Experimental Study on Ammonia Co-Firing with Coal for Carbon Reduction in the Boiler of a 300-MW Coal-Fired Power Station

To reduce CO_(2) emissions from coal-fired power plants,the development of low-carbon or carbon-free fuel combustion technologies has become urgent.As a new zero-carbon fuel,ammonia(NH_(3))can be used to address the storage and transportation issues of hydrogen energy.Since it is not feasible to completely replace coal with ammonia in the short term,the development of ammonia-coal co-combustion technology at the current stage is a fast and feasible approach to reduce CO_(2) emissions from coal-fired power plants.This study focuses on modifying the boiler and installing two layers of eight pure-ammonia burners in a 300-MW coal-fired power plant to achieve ammonia-coal co-combustion at proportions ranging from 20%to 10%(by heat ratio)at loads of 180-to 300-MW,respectively.The results show that,during ammonia-coal co-combustion in a 300-MW coal-fired power plant,there was a more significant change in NO_(x) emissions at the furnace outlet compared with that under pure-coal combustion as the boiler oxygen levels varied.Moreover,ammonia burners located in the middle part of the main combustion zone exhibited a better high-temperature reduction performance than those located in the upper part of the main combustion zone.Under all ammonia co-combustion conditions,the NH_(3) concentration at the furnace outlet remained below 1 parts per million(ppm).Compared with that under pure-coal conditions,the thermal efficiency of the boiler slightly decreased(by 0.12%-0.38%)under different loads when ammonia co-combustion reached 15 t·h^(-1).Ammonia co-combustion in coal-fired power plants is a potentially feasible technology route for carbon reduction.

Preliminary results of in situ laser-induced breakdown spectroscopy for the first wall diagnostics on EAST

Post-mortem methods cannot fulfill the requirement of monitoring the lifetime of the plasma facing components （PFC） and measuring the tritium inventory for the safety evaluation. Laserinduced breakdown spectroscopy （LIBS） is proposed as a promising method for the in situ study of fuel retention and impurity deposition in a tokamak. In this study, an in situ LIBS system was successfully established on EAST to investigate fuel retention and impurity deposition on the first wall without the need of removal tiles between plasma discharges. Spectral lines of D, H and impurities （Mo, Li, Si ） in laser-induced plasma were observed and identified within the wavelength range of 500-700 nm. Qualitative measurements such as thickness of the deposition layers, element depth profile and fuel retention on the wall are obtained by means of in situ LIBS. The results demonstrated the potential applications of LIBS for in situ characterization of fuel retention and co-deposition on the first wall of EAST.

Review of LIBS application in nuclear fusion technology

Nuclear fusion has enormous potential to greatly affect global energy production. The next-generation tokamak ITER, which is aimed at demonstrating the feasibility of energy production from fusion on a commercial scale, is under construction. Wall erosion, material transport, and fuel retention are known factors that shorten the lifetime of ITER during tokamak operation and give rise to safety issues. These factors, which must be understood and solved early in the process of fusion reactor design and development, are among the most important concerns for the community of plasma-wall interaction researchers. To date, laser techniques are among the most promising methods that can solve these open ITER issues, and laser-induced breakdown spectroscopy (LIBS) is an ideal candidate for online monitoring of the walls of current and next-generation (such as ITER) fusion devices. LIBS is a widely used technique for various applications. It has been considered recently as a promising tool for analyzing plasma-facing components in fusion devices in situ. This article reviews the experiments that have been performed by many research groups to assess the feasibility of LIBS for this purpose.

Recent research progress on the compatibility of tritium breeders with structural materials and coatings in fusion reactors

Nuclear fusion energy is considered as a clean and safe energy source.As the fuel of deuterium-tritium(D-T)fusion reactor,T must be produced by the reaction between neutron and lithium(Li)in the breeders.The blanket of fusion reactor will work in a high-temperature and radioactive environment.The long-term contact between T breeders and structural materials in such a harsh environment will result in corrosion and microstructure modification of material surfaces,and then affect the mechanical properties and thermal conductivity of materials.To protect the structural materials from corrosion,coatings are applied to their surface.In addition,the coating also plays a role in preventing T permeation.The compatibility of T breeder materials with structural materials and coatings in the breeding blanket has always been a concern.In this paper,the up-to-date data on liquid and solid blanket of fusion reactors,the interaction behavior of T breeders with candidate structural materials,including reduced activation ferritic/martensitic steel,oxide dispersion-strengthened steel,silicon carbide and vanadium alloy,and coatings are reviewed.The corrosion mechanism is also expounded.Furthermore,the corrosion behavior between different types of materials is compared comprehensively.At the end,the research and development prospects on this topic are suggested.

Plasma-tungsten interactions in experimental advanced superconducting tokamak(EAST)

Tungsten(W)is used as the armor material of the International Thermonuclear Experimental Reactor(ITER)divertor and is regarded as the potential first wall material of future fusion reactors.One of the key challenges for the successful application of W in fusion devices is effective control of W at an extremely low concentration in plasma.Understanding and control of W erosion are not only a prerequisite for W impurity control,but also vital concerns to plasma-facing component(PFC)lifetime.Since the application of ITER-like water-cooled full W divertor in EAST in 2014,great efforts were made to inves-tigate W erosion by experiment and simulation.A spectroscopic system was developed to provide a real-time measurement of W sputtering source.Both experiment and simulation results indicate that carbon(C)is the dominant impurity causing W sputtering in L-mode plasmas,which comes from the erosion of C plasma-facing material(PFM)in the lower divertor and the main chamber limiters.The mixture layer on the surface of W PFCs formed through redeposition or the wall coating can effectively suppress W erosion.Increasing the plasma density and radiation can reduce incident ion energy,thus alleviating W sputtering.In H-mode plasmas,control of edge localized mode(ELM)via resonant magnetic perturbation(RMP)proves to be capable of suppressing intra-ELM W erosion.The experiences and lessons from the EAST W divertor are beneficial to the design,manufacturing and operation of ITER and beyond.

Preface to the special issue on plasma facing materials for fusion energy

International thermonuclear experimental reactor(ITER),the largest tokamak device so far,will operate with a full-tungsten divertor to handle the steady heat fluxes of 10 MW m^(−2),the slow transients of 20 MW m^(−2)(~10 s),as well as the transient heat fluxes up to~GM m^(−2)(<1 ms).Currently,tungsten(W)is also foreseen as the most suitable plasma-facing material(PFM)for the first wall in demonstration(DEMO)and future fusion reactors,as well as the divertor.The wall material in future fusion reactors must fulfill the requirements of sufficient lifetime,negligible or small long-term retention of tritium(T)fuel and an acceptable neutron activation level in long-term operation,which are favorable for W walls.

Achieving novel copper-steel joints with a combination of high strength and ductility reinforced by in-situ Fe-rich particles

Strength and ductility are typically mutually exclusive in traditional copper-steel joints.This work pro-poses a strategy to overcome the inherent trade-off between strength and ductility through high speed electron beam welding with a preferred deflection to facilitate the in-situ formation of Fe-rich particles in the Cu matrix.The Fe-rich particles with an average diameter of 178.5 nm feature a 3D spatial network distribution across practically the entire joint.The obtained joint reinforced with such Fe-rich particles achieves ultimate high tensile strength(413 MPa)while maintaining excellent ductility(22%).The im-proved strength of the copper-steel joint is derived from the combined effects of dislocation strengthen-ing and grain refinement strengthening,while the increase in room-temperature ductility is mainly due to the high Schmid factor up to 0.454,which promotes the primary slip system to initiate easily during tensile deformation.This work provides a novel perspective on creating copper-steel joints in terms of achieving microstructural refinement and outstanding strength-ductility synergy.