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The first generation of TCAP hydrogen isotope separation process has been in service for tritium separation at the Savannah River Site since 1994. To prepare for replacement, a next-generation TCAP process has been developed. This new process simplifies the column design and reduces the equipment requirements of the thermal cycling system. An experimental twelve-meter column was fabricated and installed in the laboratory to demonstrate its performance. This new design and its initial test results were presented at the 8th International Conference on Tritium Science and Technology and published in the proceedings. We have since completed the startup and demonstration the separation of protium and deuterium in the experimental unit. The unit has been operated for more than 200 cycles. A feed of 25% deuterium in protium was separated into two streams each better than 99.7% purity.

Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.

The effect of the size and distribution of non-metal inclusions on mechanical properties of the China low activation martensitic steel (CLAM) was investigated. The tensile and Charpy V-notch impact tests showed that electroslag remelting improved the tensile properties and reduced the ductile brittle transition temperature (DBTT). Inclusion detection of CLAM by optical microscopy and scanning electron microscope showed that both the dimensions and quantities of the alumina inclusions decreased and their distributions became more uniform after remelting. The better refinement and distribution uniformity of alumina inclusions were considered as the main possible reasons for the improvement of the mechanical properties of CLAM steel after remelting.

A tritium flow model of the entire FDS-II blanket system was developed and the preliminary analysis on tritium permeation and extraction for FDS-II blanket system were done by using Tritium Analysis Software (TAS). The factors which affected tritium extraction and permeation were calculated and evaluated, such as tritium permeation reduction factor in blanket, proportion of LiPb flow in tritium extraction system and helium leakage rate, etc. The results of the presented analysis shows that further R&D efforts are still required to guarantee the tritium self-sufficient and safety, for example high quality tritium permeation barriers, efficiency of tritium extraction from LiPb and fabrication technology of the LiPb heat exchanger, etc.

In this study, the rejections of dye, salt, and dye intermediate by a cellulose acetate (CA) nanofiltration membrane were investigated. The rejections of sodium chloride and sodium sulfate were moderate and declined with the increase of concentration. The rejection of five reactive dyes by a CA nanofiltration membrane was higher than 99.0% at low concentration, and high rejection of five reactive dyes was still achievable at higher dye concentration. However, the rejections of five dye intermediates varied from low to moderate when the molecular weight increased. In the C.I. Reactive Black 5 diafiltration and concentration process, the concentration and constant volume diafiltration mode had a better desalting degree than the constant volume diafiltration and concentration mode. However, the constant volume diafiltration and concentration mode is necessary for C.I. Reactive Blue 13 solution due to its low solubility. This study illustrates that nanofiltration is a promising process for dye desalting and concentrating.

The liquid metal and solid ceramic pebble breeder blankets have become the most promising blankets for ITER-TBMs or DEMO reactors in China and the world due to their potential advantages. In recent years the corresponding research work on fusion reactor materials mainly focuses on structural materials, plasma facing materials and the functional materials for the blanket such as breeder, coating and flow channel insert etc. for the successful application of fusion energy in the near future. The R&D on those materials in the two kinds of blankets is being carried out widely in China, including fabrication and manufacturing techniques, physical/mechanical properties assessment before and after irradiation, joining techniques for structural materials, compatibility evaluation, and the development and verification of the criteria for fusion material designs. The progress on main R&D activities of fusion reactor materials in China is introduced and prospected in the paper.

Membrane process has increasingly developed as a reliable and effective means of improving product yield and reducing manufacturing costs in the reactive dye industry. In order to improve a product's quality, ultrafiltration (UF) membrane has been applied to perform Reactive Brilliant Blue KN-R desalting and concentration. The performance of this membrane's separation process was evaluated under different operating conditions, through which the influence of operating pressure, temperature, cross-flow velocity, pH, concentration of feed and operating time on permeate flux, rejection of Reactive Brilliant Blue KN-R and sodium sulfate were studied. When the permeate flux reached 63.2 L m −2 h −1 at ambient temperature, the rejection of Reactive Brilliant Blue KN-R was over 99.0% and the rejection of sodium sulfate was less than 50.0%. And operating pressure, temperature, cross-flow velocity and concentration of feed have posed significant influence on the membrane's separation performance, whereas pH and operating time had little effect on the separation performance. This study also illustrates that membrane process is a promising technique for Reactive Brilliant Blue KN-R desalting and concentration.

The liquid metal LiPb blanket design is one of the most promising designs for future fusion power reactors and under wide research in the world. The first liquid metal LiPb loop in China named DRAGON-I was built in 2005 in order to do research on characteristics of liquid metal LiPb such as its corrosion to structural materials of the blankets and so on. The first corrosion experiment in flowing LiPb with a speed of 0.08 m/s at 480 °C for 500 h was done in October 2005 on CLAM (China low activation martensitic) steel and 316L stainless steel for comparison. The weights and compositions, etc. of the specimens before and after corrosion experiment were tested and analyzed, the microstructures of the specimens were also inspected by SEM. The results show that the corrosion of CLAM steel is relatively slight, while that for 316L is obvious and very serious. Further study on corrosion behavior of CLAM for longer time experiment in liquid LiPb at different temperatures and flow speeds will be carried out in the near future.

China Low Activation Martensitic steel (CLAM), which is one of the RAFMs (Reduced Activation Ferritic/Martensitic steels) and under development in ASIPP, is considered as the primary candidate structural material and LiPb eutectic as both tritium breeder and coolant of the blankets in FDS series fusion reactors. The corrosion behavior of CLAM steel exposed to the liquid breeder LiPb is of significance. Corrosion tests of CLAM in flowing LiPb at 480°C were performed up to about 2000 hrs to analyze the corrosion mechanism of CLAM exposed to liquid LiPb. The specimens were observed and analyzed by Scanning Electron Microscopy (SEM) and Energy Dispersive X-ray Spectroscopy (EDX) after 500hrs, 1000hrs and 2000hrs corrosion experiment respectively. The corrosion is of non-uniform and the weight loss was about 0.23 mg/cm2 after 2000 hrs’ exposure, which is smaller for CLAM compared to those of other RAFMs.

The effects of yttrium on properties and microstructure for China Low Activation Martensitic (CLAM) steel have been investigated. The tensile test shows that the addition of 0.2% yttrium decreases the strength of CLAM steel before irradiation. The results obtained by positron annihilation lifetime measurements of the CLAM, F82H and T91 steels irradiated by 80 MeV 19 F ions to 10 dpa demonstrate that the irradiation resistant property of CLAM steel with yttrium is much better than that of other two steels. The SEM and TEM observations suggest the addition of yttrium both refines the martensitic lath structure and leads to the precipitation of big Y 2 Fe 17 C 0.75 compound.

Preliminary Hot Isostatic Pressing (HIP) diffusion bonding experiments on China Low Activation Martensitic (CLAM) steel have been performed. A few approaches such as dry turning, grinding and mechanical-polishing, etc. were used to prepare the samples’ surfaces. Different capsule sealing techniques were used as well. The HIP parameters are 140 MPa/3–4 h/1323–1423 K. Then post-HIP heat treatment (PHHT) was done. The tensile properties of HIPed joints of 1373 K and 1423 K with dry-turned or ground joining surface are identical to those of the base material. But the Charpy impact absorbed energy of the joints is a little low and further research is needed to increase it.

The tensile and impact properties of CLAM steel are compared to those of JLF-1 steel. Tensile testing revealed that the ultimate and yield strengths of the CLAM steel are 670 MPa and 512 MPa at room temperature, and 373 MPa and 327 MPa at 873 K, respectively. These values are higher than those measured for JLF-1. The ductile-to-brittle transition temperature (DBTT) of CLAM was found to be 171 K using one-third size Charpy V-notch specimens, which is 16 K lower than that of JLF-1. Microstructural analysis by SEM and TEM indicated that the prior austenite grain size and lath width for CLAM are smaller than those for JLF-1. The finer grain and lath structure is considered to be one of the main reasons for the higher strength and lower DBTT of the CLAM steel.

CREST is a conceptual tokamak reactor design with high β plasma, high thermal efficiency, competitive cost and water-cooled ferritic steel components. Some of its parameters are similar to those of the ITER advanced mode plasma. In this manuscript, the specific issues and analysis on damage to TF coils of CREST were carried out based on the three-dimensional model of the CREST with the widely used code MCNP/4C and the IAEA latest released FENDL/2.1 data library. Damage to some specific regions of the TF coils near large openings and at the inboard mid-plane are calculated and analyzed. Parameters such as the distributions of nuclear heat density, fast neutron flux, dose rate to the epoxy insulator, and peak displacement dose to Cu conductor for the TF coil near these regions were calculated and analyzed. The shield thicknesses at these regions are optimized.

A series of R&D activities on the structural material China Low Activation Martensitic steel (CLAM) and related blanket technology are being carried out in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). A summary of these activities is presented, mainly covering the composition design, property tests, techniques for HIP joining and coating, and activation analysis. In addition, a nuclear material database FUMDS is introduced, which is under development based on the requirement for CLAM data management.

Chinese low activation martensitic steel (CLAM) has been designed with improved composition, and its performance, such as tensile properties, ductile–brittle transition temperature (DBTT41J), creep and thermal physical properties, has been determined. The interaction experiments between CLAM and plasma were carried out in the HT-7 tokamak facility and the activities, afterheat and gamma dose rate for CLAM as a function of cooling time (CT) were calculated to obtain the required control levels of impurities in CLAM. The insulator coatings on CLAM steel prepared by the CVD process at 700 °C and 740 °C have pure Al2O3 and Al2O3 with an oxygen deficiency layer on surface about 1 μm thick. The electrical resistivity of the coating reaches about 104 Ωm2 on the surface.

Due to lack of high-resolution data both in space and time, mapping land use change in built-up areas remain challenging. In this work, we developed an approach that integrates multiple "geo-sensor" data sources, including remote sensing and social media, for mapping urban land use changes. The approach starts by mapping built-up expansion annually in small patches, using dense time periods of remotely sensed imagery. We refine these patches and identify the major categories of land use such as residential, commercial, manufacturing, recreational, and office. We further select one major category---residential---and use social media data and house trading records to classify it into three sub-categories: gated communities, ordinary communities, and urban slums. We demonstrate our approach using the multi-sourced data of Kunming, a medium-sized city in China. The results showed that our approach provides an efficient and integrative way for mapping large-scale land use change in urban regions with significant details that are not feasible only with limited remote sensing data.

Traditional GIS tools and systems are powerful for analyzing geographic information for various applications but they are not designed for processing dynamic streams of data. This paper presents a CyberGIS framework that can automatically synthesize multi-sourced data, such as social media and socioeconomic data, to track disaster events, to produce maps, and to perform spatial and statistical analysis for disaster management. Within our framework, Apache Hive, Hadoop, and Mahout are used as scalable distributed storage, computing environment and machine learning library to store, process and mine massive social media data. The proposed framework is capable of supporting big data analytics of multiple sources. A prototype is implemented and tested using the 2011 Hurricane Sandy as a case study.

Hurricane Sandy made landfall in one of the most populated areas of the United States, and affected almost 8 million people. The event provides a unique opportunity to study power outages because of the data available and the large impact to a densely populated area. Satellite nightlight imagery of "before" and "after" the landfall of the hurricane is used to quantify the light dimming caused by power outages. Geolocated tweets filtered by keywords provide valuable information on human activity at a high temporal and spatial resolution during the event. Analysis of brightness change in the satellite data and the density of power related tweets points to a spatial relationship that identifies severely impacted areas with human presence. Classification of tweets through text analysis serves to further narrow the information search to find the most relevant and reliable content. Twitter data fused with satellite imagery identifies power outage information at a street-level resolution that is not achievable with satellite imagery alone.

Since many years, liquid metal flows for applications in fusion blankets have been investigated worldwide. A review is given about modeling requirements and existing experimental facilities for investigations of liquid metal related issues in blankets with the focus on magnetohydrodynamics (MHD). Most of the performed theoretical and experimental works were dedicated to fundamental aspects of MHD flows under very strong magnetic fields as they may occur in generic elements of fusion blankets like pipes, ducts, bends, expansions and contractions. Those experiments are required to progressively validate numerical tools with the purpose of obtaining codes capable to predict MHD flows at fusion relevant parameters in complex blanket geometries, taking into account electrical and thermal coupling between fluid and structural materials. Scaled mock-up experiments support the theoretical activities and help deriving engineering correlations for cases which cannot be calculated with required accuracy up to now. © 2014 Karlsruhe Institute of Technology All rights reserved.

LiPb auxiliary system is an important external loop equipment of test blanket module (TBM) for liquid LiPb blanket design. According to the requirement of testing strategy of DFLL (dual functional lithium lead) TBM, which could demonstrate the technology of the SLL (single-coolant lithium lead) and the DLL (dual-coolant lithium lead) blankets, a conceptual design of the LiPb ancillary system is done, which could serve to achieve the dual functions. The auxiliary system can be used to achieve the recovery of tritium, heat rejection, monitoring of impurity in LiPb, LiPb purification and so on. It consists of detritiation unit, LiPb to He heat exchanger, plugging indicator, LiPb purification system and pumping system, etc.

The compact reversed shear tokamak CREST is a conceptual tokamak reactor design with high β plasma, high thermal efficiency, competitive cost and water-cooled ferritic steel components. In this manuscript neutronics analysis on CREST is presented based on the three-dimensional (3D) model and calculations by MCNP/4C code and FENDL/2.0 data library. The comparison among the results of tritium breeding ratio (TBR) for the one-dimensional (1D), two-dimensional (2D) and 3D calculations shows that the results are consistent with each other. The system would be tritium self-sufficient when considering the blanket coverage fraction, etc. and even some uncertainty. It has an energy multiplication of 1.38. The maximum neutron wall loading and maximum damage rate at the first wall (FW) are 6.30 MW/m2 and 51.9 dpa/FPY, respectively. The FW material needs to be replaced every 3.8 years of operation with an availability of 75% if the neutron damage of ferritic steel is assumed to limit to 150 dpa. The total nuclear heat for TF coils is under the limit for TF coils in ITER case. The nuclear heat density for TF coils at the mid-plane is roughly the same as the limit for ITER.

This paper reported the low activation martensitic steels which are being studied to develop the structural materials in fusion reactors. The steels were based on 9Cr1.5WVTa, but the effect of alloy elements was investigated by changing the amounts of alloy elements or adding other elements. The structure and properties of the steels were studied by tensile experiment, X-ray diffraction, SEM, TEM. Also the metallurgical process and heat treatment effect were discussed.

The Dual-cooled Waste Transmutation Blanket design of the Fusion-Driven Sub-critical System is a dual-cooled liquid metal Pb–17Li and He system with Chinese Low Activation Martensitic steel structure. The effects of the structural material on liquid metal MHD flow, including pressure drop and heat transfer, are presented. The influences of the wall conductance of the duct wall on the MHD pressure drop have been analysed and estimated. The pressure drop is too large to be accepted when Pb–17Li coolant flows in these ducts without insulating coating. This can be improved by using Al2O3 electrically insulating coating, while heat transfer must also be cautiously considered.

Reduced activation ferritic/martensitic steels (RAFM steels) are presently considered as the primary structural materials for a demonstration (DEMO) fusion plant and the first fusion power reactors because of their attractive properties. Studies on various properties of China low activation martensitic steel (CLAM) are underway. The activation level of CLAM steel was calculated with the widely used inventory code FISPACT with the latest data library FENDL/A-2 based on the first wall (FW) neutron spectrum of the fusion-driven subcritical system (FDS) from the Monte Carlo transport code MCNP/4C calculation with FENDL-2 data library. The results were compared with the activation levels of other RAFM steels, such as EUROFER97, F82H, JLF-1 and 9Cr–2WVTa etc., under the same irradiation conditions. Furthermore, the dominant nuclides to γ-ray dose rate of CLAM steel were analyzed. The required control levels of impurities in CLAM steel will soon be implemented based on the hands-on and remote recycling dose rate limits.

Aluminum based coatings on low activation martensitic (LAM) steels were formed by using CVD method at 973 and 1013 K. After coating, the cleaned specimens were oxidized in 0.133 Pa oxygen atmosphere at 973 and 1013 K to form the Al 2 O 3 layer on the surface. The typical coating thickness of the aluminized layer is 20 μm, a 10 μm thick external layer and a 10 μm thick internal layer. The thickness of the Al 2 O 3 layer is about 1 μm. The microstructure, micro-hardness profile, composition depth distribution in the cross section of the coating and the electrical resistivity of the layers were measured. The results indicate that the coatings on Chinese low activation martensitic (CLAM) steel have higher micro-hardness and higher density than that on EUROFER-97 steel produced at 1013 K. The Al 2 O 3 layer formed at 973 K has higher micro-hardness and higher density than that formed at 1013 K.

The superconducting tokamak HT-7U will perform D–D plasma experiments and the production rate of 2.45 MeV neutrons will achieve the level of 10 15 n/s. The dose rates during operation and after shutdown of HT-7U have been calculated based on a detailed three-dimensional model with the Monte Carlo code MCNP-4C, the inventory code FISPACT and the FENDL-2 nuclear data library. The so-called R2S method was used to perform the shutdown dose rate analyses. The results confirm that personnel access into the tokamak hall must be prohibited during operation. In addition the results show that the dose rate level near the outer wall of the tokamak hall slightly exceeds the dose limit. Additional shielding may be required. The shutdown dose rates of various experimental campaigns have been calculated and the waiting times for personnel access into the hall have been assessed. The results indicate that the in-vessel components of HT-7U will be considerably activated by neutrons and the shutdown dose rate level around the machine after long D–D plasma discharges is serious within a certain range of cooling time regarding the maintenance work.

In the past decade, the social sciences are undergoing a dramatic shift toward analyzing ever-increasing amounts of large-scale diverse data with rich spatial, temporal, and thematic resolution (Sh...

This paper mainly deals with stress field numerical calculations and analysis of the Dual-cooled waste transmutation blanket (DWTB). A comprehensive finite element method (FEM) structure analysis has been performed with the computer code ANSYS5.7. The analysis refers to the steady-state operating condition of an outboard blanket segment. Thermal stresses caused by the temperature gradients in the blanket structure have been calculated. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as input for a FEM stress analysis and the resulting stress distribution has been evaluated.

The HT-7U superconducting tokamak will generate neutrons at a rate of the order of 10 15 per second during the D–D operation with long pulse up to 1000 s. Nuclear heat in the Toroidal Field (TF) and Poloidal Field (PF) coils have been estimated by using the Monte Carlo particle transport code MCNP-4C and the latest version of the Fusion Evaluated Nuclear Data Library (FENDL-2) based on the three-dimensional geometrical configuration. The calculations show that the nuclear heat in the coils originates ∼10% from neutron energy deposition and ∼90% from induced gamma energy deposition. To reduce the nuclear heat in the superconducting magnets of the TF and PF coils, a double-wall-structured vacuum vessel (VV) with the boronated water layer of thickness of 5 cm is designed to shield against neutron irradiation.

The design of the radiation shield and safety system for the HT-7U fusion experimental device is presented in this contribution. An inner shield and an outer shield are considered in the shield design. Calculations and analyses have been done using 3D MCNP/4C models for the neutron spectrum and with the inventory code FISPACT for activation calculation in order to estimate the dose rate to workers after one pulse operation of the device. In addition, a computer system for radiation protection and control based on Controller Area Network has been developed in order to protect staff and publics from the radiation of neutrons and γ during and after the operation of the HT-7U device. The safety system has the advantages of multi-master protocol, real-time capability, error correction, long communication distance and high noise immunity etc.

The induced radioactivity of a few structural materials by D–T fusion neutrons and fission neutrons in a hybrid blanket have been calculated based on the fusion driven systems as the reference reactors. The neutron transport and activation calculations are done by the Monte Carlo transport code MCNP/4B and the activation inventory code FISPACT with the latest released IAEA Fusion Evaluated Nuclear Data Library (FENDL-2) and the ENDF/B-V uranium evaluated data. The effects of material types, impurities and energy spectra on the activation characteristics of materials are analyzed.

The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented.

A program for China low activation martensitic steel (CLAM) development has been underway since 2001 to satisfy the material requirements of the test blanket module (TBM) for ITER, China fusion engineering test reactor and China fusion demonstration reactor. It has been undertaken by the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences under wide domestic and international collaborations. Extensive work and efforts are being devoted to the R&D of CLAM, such as mechanical property evaluation before and after neutron irradiation, fabrication of scaled TBM by welding and additive manufacturing, improvement of its irradiation resistance as well as high temperature properties by precipitate strengthening to achieve its final successful application in fusion systems. The status and improvement of CLAM are introduced in this paper.

To investigate irradiation/helium effects on microstructure of China Low Activation Martensitic (CLAM) steel, single beam (e−) and dual-beam (e−/He+) irradiation to 17.5 dpa with 10 appm He/dpa at 400 °C and 450 °C on CLAM steel was carried out with high voltage electron microscope HVEM. Evolution of microstructure was in situ observed during irradiation and annealing procedure. Preliminary results showed that size and density of voids and helium bubbles increased with irradiation dose. And swellings of voids and bubbles are ∼0.5% and 1–1.5%, respectively after irradiation.

SiCf/SiC composites are considered as one of promising candidate blanket structural materials and functional materials for the flow channel insert of LiPb blanket for fusion reactors. One of key issues for the SiCf/SiC composites is its compatibility with LiPb. A variety of SiC materials prepared by different processes (i.e. CVD-SiC, PIP-SiC), coated SiCf/SiC composite (the coating containing Al2O3 and SiO2) and SiO2 were exposed in the static liquid LiPb at 800 °C for 200 h and 1000 h, respectively. The preliminary results showed that PIP-SiC had good compatibility with LiPb at 800 °C. The coating on the SiCf/SiC composite was corroded by liquid LiPb. In addition, the XRD pattern of CVD-SiC integrally moved left a little after 1000 h exposure.

Porosity-controlled nuclear fuel microsphere is an essential material of fabricating minor actinide-bearing dispersion-type nuclear fuel with the infiltration processes. In this paper, monodisperse and size-controlled spherical oxide nuclear fuel particles with size range of 20μm to 800μm were fabricated by means of microfluidic technology combined with sol-gel process using cerium as a surrogate for plutonium. The porous CeO2 beads with the density range of 25% to 93% T.D. were successfully prepared by the addition of polyethylene glycol 6000 used as a porogen to the feed broth. The uniform U3O8 beads were also prepared at the same experimental conditions as CeO2 beads prepared, which shows the feasibility of the method for fabricating size-controlled monodisperse nuclear fuel beads.

Liquid lead or lead-based alloy is a potential candidate coolant for fast reactors and ADS subcritical reactors because of its many unique nuclear, thermal-physical and chemical attributes. Chinese Academy of Sciences (CAS) had launched an engineering project to develop ADS system and lead-based reactors. Series CLEAR reactor conceptual designs were finished, and the preliminary engineering design for the China Lead-based research reactor (CLEAR-I) was underway. The key components prototypes for engineering validation including the control rod drive system, refueling system, fuel assembly have been constructed, the validation experiment are carrying out. KYLIN series PbBi experimental loops has already been built to perform structure material corrosion experiment, thermal-hydraulics experiment and safety experiment for CLEAR series reactors. The highly intensified neutron generator HINEG will be constructed to take the benchmark experiment of neutronics simulation codes. In this paper, the design and R&D progress are presented.Copyright © 2014 by ASME

Geographic phenomena evolve in a four-dimensional spatiotemporal world. To capture the geographical phenomena at different scales, large amount of data (big data) are produced with specific spatiotemporal patterns. Phenomena evolution and the principles driving the evolution provide pathways for developing methodological solutions to process the big spatiotemporal data. Based on experiences gained from several projects, such as climate studies and cloud computing, we introduce in this chapter modern computing technologies required for processing big data, including (1) sensor web, Earth observations, and model simulations for collecting and generating big data, (2) flexible and standard-based systems for managing big data for easy discovery and access, (3) multidimensional visual analytics for exploring and analyzing big spatiotemporal data, and (4) grid, cloud, and GPU computing for addressing the computing intensive challenges. We discuss through exemplar projects how these cutting-edge computing technologies are utilized to handle big spatiotemporal data. We expect this chapter to set a computing research context for future big data handling at different spatiotemporal granules.

Silicon carbide fiber reinforced silicon carbide composites (SiCf/SiC) are promising candidate materials for fusion applications. The compatibility of SiCf/SiC composites with liquid LiPb is one of the key issues for fusion technologies. Three-dimensional SiCf/SiC composites were fabricated by polymer infiltration and pyrolysis (PIP) process and β-SiC coating was deposited by chemical vapor deposition (CVD) process. The composites were exposed to static liquid LiPb for 500 h at 700 °C. The results showed that the liquid LiPb penetrated through the open pores of the cut surface of the composites and reacted with the matrix and fibers, which caused a weight change and a severe degradation of the mechanical property to the uncoated composites. As for the coated composites, the mechanical property was degraded and the weight increased because of the absorption of LiPb, although the CVD SiC coating was compatible with LiPb and protected fibers and matrix beneath.

The spherical tokamak (ST) has a very low aspect ratio, typically below ∼1.8, which allows operation at high beta. The ST also offers stability at high elongation which permits operation at high bootstrap current fraction leaving only a modest external current drive requirement. Results from START indicate the ST has resilience to disruptions and low halo currents when disruptions are triggered. In order to explore these potential advantages, a conceptual design of a steady-state ST power plant is being developed for which the thermodynamic, neutronic and mechanical design of the plant have been iterated, together with the plasma parameters, to give a consistent design. The baseline design uses water cooled copper for the centre rod and return limbs of the toroidal field (TF) coils with minimal steel shielding around the rod giving a simple coil design. A helium cooled ceramic pebble bed blanket with beryllium multiplier is used to generate the required tritium and achieve a high thermal efficiency.

To investigate the irradiation effects on a new version of reduced activation ferritic/martensitic steels (RAFMs) i.e. China Low Activation Martensitic steel (CLAM), neutron irradiation experiments has been being carried out under wide collaboration in China and overseas. In this paper, the mechanical properties of CLAM heats 0603A, 0408B, and 0408D were investigated before and after neutron irradiation to ∼0.02 dpa at 250 °C. The test results showed that ultimate strength and yield stress of CLAM HEAT 0603A increased about 10–30 MPa and ductile to brittle transition temperature (DBTT) shift was about 5 °C. For CLAM heats 0408B and 0408D, ultimate strength and yield stress increased about 80–150 MPa.

Surface preparation is essential for the Hot Isostatic Pressing (HIP) diffusion bonding of RAFM steels. Hot Isostatic Pressing (HIP) diffusion bonding experiments on China Low Activation Martensitic (CLAM) steel was performed to study the effect of surface preparation. A few approaches such as hand lapping, dry-milling and grinding etc., were used to prepare the faying surfaces of the HIP joints. Different sealing techniques were used as well. The HIP parameters were 150 MPa/3 h/1150 °C. After post HIP heat treatment (PHHT), the tensile and Charpy impact tests were carried out. The results showed that hand lapping was not suitable to prepare the faying surfaces of HIP diffusion bonding specimens although the surface roughness by hand lapping was very low.

The fully exposed center conductor post (CCP) in a spherical tokamak will receive strong neutron irradiation. The analytical results for a CCP in a spherical tokamak reactor are given considering the irradiation effects such as radiation damage, transmutation, nuclear and resistive heat removal, induced radioactivity and blanket tritium breeding ratio, etc. These results are compared with those for the first wall of conventional tokamaks, assuming the same technical requirements.

The fully exposed CCP in a low aspect ratio tokamak will receive high neutron damage, electrical resistive heating and neutron heating, which is one of the key components of the reactor and requires replacement at regular intervals. This paper gives the analysis results for the CCP in the low aspect ratio transmutation reactor considering parameters such as nuclear heating deposit, thermal-hydraulics parameters, radiation damage etc. Finally, those results are compared with those for the first wall of conventional tokamaks assuming the same technical requirements.