (INIS)

Kleefeldt, K.

1985-01-01

The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m 2 . Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

Pulsed activation analyses of the ITER blanket design options considered in the blanket trade-off study

International Nuclear Information System (INIS)

Wang, Q.; Henderson, D.L.

1995-01-01

Pulsed activation calculations have been performed on two blanket options considered as part of the ITER home team blanket trade-off study. The objective was to compare the activity, afterheat and waste disposal rating (WDR) results of a composite blanket-shield design for the continuous operation approximation to a pulsed operation case to determine whether the differences are at most the duty factor as predicted by the two nuclide chain model. Up to a cooling period of 100 years, the pulsed activity and afterheat values were below the continuous oepration results and well within (except for one afterheat value) the maximum deviation predicted by the two nuclide chain model. No differences in the WDR values were noted as they are, to a large extent, based on long-lived nuclides which are insensitive to short-term changes in the operation history. (orig.)

Axial blanket enrichment optimization of the NPP Krsko fuel

International Nuclear Information System (INIS)

Kromar, M.; Kurincic, B.

2001-01-01

In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

Design and analysis of ITER shield blanket

Energy Technology Data Exchange (ETDEWEB)

Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1998-12-01

This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

Processing and waste disposal needs for fusion breeder blankets system

International Nuclear Information System (INIS)

Finn, P.A.; Vogler, S.

1988-01-01

We evaluated the waste disposal and recycling requirements for two types of fusion breeder blanket (solid and liquid). The goal was to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under U.S. Nuclear Regulatory Commission regulations. Described in this paper are the radionuclides expected in fusion blanket materials, plans for reprocessing and disposal of blanket components, and estimates for the operating costs involved in waste disposal. (orig.)

Impacts of terracing on soil erosion control and crop yield in two agro-ecological zones of Rwanda

Science.gov (United States)

Rutebuka, Jules; Ryken, Nick; Uwimanzi, Aline; Nkundwakazi, Olive; Verdoodt, Ann

2017-04-01

Soil erosion remains a serious limiting factor to the agricultural production in Rwanda. Terracing has been widely adopted in many parts of the country in the past years, but its effectiveness is not yet known. Besides the standard radical (bench) terraces promoted by the government, also progressive terraces (with living hedges) become adopted mainly by the farmers. The aim of this study was to measure short-term (two consecutive rainy seasons 2016A and 2016B) run-off and soil losses for existing radical (RT) and progressive (PT) terraces versus non-protected (NP) fields using erosion plots installed in two agro-ecological zones, i.e. Buberuka highlands (site Tangata) and Eastern plateau (site Murehe) and determine their impacts on soil fertility and crop production. The erosion plot experiment started with a topsoil fertility assessment and during the experiment, maize was grown as farmer's cropping preference in the area. Runoff data were captured after each rainfall event and the collected water samples were dried to determine soil loss. Both erosion control measures reduced soil losses in Tangata, with effectiveness indices ranging from 43 to 100% when compared to the NP plots. RT showed the highest effectiveness, especially in season A. In Murehe, RT minimized runoff and soil losses in both seasons. Yet, the PT were largely inefficient, leading to soil losses exceeding those on the NP plots (ineffectiveness index of -78% and -65% in season A and B, respectively). Though topsoil fertility assessment in the erosion plots showed that the soil quality parameters were significantly higher in RT and NP plots compared to the PT plots on both sites, maize grain yield was not correlated with the physical effectiveness of the erosion control measures. Finally, the effectiveness of soil erosion control measures as well as their positive impacts on soil fertility and production differ not only by terracing type but also by agro-ecological zone and the management or

Epoxy blanket protects milled part during explosive forming

Science.gov (United States)

1966-01-01

Epoxy blanket protects chemically milled or machined sections of large, complex structural parts during explosive forming. The blanket uniformly covers all exposed surfaces and fills any voids to support and protect the entire part.

18 CFR 284.303 - OCS blanket certificates.

Science.gov (United States)

2010-04-01

... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false OCS blanket certificates. 284.303 Section 284.303 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... Pipelines on Behalf of Others § 284.303 OCS blanket certificates. Every OCS pipeline [as that term is...

In Situ analysis of CO2 laser irradiation on controlling progression of erosive lesions on dental enamel.

Science.gov (United States)

Lepri, Taísa Penazzo; Scatolin, Renata Siqueira; Colucci, Vivian; De Alexandria, Adílis Kalina; Maia, Lucianne Cople; Turssi, Cecília Pedroso; Corona, Silmara Aparecida Milori

2014-08-01

The present study aimed to evaluate in situ the effect of CO2 laser irradiation to control the progression of enamel erosive lesions. Fifty-six slabs of bovine incisors enamel (5 × 3 × 2.5 mm(3) ) were divided in four distinct areas: (1) sound (reference area), (2) initial erosion, (3) treatment (irradiated or nonirradiated with CO2 laser), (4) final erosion (after in situ phase). The initial erosive challenge was performed with 1% citric acid (pH = 2.3), for 5 min, 2×/day, for 2 days. The slabs were divided in two groups according to surface treatment: irradiated with CO2 laser (λ = 10.6 µm; 0.5 W) and nonirradiate. After a 2-day lead-in period, 14 volunteers wore an intraoral palatal appliance containing two slabs (irradiated and nonirradiated), in two intraoral phases of 5 days each. Following a cross-over design during the first intraoral phase, half of the volunteers immersed the appliance in 100 mL of citric acid for 5 min, 3×/day, while other half of the volunteers used deionized water (control). The volunteers were crossed over in the second phase. Enamel wear was determined by an optical 3D profilometer. Three-way ANOVA for repeated measures revealed that there was no significant interaction between erosive challenge and CO2 laser irradiation (P = 0.419). Erosive challenge significantly increased enamel wear (P = 0.001), regardless whether or not CO2 laser irradiation was performed. There was no difference in enamel wear between specimens CO2 -laser irradiated and non-irradiated (P = 0.513). Under intraoral conditions, CO2 laser irradiation did not control the progression of erosive lesions in enamel caused by citric acid. © 2014 Wiley Periodicals, Inc.

Development of a control system for a heavy object handling manipulator. Application to a remote maintenance system for ITER blanket module

International Nuclear Information System (INIS)

Yoshimi, Takashi; Tsuji, Kouichi; Miyagawa, Shinichi; Kubo, Tomomi; Kakudate, Satoshi; Tada, Eisuke

2001-01-01

This paper describes a control system for the heavy object handling manipulator. It has been developed for the blanket module remote maintenance system of ITER (International Thermonuclear Fusion Experimental Reactor). A rail-mounted vehicle-type manipulator is proposed for the precise handling of a blanket module which is about 4 tons in weight. Basically, this manipulator is controlled by teaching-playback technique. When grasping or releasing the module, the manipulator sags and the position of the end-effector changes about 50 [mm]. Applying only the usual teaching-playback control makes the smooth operation of setting/removing modules to/from the vacuum vessel wall difficult due to this position change. To solve this proper problem of heavy object handling manipulator, we have developed a system which uses motion patterns generated from two kinds of teaching points. These motion patterns for setting/removing heavy objects are generated by combining teaching points for positioning the manipulator with and without grasping the object. When these motion patterns are applied, the manipulator can transfer the object's weight smoothly at the setting/removing point. This developed system has been applied to the real-scale mock-up of the vehicle manipulator and through the actual module setting/removing experiments, we have verified its effectiveness and realized smooth maintenance operation. (author)

Japanese contributions to the Japan-US workshop on blanket design/technology

International Nuclear Information System (INIS)

Tone, Tatsuzo; Seki, Yasushi; Minato, Akio; Kobayashi, Takeshi; Mori, Seiji; Kawasaki, Hiromitsu; Sumita, Kenji.

1983-02-01

This report describes Japanese papers presented at the Japan-US Workshop on Blanket Design/Technology which was held at Argonne National Laboratory, November 10 - 11, 1982. Overview of Fusion Experimental Reactor (FER), JAERI's activities related to first wall/blanket/shield, summary of FER blanket and its technology development issues and summary of activities at universities on fusion reactor blanket engineering are covered. (author)

Cassette blanket and vacuum building: key elements in fusion reactor maintenance

International Nuclear Information System (INIS)

Werner, R.W.

1977-01-01

The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

[Mechanisms of grass in slope erosion control in Loess sandy soil region of Northwest China].

Science.gov (United States)

Zhao, Chun-Hong; Gao, Jian-En; Xu, Zhen

2013-01-01

By adopting the method of simulated precipitation and from the viewpoint of slope hydrodynamics, in combining with the analysis of soil resistance to erosion, a quantitative study was made on the mechanisms of grass in controlling the slope erosion in the cross area of wind-water erosion in Loess Plateau of Northwest China under different combinations of rainfall intensity and slope gradient, aimed to provide basis to reveal the mechanisms of vegetation in controlling soil erosion and to select appropriate vegetation for the soil and water conservation in Loess Plateau. The grass Astragalus adsurgens with the coverage about 40% could effectively control the slope erosion. This grass had an efficiency of more than 70% in reducing sediment, and the grass root had a greater effect than grass canopy. On bare slope and on the slopes with the grass plant or only the grass root playing effect, there existed a functional relation between the flow velocity on the slopes and the rainfall intensity and slope gradient (V = DJ(0.33 i 0.5), where V is flow velocity, D is the comprehensive coefficient which varies with different underlying surfaces, i is rainfall intensity, and J is slope gradient). Both the grass root and the grass canopy could markedly decrease the flow velocity on the slopes, and increase the slope resistance, but the effect of grass root in decreasing flow velocity was greater while the effect in increasing resistance was smaller than that of grass canopy. The effect of grass root in increasing slope resistance was mainly achieved by increasing the sediment grain resistance, while the effect of canopy was mainly achieved by increasing the slope form resistance and wave resistance. The evaluation of the soil resistance to erosion by using a conceptual model of sediment generation by overland flow indicated that the critical shear stress value of bare slope and of the slopes with the grass plant or only the grass root playing effect was 0.533, 1.672 and 0

Joint Markov Blankets in Feature Sets Extracted from Wavelet Packet Decompositions

Directory of Open Access Journals (Sweden)

Gert Van Dijck

2011-07-01

Full Text Available Since two decades, wavelet packet decompositions have been shown effective as a generic approach to feature extraction from time series and images for the prediction of a target variable. Redundancies exist between the wavelet coefficients and between the energy features that are derived from the wavelet coefficients. We assess these redundancies in wavelet packet decompositions by means of the Markov blanket filtering theory. We introduce the concept of joint Markov blankets. It is shown that joint Markov blankets are a natural extension of Markov blankets, which are defined for single features, to a set of features. We show that these joint Markov blankets exist in feature sets consisting of the wavelet coefficients. Furthermore, we prove that wavelet energy features from the highest frequency resolution level form a joint Markov blanket for all other wavelet energy features. The joint Markov blanket theory indicates that one can expect an increase of classification accuracy with the increase of the frequency resolution level of the energy features.

Neutronics analysis for aqueous self-cooled fusion reactor blankets

International Nuclear Information System (INIS)

Varsamis, G.; Embrechts, M.J.; Jaffa, R.; Steiner, D.; Deutsch, L.; Gierszewski, P.

1986-06-01

The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

EU DEMO blanket concepts safety assessment. Final report of Working Group 6a of the Blanket Concept Selection Exercise

International Nuclear Information System (INIS)

Kleefeldt, K.; Porfiri, T.

1996-06-01

The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four blanket concepts under development. Two of them use lithium ceramics, the other two concepts employ an eutectic lead-lithium alloy (Pb-17Li) as breeder material. The two most promising concepts were to select in 1995 for further development. In order to prepare the selection, a Blanket Concept Selection Exercise (BCSE) has been inititated by the participating associations under the auspices of the European Commission. This BCSE has been performed in 14 working groups which, in a comparative evaluation of the four blanket concepts, addressed specific fields. The working group safety addressed the safety implications. This report describes the methodology adopted, the safety issues identified, their comparative evaluation for the four concepts, and the results and conclusions of the working group to be entered into the overall evaluation. There, the results from all 14 working groups have been combined to yield a final ranking as a basis for the selection. In summary, the safety assessment showed that the four European blanket concepts can be considered as equivalent in terms of the safety rating adopted, each concept, however, rendering safety concerns of different quality in different areas which are substantiated in this report. (orig.) [de

Erosion and Soil Contamination Control Using Coconut Flakes And Plantation Of Centella Asiatica And Chrysopogon Zizanioides

Science.gov (United States)

Roslan, Rasyikin; Che Omar, Rohayu; Nor Zuliana Baharuddin, Intan; Zulkarnain, M. S.; Hanafiah, M. I. M.

2016-11-01

Land degradation in Malaysia due to water erosion and water logging cause of loss of organic matter, biodiversity and slope instability but also land are contaminated with heavy metals. Various alternative such as physical remediation are use but it not showing the sustainability in term of environmental sustainable. Due to that, erosion and soil contamination control using coconut flakes and plantation of Centella asiatica and Chrysopogon zizanioides are use as alternative approach for aid of sophisticated green technology known as phytoremediation and mycoremediation. Soil from cabonaceous phyllite located near to Equine Park, Sri Kembangan are use for monitoring the effect of phytoremediation and mycoremediation in reducing soil contamination and biotechnology for erosion control. Five laboratory scale prototypes were designed to monitor the effect of different proportion of coconut flakes i.e. 10%, 25%, 50% & 100% and plantation of Centella asiatica and Chrysopogon zizanioides to reduce the top soil from eroding and reduce the soil contamination. Prototype have been observe started from first week and ends after 12 weeks. Centella asiatica planted on 10% coconut flakes with 90% soil and Chrysopogon zizanioides planted on 25% coconut flakes with 75% soil are selected proportion to be used as phytoremediation and mycoremediation in reducing soil contamination and biotechnology for erosion control.

Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

International Nuclear Information System (INIS)

Reiter, F.; Tominetti, S.; Perujo, A.

1992-01-01

The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

Fusion breeder sphere - PAC blanket design

International Nuclear Information System (INIS)

Sullivan, J.D.; Palmer, B.J.F.

1987-11-01

There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

Effect of reactor size on the breeding economics of LMFBR blankets

International Nuclear Information System (INIS)

Tagishi, A.; Driscoll, M.J.

1975-02-01

The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

Assessment of alkali metal coolants for the ITER blanket

International Nuclear Information System (INIS)

Natesan, K.; Reed, C.B.; Mattas, R.F.

1994-01-01

The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

Analysis of ER string test thermally instrumented interconnect 80-K MLI blanket

International Nuclear Information System (INIS)

Daly, E.; Pletzer, R.

1992-04-01

An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m 2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m 2 . The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

A review of fusion breeder blanket technology, part 1

International Nuclear Information System (INIS)

Jackson, D.P.; Selander, W.N.; Townes, B.M.

1985-01-01

This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

Wood strands as an alternative to agricultural straw for erosion control

Science.gov (United States)

Randy B. Foltz; James H. Dooley

2004-01-01

Agricultural straw is used in forested areas of the United States for erosion control on burned areas, harvest landings, decommissioned road prisms, road cuts and fills, and other areas of disturbed soil. However, an increased agronomic and ecological value for straw; an increased utilization for energy production, fiber panels, and other higher value uses; a...

Accelerator driven heavy water blanket on circulating fuel

International Nuclear Information System (INIS)

Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

1997-01-01

A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

FW/Blanket and vacuum vessel for RTO/RC ITER

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M.

2000-01-01

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste

FW/Blanket and vacuum vessel for RTO/RC ITER

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M

2000-11-01

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, {approx}50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

Evaluation of different techniques for erosion control on different roadcuts in its first year of implantation

Science.gov (United States)

Gomez, Jose Alfonso; Rodríguez, Abraham; Viedma, Antonio; Contreras, Valentin; Vanwalleghem, Tom; Taguas, Encarnación V.; Giráldez, Juan Vicente

2014-05-01

Linear infrastructures, such as highways and railways, present a large environmental impact. Among this impact is the effect on landscape and the modification of the hydrological conditions of the area and an increase in erosive processes (Martin et al., 2011). The increase of erosive processes is specially significant in roadbanks, resulting in high maintenance costs as well as security risks for the use of the infrastructure if it is not properly controlled. Among roadbanks, roadcuts are specially challenging areas for erosion control and ecological restoration, due to their usually steep slope gradient and poor conditions for establishment of vegetation. There are several studies in Mediterranean conditions indicating how the combination of semiarid conditions with, sporadic, intense rainfall events makes a successful vegetation development and erosion control in motorway roadbanks extremely difficult (e.g. Andrés and Jorbat, 2000; Bochet and García-Fayos, 2004). This communication presents the results of the first year evaluation (hydrological year 2012-2013) of five different erosion control strategies on six different locations under different materials on roadcuts of motorways or railways in Andalusia during 2012-2013 using natural rainfall and simulated rainfall. The six sites were located on roadcuts between 10 and 20 m long on slope steepness ranging from 40 to 90%, in motorways and railways spread over different materials in Andalusia. Site 1, Huelva was located on consolidated sand material, sites 2, Osuna I, site 3, Osuna II and site 4, Mancha Real, on marls. Sites 5, Guadix, and 6, Fiñana, were located on phyllites, in comparison a harder material. At each site 12 plots (10 m long and 2 m wide) were installed using metal sheets buried 10 cm within the soil with their longest side in the direction of the roadcut maximum slope. Six different treatments were evaluated at each site, two replications each. These treatments were: 1- A control with bare

Progress in blanket designs using SiCf/SiC composites

International Nuclear Information System (INIS)

Giancarli, L.; Golfier, H.; Nishio, S.; Raffray, R.; Wong, C.; Yamada, R.

2002-01-01

This paper summarizes the most recent design activities concerning the use of SiC f /SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium-lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiC f /SiC box acting as a container for the lithium-lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li 2 O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R and D on SiC f /SiC

Overview of the TFTB lithium blanket module program

International Nuclear Information System (INIS)

Jassby, D.L.

1986-01-01

The Lithium Blanket Module (LBM) is an ∼ 80-cm 3 module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program

The history and assessment of effectiveness of soil erosion control measures deployed in Russia

Directory of Open Access Journals (Sweden)

Valentin Golosov

2013-09-01

Full Text Available Research activities aimed at design and application of soil conservation measures for reduction of soil losses from cultivated fields started in Russia in the last quarter of the 19th century. A network of "zonal agrofor-estry melioration experimental stations" was organized in the different landscape zones of Russia in the first half of the 20th century. The main task of the experiments was to develop effective soil conservation measures for Russian climatic,soil and land use conditions. The most widespread and large-scale introduction of coun-termeasures to cope with soil erosion by water and wind into agricultural practice supported by serious governmental investments took place during the Soviet Union period after the Second World War. After the Soviet Union collapse in 1991 ,general deterioration of the agricultural economy sector and the absence of investments resulted in cessation of organized soil conservation measures application at the nation-wide level. However, some of the long-term erosion control measures such as forest shelter belts, artificial slope terracing, water diversion dams above formerly active gully heads survived until the present. In the case study of sediment redistribution within the small cultivated catchment presented in this paper an attempt was made to evaluate average annual erosion rates on arable slopes with and without soil conservation measures for two time intervals. It has been found that application of conservation measures on cultivated slopes within the experimental part of the case study catchment has led to a decrease of average soil loss rates by at least 2. 5 2. 8 times. The figures obtained are in good agreement with previously published results of direct monitoring of snowmelt erosion rates, reporting approximately a 3 -fold decrease of average snowmelt erosion rates in the experimental sub-catchment compared to a traditionally cultivated control sub-catchment. A substantial decrease of soil

Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

International Nuclear Information System (INIS)

Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

1987-11-01

The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

The fusion blanket program at Chalk River

International Nuclear Information System (INIS)

Hastings, I.J.

1986-03-01

Work on the Fusion Blanket Program commenced at Chalk River in 1984 June. Co-funded by Canadian Fusion Fuels Technology Project and Atomic Energy of Canada Limited, the Program utilizes Chalk River expertise in instrumented irradiation testing, ceramics, tritium technology, materials testing and compound chemistry. This paper gives highlights of studies to date on lithium-based ceramics, leading contenders for the fusion blanket

Use of Ball Blanket in attention-deficit/hyperactivity disorder sleeping problems

DEFF Research Database (Denmark)

Hvolby, Allan; Bilenberg, Niels

2011-01-01

Objectives: Based on actigraphic surveillance, attention-deficit/hyperactivity disorder (ADHD) symptom rating and sleep diary, this study will evaluate the effect of Ball Blanket on sleep for a sample of 8-13-year-old children with ADHD. Design: Case-control study. Setting: A child and adolescent...... psychiatric department of a teaching hospital. Participants: 21 children aged 8-13 years with a diagnosis of ADHD and 21 healthy control subjects. Intervention: Sleep was monitored by parent-completed sleep diaries and 28 nights of actigraphy. For 14 of those days, the child slept with a Ball Blanket. Main...... outcome measures: The sleep latency, number of awakenings and total length of sleep was measured, as was the possible influence on parent- and teacher-rated ADHD symptom load. Results: The results of this study will show that the time it takes for a child to fall asleep is shortened when using a Ball...

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 1

International Nuclear Information System (INIS)

Malang, S.; Reimann, J.; Sebening, H.; Barleon, L.; Bogusch, E.; Bojarsky, E.; Borgstedt, H.U.; Buehler, L.; Casal, V.; Deckers, H.; Feuerstein, H.; Fischer, U.; Frees, G.; Graebner, H.; John, H.; Jordan, T.; Kramer, W.; Krieg, R.; Lenhart, L.; Malang, S.; Meyder, R.; Norajitra, P.; Reimann, J.; Schwenk-Ferrero, A.; Schnauder, H.; Stieglitz, R.; Oschinski, J.; Wiegner, E.

1991-12-01

A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary, Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated R and D-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required R and D-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

2003-10-01

Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

Aqueous self-cooled blanket concepts for fusion reactors

International Nuclear Information System (INIS)

Varsamis, G.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Gierszewski, P.

1987-01-01

A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

Choice of economical optimum blanket of hybrid reactors

Energy Technology Data Exchange (ETDEWEB)

Blinkin, V L; Novikov, V M

1981-01-01

The economical effectiveness of symbiotic power systems depends on the choice of the correlation between energy production and fissile fuel production in blankets of controlled thermonuclear fusion reactor (CTR), what is investigated here. It is shown that the optimum value of this correlation essentially depends on the ratio between the specific costs for energy production in hybrid thermonuclear reactors and that in fission reactors as part of the symbiotic system.

Comparison of measured and calculated composition of irradiated EBR-II blanket assemblies

International Nuclear Information System (INIS)

Grimm, K. N.

1998-01-01

In anticipation of processing irradiated EBR-II depleted uranium blanket subassemblies in the Fuel Conditioning Facility (FCF) at ANL-West, it has been possible to obtain a limited set of destructive chemical analyses of samples from a single EBR-II blanket subassembly. Comparison of calculated values with these measurements is being used to validate a depletion methodology based on a limited number of generic models of EBR-II to simulate the irradiation history of these subassemblies. Initial comparisons indicate these methods are adequate to meet the operations and material control and accountancy (MC and A) requirements for the FCF, but also indicate several shortcomings which may be corrected or improved

Tectonic control of erosion in the southern Central Andes

Science.gov (United States)

Val, Pedro; Venerdini, Agostina L.; Ouimet, William; Alvarado, Patricia; Hoke, Gregory D.

2018-01-01

Landscape evolution modeling and global compilations of exhumation data indicate that a wetter climate, mainly through orographic rainfall, can govern the spatial distribution of erosion rates and crustal strain across an orogenic wedge. However, detecting this link is not straightforward since these relationships can be modulated by tectonic forcing and/or obscured by heavy-tailed frequencies of catchment discharge. This study combines new and published along-strike average rates of catchment erosion constrained by 10Be and river-gauge data in the Central Andes between 28°S and 36°S. These data reveal a nearly identical latitudinal pattern in erosion rates on both sides of the range, reaching a maximum of 0.27 mm/a near 34°S. Collectively, data on topographic and fluvial relief, variability of rainfall and discharge, and crustal seismicity suggest that the along-strike pattern of erosion rates in the southern Central Andes is largely independent of climate, but closely relates to the N-S distribution of shallow crustal seismicity and diachronous surface uplift. The consistently high erosion rates on either side of the orogen near 34°S imply that climate plays a secondary role in the mass flux through an orogenic wedge where the perturbation to base level is similar on both sides.

Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

2015-10-15

Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

Continuous fine pattern formation by screen-offset printing using a silicone blanket

Science.gov (United States)

Nomura, Ken-ichi; Kusaka, Yasuyuki; Ushijima, Hirobumi; Nagase, Kazuro; Ikedo, Hiroaki; Mitsui, Ryosuke; Takahashi, Seiya; Nakajima, Shin-ichiro; Iwata, Shiro

2014-09-01

Screen-offset printing combines screen-printing on a silicone blanket with transference of the print from the blanket to a substrate. The blanket absorbs organic solvents in the ink, and therefore, the ink does not disperse through the material. This prevents blurring and allows fine patterns with widths of a few tens of micrometres to be produced. However, continuous printing deteriorates the pattern’s shape, which may be a result of decay in the absorption abilities of the blanket. Thus, we have developed a new technique for refreshing the blanket by substituting high-boiling-point solvents present on the blanket surface with low-boiling-point solvents. We analyse the efficacy of this technique, and demonstrate continuous fine pattern formation for 100 screen-offset printing processes.

Continuous fine pattern formation by screen-offset printing using a silicone blanket

International Nuclear Information System (INIS)

Nomura, Ken-ichi; Kusaka, Yasuyuki; Ushijima, Hirobumi; Nagase, Kazuro; Ikedo, Hiroaki; Mitsui, Ryosuke; Takahashi, Seiya; Nakajima, Shin-ichiro; Iwata, Shiro

2014-01-01

Screen-offset printing combines screen-printing on a silicone blanket with transference of the print from the blanket to a substrate. The blanket absorbs organic solvents in the ink, and therefore, the ink does not disperse through the material. This prevents blurring and allows fine patterns with widths of a few tens of micrometres to be produced. However, continuous printing deteriorates the pattern’s shape, which may be a result of decay in the absorption abilities of the blanket. Thus, we have developed a new technique for refreshing the blanket by substituting high-boiling-point solvents present on the blanket surface with low-boiling-point solvents. We analyse the efficacy of this technique, and demonstrate continuous fine pattern formation for 100 screen-offset printing processes. (paper)

Development of simulator for remote handling system of ITER blanket

International Nuclear Information System (INIS)

Takeda, Nobukazu; Kakudate, Satoshi; Nakanhira, Masataka; Matsumoto, Yasuhiro; Shibanuma, K.

2007-01-01

The maintenance activity in the ITER has to be performed remotely because 14 MeV neutron caused by fusion reaction induces activation of structural material and emission of gamma ray. In general, it is one of the most critical issues to avoid collision between the remote maintenance system and in-vessel components. Therefore, the visual information in the vacuum vessel is required strongly to understand arrangement of these devices and components. However, there is a limitation of arrangement of viewing cameras in the vessel because of high intensity of gamma ray. It is expected that enough numbers of cameras and lights are not available because of arrangement restriction. Furthermore, visibility of the interested area such as the contacting part is frequently disturbed by the devices and components, thus it is difficult to recognize relative position between the devices and components only by visual information even if enough cameras and lights are equipped. From these reasons, the simulator to recognize the positions of each devices and components is indispensable for remote handling systems in fusion reactors. The authors have been developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robot simulation software ''ENVISION''. The simulator is connected to the control system of the manipulator which was developed as a part of the blanket maintenance system in the EDA and can reconstruct the positions of the manipulator and the blanket module using the position data of the motors through the LAN. In addition, it can provide virtual visual information, such as the connecting operation behind the blanket module with making the module transparent on the screen. It can be used also for checking the maintenance sequence before the actual operation. The developed simulator will be modified further adding other necessary functions and finally completed as a prototype of the actual simulator for the blanket remote handling system

Acoustic contributions of a sound absorbing blanket placed in a double panel structure: absorption versus transmission.

Science.gov (United States)

Doutres, Olivier; Atalla, Noureddine

2010-08-01

The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound absorption behavior can affect the performance of the double panel structure.

The evolution of US helium-cooled blankets

International Nuclear Information System (INIS)

Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

1991-01-01

This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

Control of Eolian soil erosion from waste site surface barriers

International Nuclear Information System (INIS)

Ligotke, M.W.

1994-11-01

Physical models were tested in a wind tunnel to determine optimum surface-ravel admixtures for protecting silt-loam soil from erosion by, wind and saltating, sand stresses. The tests were performed to support the development of a natural-material surface barrier for and waste sites. Plans call for a 2-m deep silt-loam soil reservoir to retain infiltrating water from rainfall and snowmelt. The objective of the study was to develop a gravel admixture that would produce an erosion-resistant surface layer during, periods of extended dry climatic stress. Thus, tests were performed using simulated surfaces representing dry, unvegetated conditions present just after construction, after a wildfire, or during an extended drought. Surfaces were prepared using silt-loam soil mixed with various grades of sand and Travel. Wind-induced surface shear stresses were controlled over the test surfaces, as were saltating, sand mass flow rates and intensities. Tests were performed at wind speeds that approximated and exceeded local 100-year peak gust intensities. Surface armors produced by pea gravel admixtures were shown to provide the best protection from wind and saltating sand stresses. Compared with unprotected silt-loam surfaces, armored surfaces reduced erosion rates by more than 96%. Based in part on wind tunnel results, a pea gravel admixture of 15% will be added to the top 1 in of soil in a prototype barrier under construction in 1994. Field tests are planned at the prototype site to provide data for comparison with wind tunnel results

Effectiveness of the GAEC cross-compliance standard Short-term measures for runoff water control on sloping land (temporary ditches and grass strips in controlling soil erosion

Directory of Open Access Journals (Sweden)

Paolo Bazzoffi

2011-08-01

Full Text Available The agronomic measures made obligatory by the cross-compliance Standard Temporary measures for runoff water control on sloping land included in the Ministry of Agricultural, Food and Forestry Policies (MiPAAF decree on cross compliance until 2008, and by Standard 1.1 Creation of temporary ditches for the prevention of soil erosion in the 2009 decree, certainly appear to be useful for the control of soil erosion and runoff. The efficacy of temporary drainage ditches and of grass strips in controlling runoff and erosion has been demonstrated in trials conducted in field test plots in Italy. When level temporary drainage ditches are correctly built, namely with an inclination of not more than 2.5% in relation to the maximum hillslope gradient, they allow the suspended sediment eroded upstream to settle in the ditches, retaining the material carried away on the slope and, as a result, reducing the quantity of sediment delivered to the hydrographic network. In particular, among all the results, the erosion and runoff data in a trial conducted in Guiglia (Modena showed that in corn plots, temporary drainage ditches reduced soil erosion by 94%, from 14.4 Mg ha-1 year-1 (above the limit established by the NRCS-USDA of 11.2 Mg ha-1 year-1 to 0.8 Mg ha-1 year-1 (within the NRCS limit and also within the more restrictive limit established by the OECD of 6.0 Mg ha-1 year-1. With respect to the grass buffer strips the most significant research was carried out in Volterra. This research demonstrated their efficacy in reducing erosion from 8.15 Mg ha-1 to 1.6 Mg ha-1, which is approximately 5 times less than the erosion observed on bare soil. The effectiveness of temporary drainage ditches was also assessed through the application of the Revised Universal Soil Loss Equation (RUSLE erosion model to 60 areas under the control of the Agency for Agricultural Payments (AGEA in 2009, comparing the risk of erosion in these sample areas by simulating the presence and

Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

International Nuclear Information System (INIS)

Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

1984-01-01

The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

International Nuclear Information System (INIS)

Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

1978-01-01

An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

The transpiration cooled first wall and blanket concept

International Nuclear Information System (INIS)

Barleon, Leopold; Wong, Clement

2002-01-01

To achieve high thermal performance at high power density the EVOLVE concept was investigated under the APEX program. The EVOLVE W-alloy first wall and blanket concept proposes to use transpiration cooling of the first wall and boiling or vaporizing lithium (Li) in the blanket zone. Critical issues of this concept are: the Magnetohydrodynamic (MHD) pressure losses of the Li circuit, the evaporation through a capillary structure and the needed superheating of the Li at the first wall and blanket zones. Application of the transpiration concept to the blanket region results in the integrated transpiration cooling concept (ITCC) with either toroidal or poloidal first wall channels. For both orientations the routing of the liquid Li and the Li vapor has been modeled and the corresponding pressure losses have been calculated by varying the width of the supplying slot and the capillary diameter. The concept works when the sum of the active and passive pumping head is higher than the total system pressure losses and when the temperature at the inner side of the first wall does not override the superheating limit of the coolant. This cooling concept has been extended to the divertor design, and the removal of a surface heat flux of up to 10 MW/m 2 appears to be possible, but this paper will focus on the transpiration cooled first wall and blanket concept assessment

Optimal Land Use Management for Soil Erosion Control by Using an Interval-Parameter Fuzzy Two-Stage Stochastic Programming Approach

Science.gov (United States)

Han, Jing-Cheng; Huang, Guo-He; Zhang, Hua; Li, Zhong

2013-09-01

Soil erosion is one of the most serious environmental and public health problems, and such land degradation can be effectively mitigated through performing land use transitions across a watershed. Optimal land use management can thus provide a way to reduce soil erosion while achieving the maximum net benefit. However, optimized land use allocation schemes are not always successful since uncertainties pertaining to soil erosion control are not well presented. This study applied an interval-parameter fuzzy two-stage stochastic programming approach to generate optimal land use planning strategies for soil erosion control based on an inexact optimization framework, in which various uncertainties were reflected. The modeling approach can incorporate predefined soil erosion control policies, and address inherent system uncertainties expressed as discrete intervals, fuzzy sets, and probability distributions. The developed model was demonstrated through a case study in the Xiangxi River watershed, China's Three Gorges Reservoir region. Land use transformations were employed as decision variables, and based on these, the land use change dynamics were yielded for a 15-year planning horizon. Finally, the maximum net economic benefit with an interval value of [1.197, 6.311] × 109 was obtained as well as corresponding land use allocations in the three planning periods. Also, the resulting soil erosion amount was found to be decreased and controlled at a tolerable level over the watershed. Thus, results confirm that the developed model is a useful tool for implementing land use management as not only does it allow local decision makers to optimize land use allocation, but can also help to answer how to accomplish land use changes.

Optimal land use management for soil erosion control by using an interval-parameter fuzzy two-stage stochastic programming approach.

Science.gov (United States)

Han, Jing-Cheng; Huang, Guo-He; Zhang, Hua; Li, Zhong

2013-09-01

Soil erosion is one of the most serious environmental and public health problems, and such land degradation can be effectively mitigated through performing land use transitions across a watershed. Optimal land use management can thus provide a way to reduce soil erosion while achieving the maximum net benefit. However, optimized land use allocation schemes are not always successful since uncertainties pertaining to soil erosion control are not well presented. This study applied an interval-parameter fuzzy two-stage stochastic programming approach to generate optimal land use planning strategies for soil erosion control based on an inexact optimization framework, in which various uncertainties were reflected. The modeling approach can incorporate predefined soil erosion control policies, and address inherent system uncertainties expressed as discrete intervals, fuzzy sets, and probability distributions. The developed model was demonstrated through a case study in the Xiangxi River watershed, China's Three Gorges Reservoir region. Land use transformations were employed as decision variables, and based on these, the land use change dynamics were yielded for a 15-year planning horizon. Finally, the maximum net economic benefit with an interval value of [1.197, 6.311] × 10(9) $ was obtained as well as corresponding land use allocations in the three planning periods. Also, the resulting soil erosion amount was found to be decreased and controlled at a tolerable level over the watershed. Thus, results confirm that the developed model is a useful tool for implementing land use management as not only does it allow local decision makers to optimize land use allocation, but can also help to answer how to accomplish land use changes.

Fusion blanket high-temperature heat transfer

International Nuclear Information System (INIS)

Fillo, J.A.

1983-01-01

Deep penetration of 14 MeV neutrons makes two-temperature region blankets feasible. A relatively low-temperature (approx. 300 0 C) metallic structure is the vacuum/coolant pressure boundary, while the interior of the blanket, which is a simple packed bed of nonstructural material, operates at very high temperatures (>1000 0 C). The water-cooled shell structure is thermally insulated from the steam-cooled interior. High-temperature steam can dramatically increase the efficiency of electric power generation, as well as produce hydrogen and oxygen-based synthetic fuels at high-efficiency

Tritium behaviour in ceramic breeder blankets

International Nuclear Information System (INIS)

Miller, J.M.

1989-01-01

Tritium release from the candidate ceramic materials, Li 2 O, LiA10 2 , Li 2 SiO 3 , Li 4 SiO 4 and Li 2 ZrO 3 , is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

Feasibility study of fusion breeding blanket concept employing graphite reflector

International Nuclear Information System (INIS)

Cho, Seungyon; Ahn, Mu-Young; Lee, Cheol Woo; Kim, Eung Seon; Park, Yi-Hyun; Lee, Youngmin; Lee, Dong Won

2015-01-01

Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

Feasibility study of fusion breeding blanket concept employing graphite reflector

Energy Technology Data Exchange (ETDEWEB)

Cho, Seungyon, E-mail: sycho@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Woo; Kim, Eung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

Progress on DEMO blanket attachment concept with keys and pins

International Nuclear Information System (INIS)

Vizvary, Zsolt; Iglesias, Daniel; Cooper, David; Crowe, Robert; Riccardo, Valeria

2015-01-01

Highlights: • DEMO blanket attachment system with keys and pins (without using bolts). • Blanket segments are preloaded by progressively designed springs. • Blanket back plate flexibility has a major impact on spring design. • Mechanical analysis of other components indicates no unresolvable issues. • Thermal analysis indicates acceptable temperatures for the support system. - Abstract: The blanket attachment has to cope with gravity, thermal and electromagnetic loads, also it has to be installed and serviced by remote handling. Pre-stressed components suffer from stress relaxation in irradiated environments such as DEMO. To circumvent this problem pre-stressed component should be either avoided or shielded, and where possible keys and pins should be used. This strategy has been proposed for the DEMO multi-module segments (MMS). The blanket segments are held by two tapered keys each, designed to allow thermal expansions while providing contact with the vacuum vessel and to resist the poloidal and radial moments the latter being dominant at 9.1 MNm inboard and 15 MNm outboard. On the top of the blanket segment there is a pin which provides vertical support. At the bottom another vertical support has to lock them in position after installation and manage the pre-load on the segments. The pre-load is required to deal with the electromagnetic loads during disruption. This is provided by a set of springs, which require shielding as they are preloaded. These are sized to cope with the force (3 MN inboard, 1.4 MN outboard) due to halo currents and the toroidal moment which can reverse. Calculations show that the flexibility of the blanket segment itself plays a significant role in defining the required support system. The blanket segment acts as a preloaded spring and it has to be part of the attachment design as well.

Nuclear characteristics of D-D fusion reactor blankets, (1)

International Nuclear Information System (INIS)

Nakashima, Hideki; Ohta, Masao; Seki, Yasushi.

1977-01-01

Fusion reactors operating on the deuterium (D-D) cycle are considered promising for their freedom from tritium breeding in the blanket. In this paper, neutronic and photonic calculations are undertaken covering several blanket models of the D-D fusion reactor, using presently available data, with a view to comparing the nuclear characteristics of these models, in particular, the nuclear heating rates and their spatial distributions. Nine models are taken up for the study, embodying various combinations of coolant, blanket, structural and reflector materials. About 10 MeV is found to be a typical value for the total nuclear energy deposition per source neutron in the models considered here. The realization of high energy gain is contingent upon finding a favorable combination of blanket composition and configuration. The resulting implications on the thermal design aspect are briefly discussed. (auth.)

Systematic methodology for estimating direct capital costs for blanket tritium processing systems

International Nuclear Information System (INIS)

Finn, P.A.

1985-01-01

This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

International Nuclear Information System (INIS)

Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

2016-01-01

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

Water-cooled, fire boom blanket, test and evaluation for system prototype development

International Nuclear Information System (INIS)

Stahovec, J. G.; Urban, R. W.

1999-01-01

Initial development of actively cooled fire booms indicated that water-cooled barriers could withstand direct oil fire for several hours with little damage if cooling water were continuously supplied. Despite these early promising developments, it was realized that to build reliable full-scale system for Navy host salvage booms would require several development tests and lengthy evaluations. In this experiment several types of water-cooled fire blankets were tested at the Oil and Hazardous Materials Simulated Test Tank (OHMSETT). After the burn test the blankets were inspected for damage and additional tests were conducted to determine handling characteristics for deployment, recovery, cleaning and maintenance. Test results showed that water-cooled fire boom blankets can be used on conventional offshore oil containment booms to extend their use for controlling large floating-oil marine fires. Results also demonstrated the importance of using thermoset rubber coated fabrics in the host boom to maintain sufficient reserve seam strength at elevated temperatures. The suitability of passively cooled covers should be investigated to protect equipment and boom from indirect fire exposure. 1 ref., 2 tabs., 8 figs

Optimization of beryllium for fusion blanket applications

International Nuclear Information System (INIS)

Billone, M.C.

1993-01-01

The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

Runoff erosion

OpenAIRE

Evelpidou, Niki (Ed.); Cordier, Stephane (Ed.); Merino, Agustin (Ed.); Figueiredo, Tomás de (Ed.); Centeri, Csaba (Ed.)

2013-01-01

Table of Contents PART I – THEORY OF RUNOFF EROSION CHAPTER 1 - RUNOFF EROSION – THE MECHANISMS CHAPTER 2 - LARGE SCALE APPROACHES OF RUNOFF EROSION CHAPTER 3 - MEASURING PRESENT RUNOFF EROSION CHAPTER 4 - MODELLING RUNOFF EROSION CHAPTER 5 - RUNOFF EROSION AND HUMAN SOCIETIES: THE INFLUENCE OF LAND USE AND MANAGEMENT PRACTICES ON SOIL EROSION PART II - CASE STUDIES CASE STUDIES – INTRODUCTION: RUNOFF EROSION IN MEDITERRANEAN AREA CASE STUDY 1: Soil Erosion Risk...

Monthly Rainfall Erosivity Assessment for Switzerland

Science.gov (United States)

Schmidt, Simon; Meusburger, Katrin; Alewell, Christine

2016-04-01

Water erosion is crucially controlled by rainfall erosivity, which is quantified out of the kinetic energy of raindrop impact and associated surface runoff. Rainfall erosivity is often expressed as the R-factor in soil erosion risk models like the Universal Soil Loss Equation (USLE) and its revised version (RUSLE). Just like precipitation, the rainfall erosivity of Switzerland has a characteristic seasonal dynamic throughout the year. This inter-annual variability is to be assessed by a monthly and seasonal modelling approach. We used a network of 86 precipitation gauging stations with a 10-minute temporal resolution to calculate long-term average monthly R-factors. Stepwise regression and Monte Carlo Cross Validation (MCCV) was used to select spatial covariates to explain the spatial pattern of R-factor for each month across Switzerland. The regionalized monthly R-factor is mapped by its individual regression equation and the ordinary kriging interpolation of its residuals (Regression-Kriging). As covariates, a variety of precipitation indicator data has been included like snow height, a combination of hourly gauging measurements and radar observations (CombiPrecip), mean monthly alpine precipitation (EURO4M-APGD) and monthly precipitation sums (Rhires). Topographic parameters were also significant explanatory variables for single months. The comparison of all 12 monthly rainfall erosivity maps showed seasonality with highest rainfall erosivity in summer (June, July, and August) and lowest rainfall erosivity in winter months. Besides the inter-annual temporal regime, a seasonal spatial variability was detectable. Spatial maps of monthly rainfall erosivity are presented for the first time for Switzerland. The assessment of the spatial and temporal dynamic behaviour of the R-factor is valuable for the identification of more susceptible seasons and regions as well as for the application of selective erosion control measures. A combination with monthly vegetation

Design requirement on KALIMER blanket fuel assembly duct

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

1998-03-01

This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

Stress analysis of blanket vessel for JAERI experimental fusion reactor

International Nuclear Information System (INIS)

Sako, K.; Minato, A.

1979-01-01

A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

Status of blanket design for RTO/RC ITER

International Nuclear Information System (INIS)

Yamada, M.; Ioki, K.; Cardella, A.; Elio, F.; Miki, N.

2000-01-01

Design has progressed on the FW/blanket for the RTO/RC (reduced technical objective/ reduced cost) ITER. The basic functions and structures are the same as for the 1998 ITER design. However, design and fabrication methods of the FW/blanket have been improved to achieve ∝ 50% reduction of the construction cost compared to that for the 1998 ITER design. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the EDA (engineering design activity) is still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed. (orig.)

Fusion blankets for catalyzed D--D and D--He3 reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

Fusion blankets for catalyzed D--D and D--3He reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

Tritium inventory and permeation in the ITER breeding blanket

International Nuclear Information System (INIS)

Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

2000-01-01

A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 2. Detailed version

International Nuclear Information System (INIS)

John, H.; Malang, S.; Sebening, H.

1991-12-01

A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary. Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated RandD-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required RandD-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

Calculations of tritium breeding ratio and inventory distributions of FEB blanket

International Nuclear Information System (INIS)

Deng Baiquan

2001-01-01

Based on the design features of FEB reactor blanket, the tritium breeding ratio and tritium concentrations in liquid lithium of each breeding zone have been calculated after 10 days full power operation for outboard blanket and one day operation for inboard blanket. The comparisons with the results calculated by Monte-Carlo code MORSE-CGT are made. Meanwhile the inventory in beryllium multiplier after one-year full power operation has also been estimated. An important conclusion has been drew the thermal hydraulic design should be careful to guarantee the blanket temperature should not rise as high as 680 degree C

A coupled systems code-CFD MHD solver for fusion blanket design

Energy Technology Data Exchange (ETDEWEB)

Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

2015-10-15

Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

DEMO relevance of the test blanket modules in ITER-Application to the European test blanket modules

International Nuclear Information System (INIS)

Magnani, E.; Gabriel, F.; Boccaccini, L.V.; Li-Puma, A.

2010-01-01

Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.

Evaluation of chemical stabilizers and windscreens for wind erosion control of uranium mill tailings

International Nuclear Information System (INIS)

Elmore, M.R.; Hartley, J.N.

1984-08-01

Potential wind erosion of uranium mill tailings is a concern for the surface disposal of tailings at uranium mills. Wind-blown tailings may subsequently be redeposited on areas outside the impoundment. Pacific Northwest Laboratory (PNL) is investigating techniques for fugitive dust control at uranium mill tailings piles. Laboratory tests, including wind tunnel studies, were conducted to evaluate the relative effectiveness of 43 chemical stabilizers. Seventeen of the more promising stabilizers were applied to test plots on a uranium tailings pile at the American Nuclear Corporation-Gas Hills Project mill site in central Wyoming. The durabilities of these materials under actual site conditions were evaluated over time. In addition, field testing of commercially available windscreens was conducted. Test panels were constructed of eight different materials at the Wyoming test site to compare their durability. A second test site was established near PNL to evaluate the effectiveness of windscreens at reducing wind velocity, and thereby reduce the potential for wind erosion of mill tailings. Results of the laboratory land field tests of the chemical stabilizers and windscreens are presented, along with costs versus effectiveness of these techniques for control of wind erosion at mill tailings piles. 12 references, 4 figures, 6 tables

Reflux disease as an etiological factor of dental erosion

Directory of Open Access Journals (Sweden)

Stojšin Ivana

2010-01-01

Full Text Available Introduction Gastroesophageal reflux is a frequent disease which has a significant influence on the development of dental erosions. Objective The aim of this research was to determine the frequency of dental erosions among the patients with gastroesophageal reflux, as well as to verify the most common symptoms of gastroesophageal disease. Methods The research comprised of two groups, each consisting of 30 patients aged 18-80 years. The experimental group comprised of patients diagnosed with gastroesophageal reflux disease (GERD, while the control group was composed of patients who were not diagnosed with GERD. Based on the illness history data, all patients of the experimental group were registered to have gastroesophageal and extraesophageal symptoms. Dental erosions were diagnosed during a stomatological inspection by using index system according to Eccles and Jenkins. Data processing was accomplished by the Statgraphics Centurion software package. Results Dental erosions were found in 76.7% of experimental group patients, and in 53.3% of control group patients. Fortynine percent of teeth of the experimental group patients and 31.1% of the control group patients showed erosive changes. On average, the number of teeth with erosions in the experimental group was 15.7 per person and in the control group 10 per person. The teeth of the front region of the upper jaw, as well as the lower first molars had the highest average value of dental erosion index. In the experimental group 12.8% of teeth and 24% of teeth in the control group were diagnosed to have dental erosion index value 1. Furthermore, 23.4% of teeth in the experimental group and 7.1% of teeth in the control group were registered to have dental erosion index value 2. Finally, the dental erosion index value 3 was found in 13.0% of teeth in the experimental group only. The highest average value of regional erosion index in the experimental group was found in the region 13-23 equalling 1

Overview of first wall/blanket/shield technology

International Nuclear Information System (INIS)

Nygren, R.E.

1983-04-01

This brief overview of first wall, blanket, and shield technology focuses first on changes and trends in important design issues from the 1970's to the 1980's, then on current perceptions of critical issues in first wall, blanket, and shield design and related technology. The emphasis is on base technology rather than either systems engineering or materials development, on the two primary confinement systems, tokamaks and mirrors, and on production of electricity as the primary goal for development

Solid breeder test blanket module design and analysis

Energy Technology Data Exchange (ETDEWEB)

Ying, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)]. E-mail: ying@fusion.ucla.edu; Abdou, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Calderoni, P. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Sharafat, S. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Youssef, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); An, Z. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Abou-Sena, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Kim, E. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Reyes, S. [LANL, Livermore, CA (United States); Willms, S. [LANL, Los Alamos, NM (United States); Kurtz, R. [PNNL, Richland, WA (United States)

2006-02-15

This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.

Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

Energy Technology Data Exchange (ETDEWEB)

Mistrangelo, C., E-mail: chiara.mistrangelo@kit.edu; Bühler, L.

2014-10-15

Highlights: • MHD flows in model geometries of DCLL blanket manifolds. • Study of velocity, pressure distributions and flow partitioning in parallel ducts. • Flow partitioning affected by 3D MHD pressure drop and velocity distribution in the expanding zone. • Reduced pressure drop in a continuous expansion compared to a sudden expansion. - Abstract: An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure. Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules. In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.

Studies on steps affecting tritium residence time in solid blanket

International Nuclear Information System (INIS)

Tanaka, Satoru

1987-01-01

For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

Experimental programme in support of the development of the European ceramic-breeder-inside-tube test-blanket: present status and future work

International Nuclear Information System (INIS)

Proust, E.; Roux, N.; Flament, T.; Anzidei, L.; ENEA, Frascati; Casadio, S.; Dell'orco, G.

1992-01-01

Four DEMO blanket classes are under investigation within the framework of the European Test-Blanket Development Programme. One of them is featured by the use of lithium ceramic breeder pellets contained inside externally helium cooled tubes. This paper summarizes the main achievements to date of the experimental programme supporting the development of this class of blanket. It also gives an outline of the areas of the breeder material, beryllium, tritium control, and thermomechanical tests, the future work envisaged for the 92-94 period. 53 refs

Tritium inventory in Li2ZrO3 blanket

International Nuclear Information System (INIS)

Nishikawa, M.; Baba, A.

1998-01-01

Recently, we have presented the way to estimate the tritium inventory in a solid breeder blanket considering effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions. It is reported in our previous paper that the estimated tritium inventory for a LiAlO 2 blanket agrees well with data observed in various in situ experiments when the effective diffusivity of tritium from the EXOTIC-6 experiment is used and that the better agreement is obtained when existence of some water vapor is assumed in the purge gas. The same way as used for a LiAlO 2 blanket is applied to a Li 2 ZrO 3 blanket in this study and the estimated tritium inventory shows a good agreement with data obtained in such in situ experiments as MOZART, EXOTIC-6 and TRINE experiments. (orig.)

Application of vanadium alloys to a fusion reactor blanket

Energy Technology Data Exchange (ETDEWEB)

Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

1984-05-01

Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

Energy Technology Data Exchange (ETDEWEB)

Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

2016-11-15

Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

Processing and waste disposal representative for fusion breeder blanket systems

International Nuclear Information System (INIS)

Finn, P.A.; Vogler, S.

1987-01-01

This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made

Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

Science.gov (United States)

Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

2006-02-01

This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

Energy Technology Data Exchange (ETDEWEB)

Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

2016-11-01

Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

International Nuclear Information System (INIS)

Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

1985-01-01

Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

Immunohistochemical Study of p53 Expression in Patients with Erosive and Non-Erosive Oral Lichen Planus

Science.gov (United States)

Shiva, Atena; Zamanian, Ali; Arab, Shahin; Boloki, Mahsa

2018-01-01

Statement of the Problem: Oral lichen planus is a common mucocutaneous lesion with a chronic inflammatory process mediated by immune factors while a few cases of the disease become malignant. Purpose: This study aimed to determine the frequency of p53 marker as a tumor suppressor in patients with erosive and non-erosive oral lichen planus (OLP) by using immunohistochemical methods. Materials and Method: This descriptive cross-sectional study investigated the p53 expression in 16 erosive OLP, 16 non-erosive OLP samples, and 8 samples of normal oral mucosa through immunohistochemistry. The percentage of stained cells in basal and suprabasal layers, and inflammatory infiltrate were graded according to the degree of staining; if 0%, 50% of the cells were stained, they were considered as (-), (+), (++), (+++) and (++++), respectively. The obtained data was statistically analyzed and compared by using Chi square and Fisher’s exact test. Results: The mean percentage of p53 positive cells in erosive OLP (34.5±14.2) was considerably higher than that in non-erosive OLP (23.8±10.4) and normal mucosa (17.5±17). There was a significant difference among the three groups of erosive, non-erosive and control in terms of staining intensity. No significant difference existed between the patients’ age and sex in the two OLP groups. Conclusion: The increased incidence of p53 from normal mucosa to erosive OLP indicated the difference between biological behavior of erosive and non-erosive OLP. It can be claimed that the erosive OLP has great premalignant potential compared with the non-erosive one.

Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

Energy Technology Data Exchange (ETDEWEB)

Jolodosky, A. [Univ. of California, Berkeley, CA (United States); Fratoni, M. [Univ. of California, Berkeley, CA (United States)

2015-09-22

Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within a low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding

Applications of the Aqueous Self-Cooled Blanket concept

International Nuclear Information System (INIS)

Steiner, D.; Embrechts, M.J.; Varsamis, G.; Wrisley, K.; Deutch, L.; Gierszewski, P.

1986-01-01

In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids

Potential and problems of an aqueous lithium salt solution blanket for NET

International Nuclear Information System (INIS)

Kuechle, M.; Bojarsky, E.; Dorner, S.; Fischer, U.; Reimann, J.; Reiser, H.

1987-07-01

The report describes design studies on a water cooled in-vessel shield blanket for NET and its modification into an aqueous lithium salt blanket. The shield blankets are exchangable against breeding blankets and fulfill their shielding and heat removal functions. Emphasis is on simplicity and reliability. The water cooled shield is a large steel container in the shape of the blanket segment which is filled by water and containes a grid structure of poloidally arranged steel plates. The water flows several times in poloidal direction through the channels formed by the steel plates and is thereby heated up from 40degC to 70degC. When the water is replaced by an aqueous lithium salt solution the shield can be converted into a tritium breeding blanket without any design modification or invessel component replacement. When compared with other concepts this blanket has the advantage that the solution can replace water cooling also in the divertor and in segments dedicated to plasma heating and diagnostics, what increases the coverage considerably. Extensive three-dimensional neutronics calculations were done which, together with literature studies on candidate materials, corrosion, and tritium recovery led to a first assessment of the concept. There is an indication that no major corrosion problems are to be expected in the low temperature region envisaged. Tritium recovery capital costs were estimated to be in the 20 MECU to 50 MECU range and tritium breeding ratio is comparable to the best breeding blanket. (orig./GG) [de

First Wall, Blanket, Shield Engineering Technology Program

International Nuclear Information System (INIS)

Nygren, R.E.

1982-01-01

The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

Availability analysis of the ITER blanket remote handling system

International Nuclear Information System (INIS)

Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

2015-01-01

The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

Status of fusion reactor blanket evaluation studies in France

International Nuclear Information System (INIS)

Carre, F.; Chevereau, G.; Gervaise, F.; Proust, E.

1985-03-01

In the frame of recent CEA studies aiming at the evaluation and at the comparison of various candidate blanket concepts in moderate power conditions (Psub(n) approximately 2 MW/m 2 ), the present work examines the neutronic and thermomechanical performances of a water cooled Li 17 Pb 83 tubular blanket and those of a helium cooled canister blanket taking advantage of the excellent breeding capability of composite Beryllium/LiAlO 2 (85/15%) breeder elements. The purpose of the following discussion is to justify the impetus for these reference concepts and to summarize the state of their evaluation studies updated by the continuous assimilation of calculations and experiments in progress

Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

Science.gov (United States)

Xiaokang, ZHANG; Songlin, LIU; Xia, LI; Qingjun, ZHU; Jia, LI

2017-11-01

The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR). Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage, and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW. The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3. The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.

Bed erosion control at 60 degree river confluence using vanes

Science.gov (United States)

Wuppukondur, Ananth; Chandra, Venu

2017-04-01

Confluences are common occurrences along natural rivers. Hydrodynamics at the confluence is complex due to merging of main and lateral flows with different characteristics. Bed erosion occurs at the confluence due to turbulence and also secondary circulation induced by centrifugal action of the lateral flow. The eroded sediment poses various problems in the river ecosystem including river bank failure. Reservoirs are majorly affected due to sediment deposition which reduces storage capacity. The bed erosion also endangers stability of pipeline crossings and bridge piers. The aim of this experimental study is to check the performance of vanes in controlling bed erosion at the confluence. Experiments are performed in a 600 confluence mobile bed model with a non-uniform sediment of mean particle size d50 = 0.28mm. Discharge ratio (q=ratio of lateral flow discharge to main flow discharge) is maintained as 0.5 and 0.75 with a constant average main flow depth (h) of 5cm. Vanes of width 0.3h (1.5cm) and thickness of 1 mm are placed along the mixing layer at an angle of 150, 300 and 600(with respect to main flow) to perform the experiments. Also, two different spacing of 2h and 3h (10cm and 15cm) between the vanes are used for conducting the experiments. A digital point gauge with an accuracy of ±0.1mm is used to measure bed levels and flow depths at the confluence. An Acoustic Doppler Velocitimeter (ADV) with a frequency of 25Hz and accuracy of ±1mm/s is used to measure flow velocities. Maximum scour depth ratio Rmax, which is ratio between maximum scour depth (Ds) and flow depth (h), is used to present the experimental results.From the experiments without vanes, it is observed that the velocities are increasing along the mixing layer and Rmax=0.82 and 1.06, for q=0.5 and 0.75, respectively. The velocities reduce with vanes since roughness increases along the mixing layer. For q=0.5 and 0.75, Rmax reduces to 0.62 and 0.7 with vanes at 2h spacing, respectively. Similarly

Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

Science.gov (United States)

Leshukov, A. Yu.; Strebkov, Yu. S.; Sviridenko, M. N.; Safronov, V. M.; Putrik, A. B.

2017-12-01

In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 "Blanket First Wall" of February 14, 2014, and 1.6.P3.RF.01 "Blanket Module Connections" of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m2. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015-2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

In plain sight: the Chesapeake Bay crater ejecta blanket

Science.gov (United States)

Griscom, D. L.

2012-02-01

idealized calculation of the CBIS ejecta-blanket elevation profile minutes after the impact was carried out founded on well established rules for explosion and impact-generated craters. This profile is shown here to match the volume of the upland deposits ≥170 km from the crater center. Closer to the crater, much of the "postdicted" ejecta blanket has clearly been removed by erosion. Nevertheless the Shirley and fossil-free Bacons Castle Formations, located between the upland deposits and the CBIS interior and veneering the present day surface with units ∼10-20 m deep, are respectively identified as curtain- and excavation-phase ejecta. The neritic-fossil-bearing Calvert Formation external to the crater is deduced to be of Eocene age (as opposed to early Miocene as currently believed), preserved by the armoring effects of the overlying CBIS ejecta composed of the (distal) upland deposits and the (proximal) Bacons Castle Formation. The lithofacies of the in-crater Calvert Formation can only have resulted from inward mass wasting of the postdicted ejecta blanket, vestiges of which (i.e. the Bacons Castle and Shirley Formations) still overlap the crater rim and sag into its interior, consistent with this expectation. Because there appear to be a total of ∼10 000 km2 of CBIS ejecta lying on the present-day surface, future research should center the stratigraphic, lithologic, and petrologic properties of these ejecta versus both radial distance from the crater center (to identify ejecta from different ejection stages) and circumferentially at fixed radial distances (to detect possible anisotropies relating the impact angle and direction of approach of the impactor). The geological units described here may comprise the best preserved, and certainly the most accessible, ejecta blanket of a major crater on the Earth's surface and therefore promise to be a boon to the field of impact geology. As a corollary, a major revision of the current stratigraphic column of the M

Some new ideas for Tandem Mirror blankets

International Nuclear Information System (INIS)

Neef, W.S. Jr.

1981-01-01

The Tandem Mirror Reactor, with its cylindrical central cell, has led to numerous blanket designs taking advantage of the simple geometry. Also many new applications for fusion neutrons are now being considered. To the pure fusion electricity producers and hybrids producing fissile fuel, we are adding studies of synthetic fuel producers and fission-suppressed hybrids. The three blanket concepts presented are new ideas and should be considered illustrative of the breadth of Livermore's application studies. They are not meant to imply fully analyzed designs

A development of user-friendly graphical interface for a blanket simulator

International Nuclear Information System (INIS)

Lee, Young-Seok; Yoon, Seok-Heun; Han, Jung-Hoon

2010-01-01

A web-based user-friendly graphical interface (GUI) system, named GUMBIS (Graphical User-friendly Monte-Carlo-Application Blanket-Design Interface System), was developed to cut down the efforts of the researchers and practitioners who study tokamak blanket designs with the Monte Carlo MCNP/MCNPX codes. GUMBIS was also aimed at supporting them to use the codes for their study without having through understanding on the complex menus and commands of the codes. Developed on the web-based environment, GUMBIS provides task sharing capability on a network. GUMBIS, applicable for both blanket design and neutronics analysis, could facilitate not only advanced blanket R and D but also the education and training of the researchers in the R and D.

Electromagnetic effects involving a tokamak reactor first wall and blanket

International Nuclear Information System (INIS)

Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

1980-01-01

Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

2015-05-15

In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

Molten salt cooling/17Li-83Pb breeding blanket concept

International Nuclear Information System (INIS)

Sze, D.K.; Cheng, E.T.

1985-02-01

A description of a fusion breeding blanket concept using draw salt coolant and static 17 Li- 83 Pb is presented. 17 Li- 83 Pb has high breeding capability and low tritium solubility. Draw salt operates at low pressure and is inert to water. Corrosion, MHD, and tritium containment problems associated with the MARS design are alleviated because of the use of a static LiPb blanket. Blanket tritium recovery is by permeation toward the plasma. A direct contact steam generator is proposed to eliminate some generic problems associated with a tube shell steam generator

The control of divertor carbon erosion/redeposition in the DIII-D tokamak

International Nuclear Information System (INIS)

Whyte, D.G.; West, W.P.; Wong, C.P.C.

2001-01-01

The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon source rate arising from chemical erosion is found to be very low in the detached divertor. Near strikepoint regions, the rate of carbon deposition is ∼3 cm/burn-year, with a corresponding hydrogenic codeposition rate >1kg/m 2 /burn-year; rates both problematic for steady-state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon influx from the main wall is measured to be relatively large in the high-density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/redeposition and impurity plasma transport. (author)

Three procedures for estimating erosion from construction areas

International Nuclear Information System (INIS)

Abt, S.R.; Ruff, J.F.

1978-01-01

Erosion from many mining and construction sites can lead to serious environmental pollution problems. Therefore, erosion management plans must be developed in order that the engineer may implement measures to control or eliminate excessive soil losses. To properly implement a management program, it is necessary to estimate potential soil losses from the time the project begins to beyond project completion. Three methodologies are presented which project the estimated soil losses due to sheet or rill erosion of water and are applicable to mining and construction areas. Furthermore, the three methods described are intended as indicators of the state-of-the-art in water erosion prediction. The procedures herein do not account for gully erosion, snowmelt erosion, wind erosion, freeze-thaw erosion or extensive flooding

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 2

International Nuclear Information System (INIS)

Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

1991-10-01

The BOT (Breeder Outside Tube) Helium Cooled Solid Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test blankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of the test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

Instrumentation and methods evaluations for shallow land burial of waste materials: water erosion

International Nuclear Information System (INIS)

Hostetler, D.D.; Murphy, E.M.; Childs, S.W.

1981-08-01

The erosion of geologic materials by water at shallow-land hazardous waste disposal sites can compromise waste containment. Erosion of protective soil from these sites may enhance waste transport to the biosphere through water, air, and biologic pathways. The purpose of this study was to review current methods of evaluating soil erosion and to recommend methods for use at shallow-land, hazardous waste burial sites. The basic principles of erosion control are: minimize raindrop impact on the soil surface; minimize runoff quantity; minimize runoff velocity; and maximize the soil's resistance to erosion. Generally soil erosion can be controlled when these principles are successfully applied at waste disposal sites. However, these erosion control practices may jeopardize waste containment. Typical erosion control practices may enhance waste transport by increasing subsurface moisture movement and biologic uptake of hazardous wastes. A two part monitoring program is recommended for US Department of Energy (DOE) hazardous waste disposal sites. The monitoring programs and associated measurement methods are designed to provide baseline data permitting analysis and prediction of long term erosion hazards at disposal sites. These two monitoring programs are: (1) site reconnaissance and tracking; and (2) site instrumentation. Some potential waste transport problems arising from erosion control practices are identified. This report summarizes current literature regarding water erosion prediction and control

Ecological site-based assessments of wind and water erosion: informing accelerated soil erosion management in rangelands

Science.gov (United States)

Webb, Nicholas P.; Herrick, Jeffrey E.; Duniway, Michael C.

2014-01-01

Accelerated soil erosion occurs when anthropogenic processes modify soil, vegetation or climatic conditions causing erosion rates at a location to exceed their natural variability. Identifying where and when accelerated erosion occurs is a critical first step toward its effective management. Here we explore how erosion assessments structured in the context of ecological sites (a land classification based on soils, landscape setting and ecological potential) and their vegetation states (plant assemblages that may change due to management) can inform systems for reducing accelerated soil erosion in rangelands. We evaluated aeolian horizontal sediment flux and fluvial sediment erosion rates for five ecological sites in southern New Mexico, USA, using monitoring data and rangeland-specific wind and water erosion models. Across the ecological sites, plots in shrub-encroached and shrub-dominated vegetation states were consistently susceptible to aeolian sediment flux and fluvial sediment erosion. Both processes were found to be highly variable for grassland and grass-succulent states across the ecological sites at the plot scale (0.25 Ha). We identify vegetation thresholds that define cover levels below which rapid (exponential) increases in aeolian sediment flux and fluvial sediment erosion occur across the ecological sites and vegetation states. Aeolian sediment flux and fluvial erosion in the study area can be effectively controlled when bare ground cover is 100 cm in length is less than ~35%. Land use and management activities that alter cover levels such that they cross thresholds, and/or drive vegetation state changes, may increase the susceptibility of areas to erosion. Land use impacts that are constrained within the range of natural variability should not result in accelerated soil erosion. Evaluating land condition against the erosion thresholds identified here will enable identification of areas susceptible to accelerated soil erosion and the development of

Ecological site‐based assessments of wind and water erosion: informing accelerated soil erosion management in rangelands.

Science.gov (United States)

Webb, Nicholas P; Herrick, Jeffrey E; Duniway, Michael C

Accelerated soil erosion occurs when anthropogenic processes modify soil, vegetation, or climatic conditions causing erosion rates at a location to exceed their natural variability. Identifying where and when accelerated erosion occurs is a critical first step toward its effective management. Here we explored how erosion assessments structured in the context of ecological sites (a land classification based on soils, landscape setting, and ecological potential) and their vegetation states (plant assemblages that may change due to management) can inform systems for reducing accelerated soil erosion in rangelands. We evaluated aeolian horizontal sediment flux and fluvial sediment erosion rates for five ecological sites in southern New Mexico, USA, using monitoring data and rangeland-specific wind and water erosion models. Across the ecological sites, plots in shrub-encroached and shrub-dominated vegetation states were consistently susceptible to aeolian sediment flux and fluvial sediment erosion. Both processes were found to be highly variable for grassland and grass–succulent states across the ecological sites at the plot scale (0.25 ha). We identified vegetation thresholds that define cover levels below which rapid (exponential) increases in aeolian sediment flux and fluvial sediment erosion occur across the ecological sites and vegetation states. Aeolian sediment flux and fluvial erosion in the study area could be effectively controlled when bare ground cover was 100 cm in length was less than ∼35%. Land use and management activities that alter cover levels such that they cross thresholds, and/or drive vegetation state changes, may increase the susceptibility of areas to erosion. Land use impacts that are constrained within the range of natural variability should not result in accelerated soil erosion. Evaluating land condition against the erosion thresholds identified here will enable identification of areas susceptible to accelerated soil erosion and the

The role of secondary minerals in the control of erosion processes under a Mediterranean mining landcape

International Nuclear Information System (INIS)

Penas, J. M.; Garcia, G.; Manteca, J. I.

2009-01-01

The result of mining activity is the presence of several slit ponds and mining tailings spread all over the Sierra Minera (Cartagena La Union Mountains, SE Spain). These ponds, joint to other wastes deposits constitute the main source of heavy metals to the environment. Besides, these metal sources areas act as dispersion focus towards the surrounding and subsidiary areas due to the erosion processes. Interaction between metal and salts present in these environments, provoke an secondary effect on the landscape modelling. The major o minor strength of the erosion processes is controlled by the presence of salts in soil and mining wastes (silt ponds and mining tailings). The aim of this work concerns the relation- between the salt-metal compounds and the erosion and landscape modeling processes. (Author) 4 refs.

Proceedings of the eleventh international workshop on ceramic breeder blanket interactions

International Nuclear Information System (INIS)

Enoeda, Mikio

2004-07-01

This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li 2 TiO 3 and so on, fabrication technology developments and characterization of the Li 2 TiO 3 and Li 4 SiO 4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li 2 TiO 3 and Li 4 SiO 4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

Effects of buffer thickness on ATW blanket performance

International Nuclear Information System (INIS)

Yang, W. S.; Mercatali, L.; Taiwo, T. A.; Hill, R. N.

2001-01-01

This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy ( and lt; 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level

Effects of Buffer Thickness on ATW Blanket Performance

International Nuclear Information System (INIS)

Yang, W.S.; Mercatali, L.; Taiwo, T.A.; Hill, R.N.

2002-01-01

This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level. (authors)

Key achievements in elementary R and Ds on water-cooled solid breeder blanket for ITER Test Blanket Module in JAERI

International Nuclear Information System (INIS)

Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Tanigawa, H.; Tobita, K.; Akiba, M.; Hayashi, K.; Ochiai, K.; Nishitani, T.

2005-01-01

This paper presents significant progress in research and development (R and D) of key elementary technologies on the water-cooled solid breeder blanket for the ITER test blanket modules (TBMs) in JAERI. Development of module fabrication technology, bonding technology of armors, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup, and tritium release behavior from Li 2 TiO 3 pebble bed under neutron pulsed operation condition are summarized. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 deg C followed by normalizing at 930 deg C after the Hot Isostatic Pressing (HIP) process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a solid state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it was found that the thermal fatigue lifetime of F82H can be predicted by using Manson-Coffin's law. As for R and Ds on a breeder material, Li 2 TiO 3 , effective thermal conductivity of Li 2 TiO 3 pebble was measured under compressive force simulating the ITER TBM environment. The increase in the effective thermal conductivity of the pebble bed was about 2.5 % at the compressive strain of 0.9 % at 400 deg C. Neutronic performance of the blanket module mockup has been carried out by the 14 MeV neutron irradiation. It was confirmed that the measured tritium production rate agreed with the calculated values within about 10% difference. Also, tritium release from a Li 2 TiO 3 pebble bed was measured under pulsed neutron irradiation conditions simulating the ITER operation. (author)

Key achievements in elementary R and D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

International Nuclear Information System (INIS)

Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

2006-01-01

This paper presents the significant progress made in the research and development (R and D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li 2 TiO 3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 0 C followed by normalizing it at 930 0 C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R and D on the breeder material, Li 2 TiO 3 , the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li 2 TiO 3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li 2 TiO 3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation

Neutronics design aspects of reference ARIES-I fusion blanket

International Nuclear Information System (INIS)

Cheng, E.T.

1990-12-01

A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

Effects of buffer thickness on ATW blanket performances

International Nuclear Information System (INIS)

Yang, Won Sik

2001-01-01

This paper presents the preliminary results of target and buffer design studies for a lead-bismuth eutectic (LBE) cooled accelerator transmutation of waste (ATW) system, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using an 840 MWt LBE cooled ATW design, the effects of buffer thickness on the blanket performances have been studied. Varying the buffer thickness for a given blanket configuration, system performances have been estimated by a series of calculations using MCNPX and REBUS-3 codes. The effects of source importance change are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. As the irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. The results show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable

Two-phase-flow cooling concept for fusion reactor blankets

International Nuclear Information System (INIS)

Bender, D.J.; Hoffman, M.A.

1977-01-01

The new two-phase heat transfer medium proposed is a mixture of potassium droplets and helium which permits blanket operation at hih temperature and low pressure, while maintaining acceptable pumping power requirements, coolant ducting size, and blanket structure fractions. A two-phase flow model is described. The helium pumping power and the primary heat transfer loop are discussed

Blast venting through blanket material in the HYLIFE ICF reactor

International Nuclear Information System (INIS)

Liu, J.C.; Peterson, P.F.; Schrock, V.E.

1992-01-01

This work presents a numerical study of blast venting through various blanket configurations in the HYLIFE ICF reactor design. The study uses TSUNAMI -- a multi-dimensional, high-resolution, shock capturing code -- to predict the momentum exchange and gas dynamics for blast venting in complex geometries. In addition, the study presents conservative predictions of wall loading by gas shock and impulse delivered to the protective liquid blanket. Configurations used in the study include both 2700 MJ and 350 MJ fusion yields per pulse for 5 meter and 3 meter radius reactor chambers. For the former, an annular jet array is used for the blanket geometry, while in the latter, both annular jet array as well as slab geometries are used. Results of the study indicate that blast venting and wall loading may be manageable in the HYLIFE-II design by a judicious choice of blanket configuration

Peningkatan mutu blanket karet alam melalui proses predrying dan penyemprotan asap cair

Directory of Open Access Journals (Sweden)

Afrizal Vachlepi

2017-06-01

Full Text Available Most of Indonesian rubber products SIR 20 are made from the material of raw rubber obtained from smallholders. However, the quality of this material is not good enough. Thus, quality improvement has to be carried out by manufacturers. The liquid smoke used during the blanket hanging process can improve the quality of the rubber products SIR 20. This research aimed to determine and study the effects of liquid smoke spraying and blanket hanging duration on the drying factor, the dry rubber content, technical quality, vulcanization characteristics, and physical properties of vulcanized natural rubber. Treatments consisted of various hanging duration (6, 8, and 10 days, and without hanging and spraying (with and without spraying of liquid smoke. The results showed that the spraying of liquid smoke on natural rubber blankets could improve the technical quality of the natural rubber, especially the values of Po and PRI. The spraying of liquid smoke could reduce the blanket hanging duration to 6-8 days. The blankets sprayed with liquid smoke had the optimum cure time of around 15 minutes and 19 seconds and the scorch time of around 3 minutes and 22 seconds. These values indicated that the vulcanization characteristics of blankets which were sprayed with liquid smoke were generally better than those of blankets which were not sprayed with liquid smoke

Main features and potentialities of gas-blanket systems

International Nuclear Information System (INIS)

Lehnert, B.

1977-02-01

A review is given of the features and potentialities of cold-blanket systems, with respect to plasma equilibrium, stability, and reactor technology. The treatment is concentrated on quasi-steady magnetized plasmas confined at moderately high beta values. The cold-blanket concept has specific potentialities as a fusion reactor, e.g. in connection with the desired densities and dimensions of full-scale systems, refuelling, as well as ash and impurity removal, and stability. (author)

Advanced methods comparisons of reaction rates in the Purdue Fast Breeder Blanket Facility

International Nuclear Information System (INIS)

Hill, R.N.; Ott, K.O.

1988-01-01

A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab

R and D status on Water Cooled Ceramic Breeder Blanket Technology

Energy Technology Data Exchange (ETDEWEB)

Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji

2014-10-15

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li{sub 2}TiO{sub 3} pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

Conceptual design of blanket structures for fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

International Nuclear Information System (INIS)

1978-01-01

This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified

Flow balancing in liquid metal blankets

International Nuclear Information System (INIS)

Tillack, M.S.; Morley, N.B.

1995-01-01

Non-uniform flow distribution between parallel channels is one of the most serious concerns for self-cooled liquid metal blankets with electrically insulated walls. We show that uncertainties in flow distribution can be dramatically reduced by relatively simple design modifications. Several design features which impose flow uniformity by electrically coupling parallel channels are surveyed. Basic mechanisms for ''flow balancing'' are described, and a particular self-regulating concept using discrete passive electrodes is proposed for the US ITER advanced blanket concept. Scoping calculations suggest that this simple technique can be very powerful in equalizing the flow, even with massive insulator failures in individual channels. More detailed analyses and experimental verification will be required to demonstrate this concept for ITER. (orig.)

Control of two-phase erosion corrosion with the amine 5-aminopentanol: rig and plant trials

International Nuclear Information System (INIS)

Lewis, G.G.; Greene, J.C.; Tyldesley, J.D.; Wetton, E.A.M.; Fountain, M.J.

1994-01-01

Control of two-phase erosion corrosion in the once through mild steel boilers of the gas cooled nuclear power station at Wylfa was achieved by using the amine 2-amino, 2 methylpropan-1-ol (AMP). In a search to find a more cost effective amine, 5-aminopentanol (5-AP) emerged, from a laboratory based programme to determine basicity and volatility, as the most promising candidate. The effectiveness of 5-AP in controlling erosion corrosion was demonstrated in a rig test, carried out on a full scale replica of a Wylfa boiler tube. Following on from the rig test, a plant trial at Wylfa PS demonstrated 5-AP's superior thermal stability (compared to AMP). It also provided confirmation that the laboratory generated data on basicity and volatility was applicable to plant and hence also the accuracy of the figures for predicted amine usage. (orig.)

Preliminary analyses of neutronics schemes for three kinds waste transmutation blankets of fusion-fission hybrid

International Nuclear Information System (INIS)

Zhang Mingchun; Feng Kaiming; Li Zaixin; Zhao Fengchao

2012-01-01

The neutronics schemes of the helium-cooled waste transmutation blanket, sodium-cooled waste transmutation blanket and FLiBe-cooled waste transmutation blanket were preliminarily calculated and analysed by using the spheroidal tokamak (ST) plasma configuration. The neutronics properties of these blankets' were compared and analyzed. The results show that for the transmutation of "2"3"7Np, FLiBe-cooled waste transmutation blanket has the most superior transmutation performance. The calculation results of the helium-cooled waste transmutation blanket show that this transmutation blanket can run on a steady effective multiplication factor (k_e_f_f), steady power (P), and steady tritium production rate (TBR) state for a long operating time (9.62 years) by change "2"3"7Np's initial loading rate of the minor actinides (MA). (authors)

Nuclear Analyses of Indian LLCB Test Blanket System in ITER

Science.gov (United States)

Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

2017-04-01

Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

Status of EC solid breeder blanket designs and R and D for demo fusion reactors

International Nuclear Information System (INIS)

Proust, E.; Anzidei, L.; Moons, F.

1994-01-01

Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li 4 SiO 4 and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs

NOEL: a no-leak fusion blanket concept

International Nuclear Information System (INIS)

Powell, J.R.; Yu, W.S.; Fillo, J.A.; Horn, F.L.; Makowitz, H.

1980-01-01

Analysis and tests of a no-leak fusion blanket concept (NOEL-NO External Leak) are described. Coolant cannot leak into the plasma chamber even if large through-cracks develop in the first wall. Blanket modules contain a two-phase material, A, that is solid (several cm thick) on the inside of the module shell, and liquid in the interior. The solid layer is maintained by imbedded tubes carrying a coolant, B, below the freezing point of A. Most of the 14-MeV neutron energy is deposited as heat in the module interior. The thermal energy flow from the module interior to the shell keeps the interior liquid. Pressure on the liquid A interior is greater than the pressure on B, so that B cannot leak out if failures occur in coolant tubes. Liquid A cannot leak into the plasma chamber through first wall cracks because of the intervening frozen layer. The thermal hydraulics and neutronics of NOEL blankets have been investigated for various metallic (e.g., Li, Pb 2 , LiPb, Pb) and fused salt choices for material A

Space environment durability of beta cloth in LDEF thermal blankets

Science.gov (United States)

Linton, Roger C.; Whitaker, Ann F.; Finckenor, Miria M.

1993-01-01

Beta cloth performance for use on long-term space vehicles such as Space Station Freedom (S.S. Freedom) requires resistance to the degrading effects of the space environment. The major issues are retention of thermal insulating properties through maintaining optical properties, preserving mechanical integrity, and generating minimal particulates for contamination-sensitive spacecraft surfaces and payloads. The longest in-flight test of beta cloth's durability was on the Long Duration Exposure Facility (LDEF), where it was exposed to the space environment for 68 months. The LDEF contained 57 experiments which further defined the space environment and its effects on spacecraft materials. It was deployed into low-Earth orbit (LEO) in Apr. 1984 and retrieved Jan. 1990 by the space shuttle. Among the 10,000 plus material constituents and samples onboard were thermal control blankets of multilayer insulation with a beta cloth outer cover and Velcro attachments. These blankets were exposed to hard vacuum, thermal cycling, charged particles, meteoroid/debris impacts, ultraviolet (UV) radiation, and atomic oxygen (AO). Of these space environmental exposure elements, AO appears to have had the greatest effect on the beta cloth. The beta cloth analyzed in this report came from the MSFC Experiment S1005 (Transverse Flat-Plate Heat Pipe) tray oriented approximately 22 deg from the leading edge vector of the LDEF satellite. The location of the tray on LDEF and the placement of the beta cloth thermal blankets are shown. The specific space environment exposure conditions for this material are listed.

Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

2001-11-01

Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

2001-01-01

Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

18 CFR 284.284 - Blanket certificates for unbundled sales services.

Science.gov (United States)

2010-04-01

... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Blanket certificates for unbundled sales services. 284.284 Section 284.284 Conservation of Power and Water Resources... Sales by Interstate Pipelines § 284.284 Blanket certificates for unbundled sales services. (a...

Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

2016-11-01

Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

Computation Method Comparison for Th Based Seed-Blanket Cores

International Nuclear Information System (INIS)

Kolesnikov, S.; Galperin, A.; Shwageraus, E.

2004-01-01

This work compares two methods for calculating a given nuclear fuel cycle in the WASB configuration. Both methods use the ELCOS Code System (2-D transport code BOXER and 3-D nodal code SILWER) [4] are compared. In the first method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated separately for each region by the 2-D transport code. In the second method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated from Seed-Blanket Colorsets (Fig.1) calculated by the 2-D transport code. The evaluation of the error introduced by the first method is the main objective of the present study

Using REE tracers to measure sheet erosion changing to rill erosion

International Nuclear Information System (INIS)

Liu Puling; Xue Yazhou; Song Wei; Wang Mingyi; Ju Tongjun

2004-01-01

Rare Earth Elements (REE) tracer method was used to study sheet erosion changing to rill erosion on slope land. By placing different rare earth elements of different soil depth across a slope in an indoor plot, two simulated rainfalls were applied to study the change of erosion type and the rill erosion process. The results indicate that the main erosion type is sheet erosion at the beginning of the rainfalls, and serious erosion happens after rill erosion appears. Accumulated sheet and rill erosion amounts increase with the rainfalls time. The percentage of sheet erosion amount decreases and rill erosion percentage increases with time. At the end of the rainfalls, the total rill erosion amounts are 4-5 times more than sheet erosion. In this paper, a new REE tracer method was used to quantitatively distinguish sheet and rill erosion amounts. The new REE tracer method should be useful to future studying of erosion processes on slope lands. (authors)

Erosion-corrosion synergistics in the low erosion regime

International Nuclear Information System (INIS)

Corey, R.G.; Sethi, V.K.

1986-01-01

Many engineering alloys display good high temperature corrosion resistance. However, when they are used in corrosive environments where they are subjected to erosion also, the corrosion resistance has been adversely affected. The phenomenon known as erosion-corrosion is complex and requires detailed investigation of how the erosion and corrosion kinetics interact and compete. At the Kentucky Center for Energy Research Laboratory, an erosion-corrosion tester was used to perform erosion-oxidation tests on 2 1/4 Cr-1 Mo steel at 500-600 0 C using alumina abrasive at low velocities. The erosion-oxidation rate data and morphology of exposed surfaces are consistent with oxide chipping and fracturing being the mode of material loss

Synthesis and Characterization of Fibre Reinforced Silica Aerogel Blankets for Thermal Protection

Directory of Open Access Journals (Sweden)

S. Chakraborty

2016-01-01

Full Text Available Using tetraethoxysilane (TEOS as the source of silica, fibre reinforced silica aerogels were synthesized via fast ambient pressure drying using methanol (MeOH, trimethylchlorosilane (TMCS, ammonium fluoride (NH4F, and hexane. The molar ratio of TEOS/MeOH/(COOH2/NH4F was kept constant at 1 : 38 : 3.73 × 10−5 : 0.023 and the gel was allowed to form inside the highly porous meta-aramid fibrous batting. The wet gel surface was chemically modified (silylation process using various concentrations of TMCS in hexane in the range of 1 to 20% by volume. The fibre reinforced silica aerogel blanket was obtained subsequently through atmospheric pressure drying. The aerogel blanket samples were characterized by density, thermal conductivity, hydrophobicity (contact angle, and Scanning Electron Microscopy. The radiant heat resistance of the aerogel blankets was examined and compared with nonaerogel blankets. It has been observed that, compared to the ordinary nonaerogel blankets, the aerogel blankets showed a 58% increase in the estimated burn injury time and thus ensure a much better protection from heat and fire hazards. The effect of varying the concentration of TMCS on the estimated protection time has been examined. The improved thermal stability and the superior thermal insulation of the flexible aerogel blankets lead to applications being used for occupations that involve exposure to hazards of thermal radiation.

Considerations on techniques for improving tritium confinement in helium-cooled ceramic breeder blankets

International Nuclear Information System (INIS)

Fuetterer, M.A.; Raepsaet, X.; Proust, E.; Leger, D.

1994-01-01

Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (author) 8 refs.; 2 figs

The role of forest stand density in controlling soil erosion: implications to sediment-related disasters in Japan.

Science.gov (United States)

Razafindrabe, Bam H N; He, Bin; Inoue, Shoji; Ezaki, Tsugio; Shaw, Rajib

2010-01-01

The role of forest stand density in controlling soil erosion was investigated in Ehime Prefecture, Japan. The main objective was to compare soil erosion under different forest conditions including forest type, species composition, and stand density as influenced by thinning operations. Relative yield index (Ry) was used as an indicator of stand density to reflect the degree of management operations in the watershed. Eleven treatments were established based on the above forest conditions. Soil loss was collected in each of the 11 treatments after each rainfall event for a period of 1 year. The paper presents summary data on soil loss as affected by forest conditions and rainfall patterns. Findings showed that an appropriate forest management operation, which can be insured by stand density control, is needed to reduce soil loss. The present study plays an important role in clarifying technical processes related to soil erosion, while it helps linking these elements to current Japanese forestry issues and bringing new inputs to reducing sediment-related disasters in Japan.

Structural effects on fusion reactor blankets due to liquid metals in magnetic fields

International Nuclear Information System (INIS)

Lehner, J.R.; Reich, M.; Powell, J.R.

1976-01-01

The transient stress distribution caused in the blanket structure when the plasma current suddenly switches off in a time short compared to the L/R decay time of the liquid metal blanket was studied. Poloidal field of the plasma will induce a current to flow in the liquid metal and blanket walls. Since the resistance of the liquid lithium will be much less than that of the metal walls, the current can be considered as flowing around the blanket near the cross section perimeter, but in the lithium

Design study of blanket structure based on a water-cooled solid breeder for DEMO

Energy Technology Data Exchange (ETDEWEB)

Someya, Youji; Tobita, Kenji; Utoh, Hiroyasu; Tokunaga, Shinji; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru

2015-10-15

Highlights: • Neutronics design of a water-cooled solid mixed breeder blanket was presented. • The blanket concept achieves a self-sufficient supply of tritium by neutronics analysis. • The overall outlet coolant temperature was 321 °C, which is in the acceptable range. - Abstract: Blanket concept with a simplified interior for mass production has been developed using a mixed bed of Li{sub 2}TiO{sub 3} and Be{sub 12}Ti pebbles, coolant conditions of 15.5 MPa and 290–325 °C and cooling pipes without any partitions. Considering the continuity with the ITER test blanket module option of Japan and the engineering feasibility in its fabrication, our design study focused on a water-cooled solid breeding blanket using the mixed pebbles bed. Herein, we propose blanket segmentation corresponding to the shape and dimension of the blanket and routing of the coolant flow. Moreover, we estimate the overall tritium breeding ratio (TBR) with a torus configuration, based on the segmentation using three-dimensional (3D) Monte Carlo N-particle calculations. As a result, the overall TBR is 1.15. Our 3D neutronics analysis for TBR ensures that the blanket concept can achieve a self-sufficient supply of tritium.

Erosion and corrosion of nuclear power plant materials

International Nuclear Information System (INIS)

1994-01-01

This conference is composed of 23 papers, grouped in 3 sessions which main themes are: analysis of corrosion and erosion damages of nuclear power plant equipment and influence of water chemistry, temperature, irradiations, metallurgical and electrochemical factors, flow assisted cracking, stress cracking; monitoring and control of erosion and corrosion in nuclear power plants; susceptibility of structural materials to erosion and corrosion and ways to improve the resistance of materials, steels, coatings, etc. to erosion, corrosion and cracking

Comparative analysis of a fusion reactor blanket in cylindrical and toroidal geometry using Monte Carlo

International Nuclear Information System (INIS)

Chapin, D.L.

1976-03-01

Differences in neutron fluxes and nuclear reaction rates in a noncircular fusion reactor blanket when analyzed in cylindrical and toroidal geometry are studied using Monte Carlo. The investigation consists of three phases--a one-dimensional calculation using a circular approximation to a hexagonal shaped blanket; a two-dimensional calculation of a hexagonal blanket in an infinite cylinder; and a three-dimensional calculation of the blanket in tori of aspect ratios 3 and 5. The total blanket reaction rate in the two-dimensional model is found to be in good agreement with the circular model. The toroidal calculations reveal large variations in reaction rates at different blanket locations as compared to the hexagonal cylinder model, although the total reaction rate is nearly the same for both models. It is shown that the local perturbations in the toroidal blanket are due mainly to volumetric effects, and can be predicted by modifying the results of the infinite cylinder calculation by simple volume factors dependent on the blanket location and the torus major radius

Slope stability and erosion control: Ecotechnological solutions

NARCIS (Netherlands)

Norris, J.E.; Stokes, A.; Mickovski, S.B.; Cammeraat, E.; van Beek, R.; Nicoll, B.C.; Achim, A.

2008-01-01

This book is designed to assist the civil and geotechnical engineer, geomorphologist, forester, landscape architect or ecologist in choosing ecotechnological solutions for slopes that are prone to a variety of mass movements e.g. shallow failure or erosion. Within this book, the 'engineer' is used

Contour hedgerows and grass strips in erosion and runoff control in semi-arid Kenya

NARCIS (Netherlands)

Kinama, J.M.; Stigter, C.J.; Ong, C.K.; Ng'ang'a, J.K.; Gichuki, F.N.

2007-01-01

Most early alley cropping studies in semi-arid Kenya were on fairly flat land while there is an increase in cultivated sloping land. The effectiveness of aging contour hedgerows and grass strips for erosion control on an about 15% slope of an Alfisol was compared. The five treatments were Senna

Escoamento superficial na interação: cobertura vegetal e práticas de controle de erosão Erosion losses from runoff: interaction of soil cover and erosion control practice

Directory of Open Access Journals (Sweden)

Marco A. R. de Carvalho

2012-12-01

Full Text Available O escoamento da água oriunda das terras agricultadas é o principal fator poluente dos mananciais hídricos nas áreas rurais. Devido a esse fato, faz-se necessário o desenvolvimento e a aplicação de tecnologias que venham a reduzir descargas de resíduos indesejáveis. Nesse sentido, conduziu-se um experimento na área experimental do Departamento de Engenharia Rural - ESALQ/USP, Piracicaba - SP, com o objetivo de avaliar o efeito de diferentes condições de solo, (feijão, gramínea e solo nu e diferentes práticas de controle de erosão (sulco de infiltração, terraço de infiltração e sem práticas de controle de erosão, buscando-se estimar o escoamento superficial. O delineamento estatístico adotado foi o em blocos aleatorizados, em esquema fatorial 3x3, perfazendo 9 tratamentos com 3 repetições. O período de coleta de dados pluviométricos foi de 06 de dezembro de 2007 a 11 de abril de 2008; para isto, utilizou-se de um pluviômetro, com 21,1 cm de diâmetro, instalado na área experimental. Observando-se as perdas de água, em relação às estruturas, tem-se em ordem decrescente de eficiência: Terraço, Sulco e Rampa; e com relação às coberturas, tem-se em ordem decrescente de eficiência: Feijão, Capim e Solo Nu.The flow of sediment from cropped land is the main pollutant of water sources in rural areas. Due to this fact, it is necessary to develop and implement technologies that will reduce water and sediment discharges. Accordingly, an experiment was conducted in the Department of Biosystems Engineering - ESALQ / USP, Piracicaba - SP with the objective to evaluate the effect of different soil cover (bean, grass and bare ground and erosion control practices (wide base terraces and infiltration furrows in slopes (no practices to control erosion while measuring water losses in runoff. The statistical design adopted was randomized blocks in a 3x3 factorial scheme resulting in 9 treatments with 3 replicates (blocks. The

Dental Erosion in Children with Gastroesophageal Reflux Disease.

Science.gov (United States)

De Oliveira, Patricia Alves Drummond; Paiva, Saul Martins; De Abreu, Mauro Henrique Nogueira Guimarães; Auad, Sheyla Márcia

2016-01-01

The purpose of this study was to investigate the impact of gastroesophageal reflux disease (GERD) on dental erosion (DE) in children and analyze the association between dental erosion and diet, oral hygiene, and sociodemographic characteristics. This case-control study encompassed 43 two- to 14-year-olds diagnosed positive for GERD by the 24-hour pH monitoring, paired by age group with 136 healthy controls, in Belo Horizonte, Minas Gerais, Brazil. DE was assessed by one calibrated examiner using the O'Sullivan index. A questionnaire was self-administered by parents collecting information regarding sociodemographics, oral hygiene, and dietary habits. Dental erosion experience was compared between the groups, and a stratified analysis was performed (PDental erosion was diagnosed in 10.6 percent (N equals 19) of all the children; 25.6 percent (N equals 11) of GERD children and 5.9 percent (N equals eight) of children without GERD, P=0.001). Dental erosion was not associated with dietary consumption or sociodemographic characteristics in both groups (P≥0.05). Children who used adult toothpaste had a 5.79 higher chance of having dental erosion in the group with GERD. Children diagnosed with gastroesophageal reflux disease were at an increased risk of having dental erosion when compared to healthy subjects; among the GERD children, dental erosion was associated with the use of adult toothpaste.

Technical evaluation of major candidate blanket systems for fusion power reactor

International Nuclear Information System (INIS)

Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

1987-03-01

The key functions required for tritium breeding blankets for a fusion power reactor are: (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li 2 O/H 2 O/Be, Mo-alloy/Li 2 O/He/Be, Mo-alloy/LiAlO 2 /He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. (author)

Liquid metal flows in insulating elements of self-cooled blankets

International Nuclear Information System (INIS)

Molokov, S.

1995-01-01

Liquid metal flows in insulating rectangular ducts in strong magnetic fields are considered with reference to poloidal concepts of self-cooled blankets. Although the major part of the flow in poloidal blanket concepts is close to being fully developed, manifolds, expansions, contractions, elbows, etc., which are necessary elements in blanket designs, cause three-dimensional effects. The present investigation demonstrates the flow pattern in basic insulating geometries for actual and more advanced liquid metal blanket concepts and discusses the ways to avoid pressure losses caused by flow redistribution. Flows in several geometries, such as symmetric and non-symmetric 180 turns with and without manifolds, sharp and linear expansions with and without manifolds, etc., have been considered. They demonstrate the attractiveness of poloidal concepts of liquid metal blankets, since they guarantee uniform conditions for heat transfer. If changes in the duct cross-section occur in the plane perpendicular to the magnetic field (ideally a coolant should always flow in the radial-poloidal plane), the disturbances are local and the slug velocity profile is reached roughly at a distance equivalent to one duct width from the manifolds, expansions, etc. The effects of inertia in these flows are unimportant for the determination of the pressure drop and velocity profiles in the core of the flow but may favour heat transfer characteristics via instabilities and strongly anisotropic turbulence. (orig.)

Corrective measures technology for shallow land burial at arid sites: field studies of biointrusion barriers and erosion control

International Nuclear Information System (INIS)

Nyhan, J.W.; Hakonson, T.E.; Lopez, E.A.

1986-03-01

The field research program involving corrective measures technologies for arid shallow land burial (SLB) sites is described. Results of field testing of a biointrusion barrier installed at a close-out waste disposal site (Area B) at Los Alamos are presented. Soil erosion and infiltration of water into a simulated trench cap with various surface treatments were measured, and the interaction between erosion control and subsurface water dynamics is discussed relative to waste management

Direct LiT Electrolysis in a Metallic Fusion Blanket

Energy Technology Data Exchange (ETDEWEB)

Olson, Luke [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-30

A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

Direct LiT Electrolysis in a Metallic Fusion Blanket

International Nuclear Information System (INIS)

Olson, Luke

2016-01-01

A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

European research and development programme for water-cooled lithium-lead blankets: present status and future work

International Nuclear Information System (INIS)

Giancarli, L.; Leroy, P.; Proust, E.; Raepsaet, X.

1992-01-01

The European R and D programme in support of the development of water-cooled Pb-17Li blankets for DEMO aims at improving the data base concerning tritium behaviour and compatibility between blanket materials. The four main areas of the experimental programme are structural material corrosion by Pb-17Li, tritium extraction and permeation control.=, Pb-17Li physico-chemistry, and water/Pb-17Li interaction. This paper describes the most significant results obtained to date in the various experiments performed in Europe and the future programme required to complete the data base by 1994. 28 refs

A passively-safe fusion reactor blanket with helium coolant and steel structure

Energy Technology Data Exchange (ETDEWEB)

Crosswait, Kenneth Mitchell [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

1994-04-01

Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

Effects of fertile blanket on 600 MWth gas-cooled fast reactors: reactor and fuel cycle model

International Nuclear Information System (INIS)

Choi, Hang Bok

2002-07-01

A physics study has been performed to search for an optimum size of blanket for a 600 MWth gas-cooled fast reactor under fixed fuel and core specifications. The variables considered in this study are the reflector material, reflector thickness and blanket volume. The parametric calculations have shown that a positive breeding gain can be obtained by deploying 8 m 3 natural uranium blanket on the axial and radial boundaries of the core, surrounded by 40 cm Zr 3 Si 2 reflector. However the blanket core has disadvantages compared to the no-blanket core from the viewpoints of fuel fabrication cost and proliferation risk. On the other hand, the no-blanket core has large uncertainties in the possibility of achieving a positive breeding gain. Therefore further studies are recommended for the no-blanket option to improve the breeding gain and achieve a fissile self-sufficient fuel cycle, which is also proliferation-resistant. As an alternative, the blanket option can be considered, that ensures a positive breeding gain

ITER blanket module shield block design and analysis

International Nuclear Information System (INIS)

Mitin, D.; Khomyakov, S.; Razmerov, A.; Strebkov, Yu.

2008-01-01

This paper presents the alternative design of the shield block cooling path for a typical ITER blanket module with a predominantly sequential flow circuit. A number of serious disadvantages have been observed for the reference design, where the parallel flow circuit is used, which is inherent in the majority of blanket modules. The paper discusses these disadvantages and demonstrates the benefit of the alternative design based on the detailed design and the technological, hydraulic, thermal, structural and strength analyses, conducted for module no. 17

Nuclear Analysis of an ITER Blanket Module

Science.gov (United States)

Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

2013-08-01

ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

Energy Technology Data Exchange (ETDEWEB)

Takatsu, H; Enoeda, M [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1998-03-01

Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

Energy Technology Data Exchange (ETDEWEB)

Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

2016-11-01

Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

APT 3He target/blanket. Topical report

International Nuclear Information System (INIS)

1995-03-01

The 3 He target/blanket (T/B) preconceptual design for the 3/8-Goal facility is based on a 1000-MeV, 200-mA accelerator to produce a high-intensity proton beam that is expanded and then strikes one of two T/B modules. Each module consists of a centralized neutron source made of tungsten and lead, a proton beam backstop region made of zirconium and lead, and a moderator made of D 2 O. Helium-3 gas is circulated through the neutron source region and the blanket to create tritium through neutron capture. The gas is continually processed to extract the tritium with an online separation process

Impact of Soil Conservation Measures on Erosion Control and Soil Quality

International Nuclear Information System (INIS)

2011-10-01

This publication summarises the lessons learnt from a FAO/IAEA coordinated research project on the impact of soil conservation measures on erosion control and soil quality over a five-year period across a wide geographic area and range of environments. It demonstrates the new trends in the use of fallout radionuclide-based techniques as powerful tools to assess the effectiveness of soil conservation measures. As a comprehensive reference material it will support IAEA Member States in the use of these techniques to identify practices that can enhance sustainable agriculture and minimize land degradation.

Composite beryllium-ceramics breeder pin elements for a gas cooled solid blanket

International Nuclear Information System (INIS)

Carre, F.; Chevreau, G.; Gervaise, F.; Proust, E.

1986-06-01

Helium coolant have main advantages compared to water for solid blankets. But limitations exist too and the development of attractive helium cooled blankets based on breeder pin assemblies has been essentially made possible by the derivation from recent CEA neutronic studies of an optimized composite beryllium/ceramics breeder arrangement. Description of the proposed toroidal blanket layout for Net is made together with the analysis of its main performance. Merits of the considered composite Be/ceramics breeder elements are discussed

Applications of the aqueous self-cooled blanket (ASCB) concept to the Next European Torus (NET)

International Nuclear Information System (INIS)

Embrechts, M.J.; Bogaerts, W.; Cardella, A.; Chazalon, M.; Danner, W.; Dinner, P.; Libin, B.

1987-01-01

The Aqueous Self-Cooled Blanket Concept (ASCB) leads to a low-technology blanket design that relies on just structural material and coolant with small amounts of lithium compound dissolved in the coolant to provide for tritium production. The application of the ASCB concept in NET is being considered as a driver blanket that would operate at low temperature and low pressure and provide a reliable environment for machine operation during the technology phase. Shielding and tritium production are the primary objectives for such a low-technology blanket. Net tritium breeding is not a design requirement per se for a driver blanket for NET. A DEMO relevant ASCB based blanket test module with (local) tritium self-sufficiency and energy recovery as primary objectives might also be tested in NET if future developments confirm their viability

Anthropogenic Increase Of Soil Erosion In The Gangetic Plain Revealed By Geochemical Budget Of Erosion

Science.gov (United States)

Galy, V.; France-Lanord, C.; Galy, A.; Gaillardet, J.

2007-12-01

Tectonic and climatic factors are the key natural variables controlling the erosion through complex interactions. Nonetheless, over the last few hundred years, human activity also exerts a dominant control in response to extensive land use. The geochemical budget of erosion allows the balance between the different erosion processes to be quantified. The chemical composition of river sediment results from the chemical composition of the source rock modified by (1) weathering reactions occurring during erosion and (2) physical segregation during transport. If erosion is at steady state, the difference between the chemical composition of source rocks and that of river sediments must therefore be counterbalanced by the dissolved flux. However, climatic variations or anthropic impact can induce changes in the erosion distribution in a given basin resulting in non steady state erosion. Using a mass balance approach, the comparison of detailed geochemical data on river sediments with the current flux of dissolved elements allows the steady state hypothesis to be tested. In this study, we present a geochemical budget of weathering for the Ganga basin, one of the most densely populated basin in the world, based on detailed sampling of Himalayan rivers and of the Ganga in the delta. Sampling includes depth profile in the river, to assess the variability generated by transport processes. Himalayan river sediments are described by the dilution of an aluminous component (micas + clays + feldspars) by quartz. Ganga sediments on the other hand correspond to the mixing of bedload, similar to coarse Himalayan sediments, with an aluminous component highly depleted in alkaline elements. Compared with the dissolved flux, the depletion of alkaline elements in Ganga sediments shows that the alkaline weathering budget is imbalanced. This imbalance results from an overabundance of fine soil material in the Ganga sediment relative to other less weathered material directly derived from

Erosion control technology: a user's guide to the use of the Universal Soil Loss Equation at waste burial facilities

International Nuclear Information System (INIS)

Nyhan, J.W.; Lane, L.J.

1986-05-01

The Universal Soil Loss Equation (USLE) enables the operators of shallow land burial sites to predict the average rate of soil erosion for each feasible alternative combination of plant cover and land management practices in association with a specified soil type, rainfall pattern, and topography. The equation groups the numerous parameters that influence erosion rate under six major factors, whose site-specific values can be expressed numerically. Over a half century of erosion research in the agricultural community has supplied information from which approximate USLE factor values can be obtained for shallow land burial sites throughout the United States. Tables and charts presented in this report make this information readily available for field use. Extensions and limitations of the USLE to shallow land burial systems in the West are discussed, followed by a detailed description of the erosion plot research performed by the nuclear waste management community at Los Alamos, New Mexico. Example applications of the USLE at shallow land burial sites are described, and recommendations for applications of these erosion control technologies are discussed

Blanket/first wall challenges and required R&D on the pathway to DEMO

International Nuclear Information System (INIS)

Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

2015-01-01

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

Blanket/first wall challenges and required R&D on the pathway to DEMO

Energy Technology Data Exchange (ETDEWEB)

Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

2015-11-15

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

The EC conceptual design proposal of a water-cooled convertible blanket for ITER

International Nuclear Information System (INIS)

Giancarli, L.; Proust, E.; Baraer, L.; Bielak, B.; Raepsaet, X.; Salavy, J.F.; Sedano, L.; Szczepanski, J.; Quintric-Bossy, J.; Severi, Y.

1993-01-01

For several years the EC laboratories have developed breeding blankets for DEMO. From this experience, it has been derived a proposal of tritium breeding blanket for the Extended Performance Phase (EPP) of ITER. The general basic ideas are the following: (i) the switch from the shielding blanket used during the BPP to the breeding blanket for the EPP should not require segments replacement ('convertible' blanket): (ii) its use should not have significant impact on the Basic Performance Phase (BPP); (iii) design and used materials should assure good safety standards and acceptable public perception; (iv) the blanket coolant should be compatible with the coolant required in the high heat-flux components (e.g. divertor, etc.; (v) the required R and D should fit with the ITER time schedule; (vi) the blanket should be able to withstand large power excursions and to accept long downtimes. The proposed design consists of a water-cooled liquid metal blanket, using the eutectic Pb-17Li during the EPP and a non-breeding Pb-alloy (Pb-18Mg or Pb-50Bi) during the BPP. Each segment is basically formed by a box containing the alloy, cooled by an array of poloidal hairpin-type cooling tubes and reinforced by toroidal and radial stiffeners. The coolant tubes are double-walled tubes allowing leak detections. The selected First Wall (FW) is a toroidally-drilled steel plate with brazed water-cooling U-tube. The structural material is austenitic stainless steel (316L(N)) which limits the maximum acceptable neutron fluence to about 1 MWa/m 2 . The advantages of using other structural materials requiring longer leadtimes, such as ferritic/martensitic steels, are also briefly discussed

High temperature blankets for the production of synthetic fuels

International Nuclear Information System (INIS)

Powell, J.R.; Steinberg, M.; Fillo, J.; Makowitz, H.

1977-01-01

The application of very high temperature blankets to improved efficiency of electric power generation and production of H 2 and H 2 based synthetic fuels is described. The blanket modules have a low temperature (300 to 400 0 C) structure (SS, V, Al, etc.) which serves as the vacuum/coolant pressure boundary, and a hot (>1000 0 C) thermally insulated interior. Approximately 50 to 70% of the fusion energy is deposited in the hot interior because of deep penetration by high energy neutrons. Separate coolant circuits are used for the two temperature zones: water for the low temperature structure, and steam or He for the hot interior. Electric generation efficiencies of approximately 60% and H 2 production efficiencies of approximately 50 to 70%, depending on design, are projected for fusion reactors using these high temperature blankets

Wind erosion modelling in a Sahelian environment

NARCIS (Netherlands)

Faye-Visser, S.M.; Sterk, G.; Karssenberg, D.

2005-01-01

In the Sahel field observations of wind-blown mass transport often show considerable spatial variation related to the spatial variation of the wind erosion controlling parameters, e.g. soil crust and vegetation cover. A model, used to predict spatial variation in wind erosion and deposition is a

Conceptual design of ITER shielding blanket

International Nuclear Information System (INIS)

Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

1995-03-01

The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

Reduction of surface erosion in fusion reactors

International Nuclear Information System (INIS)

Rossing, T.D.; Das, S.K.; Kaminsky, M.

1976-01-01

Some of the major processes leading to surface erosion in fusion reactors are reviewed briefly, including blistering by implanted gas, sputtering by ions, atoms, and neutrons, and vaporization by local heating. Surface erosion affects the structural integrity and limits the lifetime of reactor components exposed to plasma radiation. In addition, some of the processes leading to surface erosion also cause the release of plasma contaminants. Methods proposed to reduce surface erosion have included control of surface temperature, selection of materials with a favorable microstructure, chemical and mechanical treatment of surfaces, and employment of protective surface coatings, wall liners, and divertors. The advantages and disadvantages of some of these methods are discussed

Overview of the TFTR Lithium Blanket Module program

International Nuclear Information System (INIS)

Jassby, D.L.

1986-01-01

The LBM (Lithium Blanket Module) is an approximately cubic module, about 80 cm on each side, with construction representative of a helium-cooled lithium oxide fusion reactor blanket module. Measurements of neutron transport and tritium breeding in the LBM will be made in irradiation programs first with a point-neutron source, and subsequently with the D-D and D-T fusion-neutron sources of the TFTR. This paper summarizes the objectives of the LBM program, the design, development and construction of the LBM, and progress in the experimental tests

Conceptual study on high performance blanket in a spherical tokamak fusion-driven transmuter

International Nuclear Information System (INIS)

Chen Yixue; Wu Yican

2000-01-01

A preliminary conceptual design on high performance dual-cooled blanket of fusion-driven transmuter is presented based on neutronic calculation. The dual-cooled system has some attractive advantages when utilized in transmutation of HLW (High Level Wastes). The calculation results show that this kind of blanket could safely transmute about 6 ton minor actinides (produced by 170 GW(e) Year PWRs approximately) and 0.4 ton fission products per year, and output 12 GW thermal power. In addition, the variation of power and critical factor of this blanket is relatively little during its 1-year operation period. This blanket is also tritium self-sustainable

Improved structure and long-life blanket concepts for heliotron reactors

International Nuclear Information System (INIS)

Sagara, A.; Imagawa, S.; Mitarai, O.

2005-01-01

New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14m is selected to permit a blanket-shield thickness of about 1m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R and D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated. (author)

Improved structure and long-life blanket concepts for heliotron reactors

Science.gov (United States)

Sagara, A.; Imagawa, S.; Mitarai, O.; Dolan, T.; Tanaka, T.; Kubota, Y.; Yamazaki, K.; Watanabe, K. Y.; Mizuguchi, N.; Muroga, T.; Noda, N.; Kaneko, O.; Yamada, H.; Ohyabu, N.; Uda, T.; Komori, A.; Sudo, S.; Motojima, O.

2005-04-01

New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

International Nuclear Information System (INIS)

Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

2004-01-01

RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

Evaluation of organic moderator/coolants for fusion breeder blankets

International Nuclear Information System (INIS)

Romero, J.B.

1980-03-01

Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

Blanket design for imploding liner systems

International Nuclear Information System (INIS)

Schaffer, M. J.

1980-01-01

The blanket design comprises hot, molten, rotating liquid vortex systems suitable for rapidly compressing confined plasmas, in which stratified immiscible liquid layers having successively greater mass densities outwardly of the axis of rotation are provided

Evaluation of Mediterranean plants for controlling gully erosion

International Nuclear Information System (INIS)

Baets, S. de; Poesen, J.; Muys, B.

2009-01-01

In Mediterranean environments, gullies are responsible for large soil losses causing loss of fertile cropland soil, reservoir sedimentation and flooding. To limit soil loss and sediment export it is important to prevent the initiation or rills and to stabilise gullies. This can be done by establishing vegetation at vulnerable places in the landscape. Although in the past, the effects of vegetation on soil erosion rates were predicted using above-ground biomass characteristics only, plant roots also play an important role in protecting the soil against erosion by concentrated runoff. Especially in conditions where the above-ground biomass becomes very scarce (e.g. due to drought, harvest, overgrazing or fire) the effects of vegetation will be underestimated when only above-ground plant characteristics are taken into account. (Author) 6 refs.

Evaluation of Mediterranean plants for controlling gully erosion

Energy Technology Data Exchange (ETDEWEB)

Baets, S. de; Poesen, J.; Muys, B.

2009-07-01

In Mediterranean environments, gullies are responsible for large soil losses causing loss of fertile cropland soil, reservoir sedimentation and flooding. To limit soil loss and sediment export it is important to prevent the initiation or rills and to stabilise gullies. This can be done by establishing vegetation at vulnerable places in the landscape. Although in the past, the effects of vegetation on soil erosion rates were predicted using above-ground biomass characteristics only, plant roots also play an important role in protecting the soil against erosion by concentrated runoff. Especially in conditions where the above-ground biomass becomes very scarce (e.g. due to drought, harvest, overgrazing or fire) the effects of vegetation will be underestimated when only above-ground plant characteristics are taken into account. (Author) 6 refs.

Flibe blanket concept for transmuting transuranic elements and long lived fission products

International Nuclear Information System (INIS)

Gohar, Y.

2000-01-01

A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

Overview of the Last Progresses for the European Test Blanket Modules Projects

International Nuclear Information System (INIS)

Salavy, J.-F.; Rampal, G.; Boccaccini, L.V.; Meyder, R.; Neuberger, H.; Laesser, R.; Poitevin, Y.; Zmitko, M.; Rigal, E.

2006-01-01

The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R-and-D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

A terminological matter: paragenesis, antigravitative erosion or antigravitational erosion ?

Directory of Open Access Journals (Sweden)

Pasini G.

2009-07-01

Full Text Available In the speleological literature three terms are utilized to designate the “ascending erosion”: paragenesis (= paragénésis, coined in1968, antigravitative erosion (= erosione antigravitativa, coined in 1966 and antigravitational erosion (wrong English translation ofthe Italian term erosione antigravitativa, utilized later on. The term paragenesis should be abandoned because of the priority of theterm erosione antigravitativa - on the ground of the “law of priority” – and because of its ambiguous etimology. On the other hand,the term antigravitational erosion should be forsaken in favour of the term antigravitative erosion, given the meaning that the termsgravitation and gravity have in Physics. Therefore, to designate the phenomenon of the “ascending erosion” there would be nothingleft but the term antigravitative erosion.The antigravitative erosion process and its recognizability are illustrated.Examples of caves with evident antigravitative erosion phenomena, developed in different karstifiable rocks and in several partsof the world, are given.It is recalled that the antigravitative erosion is a phenomenon well-known since 1942 and widely proven and supported, and that it isrelatively easy – in many cases - to recognize the antigravitative origin of karstic passages.It is stressed that the antigravitative erosion is an important phenomenon, exclusive of the karstic caves and unique in nature.

Gastroesophageal reflux is not associated with dental erosion in children.

Science.gov (United States)

Wild, Yvette K; Heyman, Melvin B; Vittinghoff, Eric; Dalal, Deepal H; Wojcicki, Janet M; Clark, Ann L; Rechmann, Beate; Rechmann, Peter

2011-11-01

Dental erosion is a complication of gastroesophageal reflux (GER) in adults; in children, it is not clear if GER has a role in dental pathologic conditions. Dietary intake, oral hygiene, high bacterial load, and decreased salivary flow might contribute independently to GER development or dental erosion, but their potential involvement in dental erosion from GER is not understood. We investigated the prevalence of dental erosion among children with and without GER symptoms, and whether salivary flow rate or bacterial load contribute to location-specific dental erosion. We performed a cross-sectional study of 59 children (ages, 9-17 y) with symptoms of GER and 20 asymptomatic children (controls); all completed a questionnaire on dietary exposure. Permanent teeth were examined for erosion into dentin, erosion locations, and affected surfaces. The dentist was not aware of GER status, and the gastroenterologist was not aware of dental status. Stimulated salivary flow was measured and salivary bacterial load was calculated for total bacteria, Streptococcus mutans, and Lactobacilli. Controlling for age, dietary intake, and oral hygiene, there was no association between GER symptoms and dental erosion by tooth location or affected surface. Salivary flow did not correlate with GER symptoms or erosion. Erosion location and surface were independent of total bacteria and levels of Streptococcus mutans and Lactobacilli. Location-specific dental erosion is not associated with GER, salivary flow, or bacterial load. Prospective studies are required to determine the pathogenesis of GER-associated dental erosion and the relationship between dental caries to GER and dental erosion. Copyright © 2011 AGA Institute. Published by Elsevier Inc. All rights reserved.

Modeling the fluid/soil interface erosion in the Hole Erosion Test

Directory of Open Access Journals (Sweden)

Kissi B.

2012-07-01

Full Text Available Soil erosion is a complex phenomenon which yields at its final stage to insidious fluid leakages under the hydraulic infrastructures known as piping and which are the main cause of their rupture. The Hole Erosion Test is commonly used to quantify the rate of piping erosion. In this work, The Hole Erosion Test is modelled by using Fluent software package. The aim is to predict the erosion rate of soil during the hole erosion test. The renormalization group theory – based k–ε turbulence model equations are used. This modelling makes it possible describing the effect of the clay concentration in flowing water on erosion. Unlike the usual one dimensional models, the proposed modelling shows that erosion is not uniform erosion along the hole length. In particular, the concentration of clay is found to increase noticeably the erosion rate.

Proceedings of the fifteenth international workshop on ceramic breeder blanket interactions

International Nuclear Information System (INIS)

Tanigawa, Hisashi; Enoeda, Mikio

2010-03-01

This report is the Proceedings of 'the Fifteenth International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors. This workshop was held in Sapporo, Japan on 3-4, Sept. 2009. Twenty six participants from EU, Japan, India, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket development. By this workshop, advance of key technologies for solid breeder blanket development was shared among the participants. Also, desired direction of further investigation and development was recognized. The 20 of the presented papers are indexed individually. (J.P.N.)

Proceedings of the fifteenth international workshop on ceramic breeder blanket interactions

Energy Technology Data Exchange (ETDEWEB)

Tanigawa, Hisashi; Enoeda, Mikio [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki (Japan)

2010-03-15

This report is the Proceedings of 'the Fifteenth International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors. This workshop was held in Sapporo, Japan on 3-4, Sept. 2009. Twenty six participants from EU, Japan, India, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket development. By this workshop, advance of key technologies for solid breeder blanket development was shared among the participants. Also, desired direction of further investigation and development was recognized. The 20 of the presented papers are indexed individually. (J.P.N.)

Heating facility for blanket and performance test

Energy Technology Data Exchange (ETDEWEB)

Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

1999-03-01

A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

Recent designs for advanced fusion reactor blankets

International Nuclear Information System (INIS)

Sze, D.K.

1994-01-01

A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

International Nuclear Information System (INIS)

Catalan, J.P.; Ogando, F.; Sanz, J.; Palermo, I.; Veredas, G.; Gomez-Ros, J.M.; Sedano, L.

2011-01-01

A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO F US based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils and material damage (dpa, gas production) to estimate the operational life of the steel-made structural components in the blanket and vacuum vessel (VV). In order to optimize the shielding of the coils different combinations of water and steel have been considered for the gap of the VV. The used neutron source is based on an axi-symmetric 2D fusion reaction profile for the given plasma equilibrium configuration. MCNPX has been used for transport calculations and ACAB has been used to handle gas production and damage energy cross sections.

Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor

Directory of Open Access Journals (Sweden)

Bong Guen Hong

2018-02-01

Full Text Available A configuration of a fusion-driven transmutation reactor with a low aspect ratio tokamak-type neutron source was determined in a self-consistent manner by using coupled analysis of tokamak systems and neutron transport. We investigated the impact of blanket configuration on the characteristics of a fusion-driven transmutation reactor. It was shown that by merging the TRU burning blanket and tritium breeding blanket, which uses PbLi as the tritium breeding material and as coolant, effective transmutation is possible. The TRU transmutation capability can be improved with a reduced blanket thickness, and fast fluence at the first wall can be reduced.  Article History: Received: July 10th 2017; Received: Dec 17th 2017; Accepted: February 2nd 2018; Available online How to Cite This Article: Hong, B.G. (2018 Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor. International Journal of Renewable Energy Development, 7(1, 65-70. https://doi.org/10.14710/ijred.7.1.65-70

The relative importance of different grass components in controlling runoff and erosion on a hillslope under simulated rainfall

Science.gov (United States)

Li, Changjia; Pan, Chengzhong

2018-03-01

The effects of vegetation cover on overland flow and erosion processes on hillslopes vary with vegetation type and spatial distribution and the different vegetation components, including the above- and below-ground biomass. However, few attempts have been made to quantify how these factors affect erosion processes. Field experimental plots (5 m × 2 m) with a slope of approximately 25° were constructed and simulated rainfall (60 mm hr-1) (Rainfall) and simulated rainfall combined with upslope overland flow (20 L min-1) (Rainfall + Flow) were applied. Three grass species were planted, specifically Astragalus adsurgens (A. adsurgens), Medicago sativa (M. sativa) and Cosmos bipinnatus (C. bipinnatus). To isolate and quantify the relative contributions of the above-ground grass parts (stems, litter cover and leaves) and the roots to reducing surface runoff and erosion, each of the three grass species was subjected to three treatments: intact grass control (IG), no litter or leaves (only the grass stems and roots were reserved) (NLL), and only roots remaining (OR). The results showed that planting grass significantly reduced overland flow rate and velocity and sediment yield, and the mean reductions were 21.8%, 29.1% and 67.1%, respectively. M. sativa performed the best in controlling water and soil losses due to its thick canopy and dense, fine roots. Grasses reduced soil erosion mainly during the early stage of overland flow generation. The above-ground grass parts primarily contributed to reducing overland flow rate and velocity, with mean relative contributions of 64% and 86%, respectively. The roots played a predominant role in reducing soil erosion, with mean contribution of 84%. Due to the impact of upslope inflow, overland flow rate and velocity and sediment yield increased under the Rainfall + Flow conditions. The results suggest that grass species on downslope parts of semi-arid hillslopes performed better in reducing water and soil losses. This study is

Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

International Nuclear Information System (INIS)

Beller, D.E.; Ott, K.O.; Terry, W.K.

1987-01-01

A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

HIP technologies for fusion reactor blankets fabrication

International Nuclear Information System (INIS)

Le Marois, G.; Federzoni, L.; Bucci, P.; Revirand, P.

2000-01-01

The benefit of HIP techniques applied to the fabrication of fusion internal components for higher performances, reliability and cost savings are emphasized. To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA. A coiled tube concept that allows cooling arrangement flexibility, strong reduction of the machining and number of welds is proposed for ITER IAM. Medium size mock-ups according to the WCLL breeding blanket concept have been manufactured. The fabrication of a large size mock-up is under progress. These activities are supported by numerical calculations to predict the deformations of the parts during HIP'ing. Finally, several HIP techniques issues have been identified and are discussed

Tokamak blanket design study, final report

International Nuclear Information System (INIS)

1980-08-01

A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m 2 and a particle heat flux of 1 MW/m 2 . Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

Tokamak blanket design study, final report

Energy Technology Data Exchange (ETDEWEB)

1980-08-01

A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

Thermalhydraulics of flowing particle-bed-type fusion reactor blankets

International Nuclear Information System (INIS)

Nietert, R.E.; Abdelk-Khalik, S.I.

1982-01-01

An experimental investigation has been conducted to determine the heat transfer characteristics of gravity-flowing particle beds using a special heat transfer loop. Glass microspheres were allowed to flow by gravity at controlled rates through an electrically heated stainless steel tubular test section. Values of the local and average convective heat transfer coefficient as a function of the average bed velocity, particle size and heat flux were determined. Such information is necessary for the design of gravity-flowing particle-bed type fusion reactor-blankets and associated tritium recovery systems. (orig.)

Required momentum, heat, and mass transport experiments for liquid-metal blankets

International Nuclear Information System (INIS)

Tillack, M.S.; Sze, D.K.; Abdou, M.A.

1986-01-01

Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

Neutronic optimization of a LiAlO2 solid breeder blanket

International Nuclear Information System (INIS)

Levin, P.; Ghoniem, N.M.

1986-02-01

In this report, a pressurized lobular blanket configuration is neutronically optimized. Among the features of this blanket configuration are the use of beryllium and LiAlO 2 solid breeder pins in a cross-flow configuration in a helium coolant. One-dimensional neutronic optimization calculations are performed to maximize the tritium breeding ratio (TER). The procedure involves spatial allocations of Be, LiAlO 2 , 9-C (ferritic steel), and He; in such a way as to maximize the TBR subject to several material, engineering and geometrical constraints. A TBR of 1.17 is achieved for a relatively thin blanket (approx. = 43 cm depth), and consistency with all imposed constraints

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 1

International Nuclear Information System (INIS)

Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

1991-10-01

The BOT (Breeder Outside Tube) Helium Cooled Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test plankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

Thermal comfort and safety of cotton blankets warmed at 130°F and 200°F.

Science.gov (United States)

Kelly, Patricia A; Cooper, Susan K; Krogh, Mary L; Morse, Elizabeth C; Crandall, Craig G; Winslow, Elizabeth H; Balluck, Julie P

2013-12-01

In 2009, the ECRI Institute recommended warming cotton blankets in cabinets set at 130°F or less. However, there is limited research to support the use of this cabinet temperature. To measure skin temperatures and thermal comfort in healthy volunteers before and after application of blankets warmed in cabinets set at 130 and 200°F, respectively, and to determine the time-dependent cooling of cotton blankets after removal from warming cabinets set at the two temperatures. Prospective, comparative, descriptive. Participants (n = 20) received one or two blankets warmed in 130 or 200°F cabinets. First, skin temperatures were measured, and thermal comfort reports were obtained at fixed timed intervals. Second, blanket temperatures (n = 10) were measured at fixed intervals after removal from the cabinets. No skin temperatures approached levels reported in the literature that cause epidermal damage. Thermal comfort reports supported using blankets from the 200°F cabinet, and blankets lost heat quickly over time. We recommend warming cotton blankets in cabinets set at 200°F or less to improve thermal comfort without compromising patient safety. Copyright © 2013 American Society of PeriAnesthesia Nurses. Published by Elsevier Inc. All rights reserved.

On the conditions of existence of cold-blanket systems

International Nuclear Information System (INIS)

Lehnert, B.

1977-12-01

An extende analysis of the partially ionized boundary layer of a magnetized plasma has been performed, leading to the following results: (i) In a first approximation the ion density at the inner ''edge'' of the layer becomes related to the wall-near neutral gas density, in a way being independent of the spatial distribution of the ionization rate. (ii) The particle and momentum balance equations, and the associated impermeability condition of the plasma with respect to neutral gas penetration, are not sufficient to specify a cold-blanket state, but have to be combined with considerations of the heat blance. This leads to lower and upper power input limits, thus defining conditions for the existence of a cold-blanket state. At decreasing beta values , or increasing radiation losses, there are situations where such a state cannot exist at all. (iii) It should become possible to fulfill the cold-blanket conditions in full-scale reactors as well as in certain model experiments. Probably these conditions can also be satisfied in large tokamaks like JET, and by fast gas injection in devices such as Alcator, but not in medium-size tokamaks being operated at moderately high ion densities. (iv) A strong ''boundary layer stabilization'' mechanism due to the joint viscosity-resistivity-pressure effects is available under cold-blanket conditions. (author)

Integration of test modules in the main blanket and vacuum vessel design

International Nuclear Information System (INIS)

Nakahira, Masataka; Kurasawa, Toshimasa; Sato, Satoshi; Furuya, Kazuyuki; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

1995-07-01

Typical test modules for water-cooled and helium-cooled ceramic breeder blankets have been designed, and their major design parameters are summarized. Among various candidates studied in Japan at present, BOT (Breeder Out of Tube) type of blanket is exemplified here. The integration scheme of the test module into ITER basic machine is also shown. Even with other type of blanket, the integration scheme won't be affected. The composition and space requirement of cooling and tritium recovery systems for the test module have also been studied. (author)

First wall and blanket design for the STARFIRE commercial tokamak power reactor

International Nuclear Information System (INIS)

Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

1979-01-01

The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

International Nuclear Information System (INIS)

Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

1982-01-01

Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

International Nuclear Information System (INIS)

Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

1980-09-01

The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design

An empirical approach to estimate soil erosion risk in Spain.

Science.gov (United States)

Martín-Fernández, Luis; Martínez-Núñez, Margarita

2011-08-01

Soil erosion is one of the most important factors in land degradation and influences desertification worldwide. In 2001, the Spanish Ministry of the Environment launched the 'National Inventory of Soil Erosion (INES) 2002-2012' to study the process of soil erosion in Spain. The aim of the current article is to assess the usefulness of this National Inventory as an instrument of control, measurement and monitoring of soil erosion in Spain. The methodology and main features of this National Inventory are described in detail. The results achieved as of the end of May 2010 are presented, together with an explanation of the utility of the Inventory as a tool for planning forest hydrologic restoration, soil protection, erosion control, and protection against desertification. Finally, the authors make a comparative analysis of similar initiatives for assessing soil erosion in other countries at the national and European levels. Copyright © 2011 Elsevier B.V. All rights reserved.

Ceramic BOT type blanket with poloidal helium cooling

International Nuclear Information System (INIS)

Cardella, A.; Daenenr, W.; Iseli, M.; Ferrari, M.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.

1989-01-01

This paper briefly describes the work done and results achieved over the past two years on the ceramic breeder BOT blanket with poloidal helium cooling. A conclusive remark on the brick/plate option described previously is followed by short descriptions of the low and high performance pebble bed options elaborated as alternatives for both NET and DEMO. The results show, togethre with those about the poloidal cooling of the First Wall, good prospects for this blanket type provided that the questions connected wiht an extensive use of beryllium find a satisfactor answer. (author). 5 refs.; 7 figs.; 1 tab

Ceramic sphere-pac breeder design for fusion blankets

International Nuclear Information System (INIS)

Gierszewski, P.J.; Sullivan, J.D.

1991-01-01

Randomly packed beds of ceramic spheres are a practical approach to surrounding fusion plasmas with tritium-breeding material. This paper examines the general properties of sphere-pac beds for application in fusion breeder blankets. The design considerations and models are reviewed for packing, tritium breeding and recovery, thermal conductivity, purge-gas pressure drop, mechanical behavior and fabrication. The design correlations are compared against available fusion ceramic data. Specific conclusions are that ternary (three-size) beds are not attractive for fusion blankets, and that the fusion spheres should be as large as possible subject primarily to packing constraints. (orig.)

Lead cooled heterogeneous accelerator driven molten-fluoride blanket for incineration of long-lived radioactive wastes

International Nuclear Information System (INIS)

Lopatkin, A.V.; Matyushechkin, V.M.; Tretyakov, I.T.; Blagovolin, P.P.; Kazaritsky, V.D.

1997-01-01

This paper presents a tentative design description and evaluation of the basic parameters of a lead cooled heterogeneous accelerator driven molten fluoride blanket. The proton beam of a 1 GeV accelerator strikes the blanket from below and generates spallation neutrons in the flow of lead, which serves as a target. These neutrons leave the target zone and get into a heterogeneous blanket with separated volumes of molten salts and lead. Fissile materials are dissolved in the salt. On getting into the molten salt volume the neutrons cause fission (transmutation) of the actinides, the produced heat being removed by circulation of molten lead. Two versions of the blanket design are examined. The first version: molten salt circulates in the fuel channels, while lead cools the channels flowing through the interchannel space (the salt channel design). The second version: it is lead that circulates in the channels, while molten salt takes up the interchannel space (the lead channel design). A preliminary blanket design study showed that both blanket designs possess a potential for improving performance. At present time the blanket design, mentioned above as the salt channel design, seems to be more promising. 1 ref., 2 figs., 2 tabs

Helium Loop for the HCPB Test Blanket Module

International Nuclear Information System (INIS)

Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

2006-01-01

In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

Numerical modelling of concentrated leak erosion during Hole Erosion Tests

OpenAIRE

Mercier, F.; Bonelli, S.; Golay, F.; Anselmet, F.; Philippe, P.; Borghi, R.

2015-01-01

This study focuses on the numerical modelling of concentrated leak erosion of a cohesive soil by a turbulent flow in axisymmetrical geometry, with application to the Hole Erosion Test (HET). The numerical model is based on adaptive remeshing of the water/soil interface to ensure accurate description of the mechanical phenomena occurring near the soil/water interface. The erosion law governing the interface motion is based on two erosion parameters: the critical shear stress and the erosion co...

Effect of blanket assembly shuffling on LMR neutronic performance

International Nuclear Information System (INIS)

Khalil, H.; Fujita, E.K.

1987-01-01

Neutronic analyses of advanced liquid-metal reactors (LMRs) have generally been performed with assemblies in different batches scatter-loaded but not shuffled among the core lattice positions between cycles. While this refueling approach minimizes refueling time, significant improvements in thermal performance are believed to be achievable by blanket assembly shuffling. These improvements, attributable to mitigation of the early-life overcooling of the blankets, include reductions in peak clad temperatures and in the temperature gradients responsible for thermal striping. Here the authors summarize results of a study performed to: (1) assess whether the anticipated gains in thermal performance can be realized without sacrificing core neutronic performance, particularly the burnup reactivity swing rho/sub bu/, which determines the rod ejection worth; (2) determine the effect of various blanket shuffling operations on reactor performance; and (3) determine whether shuffling strategies developed for an equilibrium (plutonium-fueled) core can be applied during the transition from an initial uranium-fueled core as is being considered in the US advanced LMR program

Study on compact design of remote handling equipment for ITER blanket maintenance

International Nuclear Information System (INIS)

Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi

2006-03-01

In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

Investigation of heat treatment conditions of structural material for blanket fabrication process

International Nuclear Information System (INIS)

Hirose, Takanori; Suzuki, Satoshi; Akiba, Masato; Shiba, Kiyoyuki; Sawai, Tomotsugu; Jitsukawa, Shiro

2004-01-01

This paper presents recent results of thermal hysteresis effects on ceramic breeder blanket structural material. Reduced activation ferritic/martensitic (RAF) steel is the leading candidates for the first wall structural materials of breeding blankets. RAF steel demonstrates superior resistance to high dose neutron irradiation, because the steel has tempered martensite structure which contains the number of sink site for radiation defects. This microstructure obtained by two-step heat treatment, first is normalizing at temperature above 1200 K and the second is tempering at temperature below 1100 K. Recent study revealed the thermal hysteresis has significant impacts on the post-irradiation mechanical properties. The breeding blanket has complicated structure, which consists of tungsten armor and thin first wall with cooling pipe. The blanket fabrication requires some high temperature joining processes. Especially hot isostatic pressing (HIP) is examined as a near-net-shape fabrication process for this structure. The process consists of heating above 1300 K and isostatic pressing at the pressure above 150 MPa followed by tempering. Moreover ceramics pebbles are packed into blanket module and the module is to be seamed by welding followed by post weld heat treatment in the final assemble process. Therefore the final microstructural features of RAFs strongly depend on the blanket fabrication process. The objective of this work is to evaluate the effects of thermal hysteresis corresponding to blanket fabrication process on RAFs microstructure in order to establish appropriate blanket fabrication process. Japanese RAFs F82H (Fe-0.1C-8Cr-2W-0.2V-0.05Ta) was investigated by metallurgical method after isochronal heat treatment up to 1473 K simulating high temperature bonding process. Although F82H showed significant grain growth after conventional solid HIP conditions (1313 K x 2 hr.), this coarse grained microstructure was refined by the post HIP normalizing at

Design and safety analysis of the helium cooled solid breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

2016-05-15

This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

International Nuclear Information System (INIS)

Davidson, J.W.; Battat, M.E.

1993-01-01

Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

Heat-pipe liquid-pool-blanket concept for the Tandem Mirror Reactor

International Nuclear Information System (INIS)

Hoffman, M.A.; Werner, R.W.; Johnson, G.L.

1981-01-01

The blanket concept for the tandem mirror reactor described in this paper was developed to produce the medium temperature heat (approx. 850 to 950 K) for the General Atomic sulfur-iodine thermochemical process for producing hydrogen. This medium temperature heat from the blanket constitutes about 81% of the total power output of the fusion reactor

Phase-IIC experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics

International Nuclear Information System (INIS)

Oyama, Yukio

1992-12-01

Neutronics experiments on two types of heterogeneous blankets have been performed as the Phase-IIC experiment of JAERI/USDOE collaborative program on fusion blanket neutronics. The experimental system was used in the same geometry as the previous Phase-IIA series which was a closed geometry using neutron source enclosure of lithium carbonate. The heterogeneous blankets selected here are the beryllium edge-on and the water coolant channel assemblies. In the former the beryllium and lithium-oxide layers are piled up alternately in the front part of test blanket. In the latter, the three simulated water cooling channels are settled in the Li 2 O blanket. These are producing steep gradient of neutron flux around material boundary. The calculation accuracy and measurement method for these features is a key of interest in the experiments. The measurements were performed for tritium production rate and the other nuclear parameters as well as the previous experiments. This report describes the experimental detail and the results enough to use for the benchmark data for testing the data and method of design calculation of fusion reactors. (author)

Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

International Nuclear Information System (INIS)

Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin

2016-01-01

Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).