Seawater desalination plant using nuclear heating reactor coupled with MED process

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the institute of Nuclear Energy Tech- nology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh wa ter and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven tech- nologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) pro cess. The NHR design possesses intrinsic and passive safety features, which was demon strated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED de salination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m^3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

High performance ORR electrocatalysts prepared via one-step pyrolysis of riboflavin

Efficient,cost‐effective electrocatalysts for an oxygen reduction reaction(ORR)are currently required for fuel cells.In the present work,riboflavin was used as a cheap,nontoxic carbon and nitrogen precursor to prepare Fe-N-C catalysts via one‐step pyrolysis in the presence of anhydrous iron chloride.Raman spectroscopy indicated that the catalyst containing nitrogen created a great quantity of defects in the carbon structures,while nitrogen adsorption‐desorption isotherms showed that the catalyst was mesoporous.Transmission electron microscopy demonstrated that the Fe-N-C catalyst was composed of very thin,curved and porous graphene layers together with some Fe2O3nanoparticles,and X‐ray diffraction patterns confirmed that the carbon in the catalyst was highly graphitized.X‐ray photoelectron spectroscopy indicated that the active sites for the ORR were primarily composed of graphitic nitrogen,although Fe sites also played an important role.The ORR activity of the Fe-N-C catalyst reached a maximum of4.16mA cm-2,and its chronoamperometric response was found to decrease by only3%after operating for3h at0.66V(vs RHE)in an O2‐saturated0.1mol L-1KOH solution.In contrast,a commercial40wt%Pt/C catalyst with a loading of0.2mgPt cm-2exhibited an activity of4.46mA cm-2and a40%loss of response.The electrochemical performance of this new Fe-N-C catalyst was therefore comparable to that of the Pt/C catalyst while showing significantly better stability.

Laser Cleaning of Mirror Surface for Optical Diagnostic Systems of the International Thermonuclear Experimental Reactor

The development of cleaning optics and deposition-mitigating techniques is a key factor in the construction and operation of optical diagnostics in ITER. The cleaning of optical surface by pulsed radiation from a fiber laser is an effective method that can recover optical properties of the mirror surface. The possibility of cleaning metallic mirrors from films with complex composition by pulsed radiation from a fiber laser has been experimentally researched. It has been shown that the high initial reflection characteristics of optical elements can be recovered by choosing regimes of radiation effect on the deposited surface. Efficient cleaning is ensured by radiation with the power density of less than 107 W/cm2. At this relatively low power density, pollutions are removed in a solid phase and the thermal effect on the mirror is insignificant. Preliminary experiments of the metal mirrors cleaning by fiber laser radiation have demonstrated the possibility of hardware implementation techniques.

Overview of Fusion Reactor Design

In this paper, the progress of the research activities of fusion reactor design at SWIP in 2006 is introduced.

Hydrogen Concentration of Co-Deposited Carbon Films Produced in the Vicinity of Local Island Divertor in LHD

In international thermonuclear experimental reactor （ITER）, one of major concerns is an in-vessel tritium inventory from a point of safety. It is believed that the carbon-tritium co-deposited film produced by the erosion of carbon diverter walls has a high tritium concentration. However, no systematic evaluation for the tritium concentration has been conducted yet. In the present study, the carbon-hydrogen co-deposited films were prepared at the wall of pumping duct in Local Island Divertor experiments of LHD, in order to evaluate the tritium concentration of the co-deposited films produced in ITER. The hydrogen concentration was obtained by measuring the amount of retained hydrogen in the film and the mass density of the film. The hydrogen concentration of the co-deposited carbon film at the wall not facing to the plasma with a low temperature was extremely high, 1.3 in the atomic ratio of H/C. This value is triple times higher than the previous value obtained so far. The crystal structure of the co-deposited carbon film observed by Raman spectroscopy showed very unique structure （polymeric aC：H）, which is well consistent with the high hydrogen concentration. The present study suggests that the tritium concentration of the co-deposited film in ITER depends on the wall position and becomes quite high as high as T/C-0.65. The results obtained contribute to evaluate the in-vessel tritium inventory owing to the co-deposited carbon films.

Investigation of the First Mirror on Tokamak

The study progress of NNSF in 2006 has been presented in this paper. The aim and plan of this item have been described briefly. Three test mirror exposure experiments have been done in this year. The main components of the deposited materials on the first mirror are carbon and iron analyzed by AES. The first mirror is polluted easier during glow discharge and wall processing than omhic discharge.

Observations of Runaway Electron Generation during Disruptions on HL-2A Tokamak

The major disruption can not only lead to the great heat loads and produce the large electromagnetic force on the first wall and divertor plates, but also can generate the runaway electrons. The runaway electrongeneration is a dangerous event for tokamak operation.

Research and Development of SWITRIM Code and Its Applications

The ftrst numerical simulation code package WITRIM has been developed to calculate the tritium inventory distribution and time-evolution in all sub-systems of FEB fusion reactor. The applications during recent six years indicate that it is reasonable and fully admitted by colleagues abroad. Some creative papers with new concept are published. For instance, we first time pointed out a new phenomenon of ＂tritium well depth and tritium well time＂ during the fusion reactor start-up phase. This is somewhat similar to, but quite different from the ＂iodine well depth and iodine well time＂ poisoning problem during restart-up process of a fission reactor. The authors not only proposed but also numerically solved this new phenomenon. The combination of the SWITRIM code package, user＇s guide, and application example are briefly introduced in this article.

Development of Vacuum Monitoring System in the HL-2A tokamak

HL-2A is the first divertor tokamak in China. The vacuum system is one of the important parts of the HL-2A tokamak, which has to be work well during the physics campaign in 2004. As the tokamak machine is separated from the operation staff during discharges, to guarantee the safe and reliable operation of the vacuum system, it is necessary to watch the vacuum system outside the machine hall with an automatic monitoring system. The paper describes the design and manufacture of the monitoring system.

Nanofluid Heat Transfer Enhancement for Nuclear Reactor Applications

Colloidal dispersions of nanoparticles are known as ＇nanofluids＇. Such engineered fluids offer the potential for enhancing heat transfer, particularly boiling heat transfer, while avoiding the drawbacks （e.g., erosion, settling, clogging） that hindered the use of particle-laden fluids in the past. At Massachusetts Institute of Technology （MIT）, the authors have been studying the heat transfer characteristics of nanofluids for the past five years, with the goal of evaluating their benefits for and applicability to nuclear power systems （e.g., primary coolant, safety systems, severe accident mitigation strategies）. This paper summarizes the MIT research in this area with particular emphasis to boiling behavior, including, prominently, the Critical Heat Flux limit and quenching phenomena.

Getting Free Heat Energy Based on Cavitation and Nuclear Fusion according to Revinov＇s Pilot Plants

In this report, the author describes and compares two innovative processes for producing thermal energy based on cavitation and nuclear fusion reactions in the indoor environment. Experiments conducted in the laboratory IHS （Intensive Heating System） of IE ＂Revinov N. M.＂ indicate that the process of lasso-vortex cavitation, which generates by means of an electric arc obtained HV-EI, gas-liquid plasma state in the EHH-CTC （electro-hydraulic heater with cavitation thermal camera） is not inferior and sometimes even superior in heat transfer to NF-AC （nuclear fusion reactions in ambient conditions）.

Comissioning of the lEA-R1 Nuclear Reactor New Heat Exchanger

This work presents results on the commissioning of the new heat exchanger of the IEA-R1 nuclear reactor in the occasion of its operational power upgrade from 2 MW to 5 MW, in comparison to the values calculated in the project of IESA Design and Equipments Company. This reactor is a swimming pool type, light water moderated and with graphite reflectors, used for research purposes and medical radioisotopes production. During monitoring procedures, issues were observed on the reactor operation at 5 MW mainly due to the ageing of the reactor＇s oldest heat exchanger （TC-A） and excessive vibrations at high flow rates on the other installed heat exchanger （TC-B）. So it was decided to provide a new IESA heat exchanger with 5 MW capacity to definitely substitute the TC-A heat exchanger. The results show that the IEA-R1 nuclear reactor can be operated safely and continuously at 5 MW with the new IESA heat exchanger.