[en] The activation aspects of pure fusion and hybrid fusion technology is studied to assess the radioactive safety of various fusion concepts including tokamak pure fusion, fissile fuel producing hybrid and radio waste transmuting hybrid. The activation properties of breeding, coolant and structural materials in fusion reactors might be quite different from those in fission reactors because of the high energy D-T fusion neutrons from the fusion cores. A study on the involved activation reactions and the uncertainties of the associated nuclear cross-sections is carried. The activation properties of various first wall concepts and blanket concepts are discussed. The radioactive inventory during the operation lifetime and the potential hazard of the radioactive nuclides with respect to near term (reprocessing) and long term (waste disposal) aspects are calculated, with reference to ITER/NET (International Thermonuclear Experiment Reactor/Next European Torus), STARFIRE (a commercial tokamak fusion power reactor), HEHR (Hefei Experiment Hybrid Reactor), HCHR (Hefei Commercial Hybrid Reactor) and THWT (Tokamak Hybrid Waste Transmuter), using the improved general one-dimensional multi-group radioactivity calculation code FDKR as well as the improved decay chain data library AF-DCDLIB. Also, a comparison among various conceptual fusion reactors, hybrid reactors and fission reactors including LWR (Light Water Reactor), HTGR (High Temperature Gas Reactor) and FBR (Fast Neutron Breeding Reactor) is carried out

[en] Highlights: • A modified ensemble Kalmen filter data assimilation method is proposed. • The method can consider four main uncertain parameters in the puff model. • The prediction of radioactive material atmospheric dispersion is improved. • The source release rate and plume rise height are successfully reconstructed. • It can shorten the time lag in the response of ensemble Kalmen filter. - Abstract: Atmospheric dispersion models play an important role in nuclear power plant accident management. A reliable estimation of radioactive material distribution in short range (about 50 km) is in urgent need for population sheltering and evacuation planning. However, the meteorological data and the source term which greatly influence the accuracy of the atmospheric dispersion models are usually poorly known at the early phase of the emergency. In this study, a modified ensemble Kalman filter data assimilation method in conjunction with a Lagrangian puff-model is proposed to simultaneously improve the model prediction and reconstruct the source terms for short range atmospheric dispersion using the off-site environmental monitoring data. Four main uncertainty parameters are considered: source release rate, plume rise height, wind speed and wind direction. Twin experiments show that the method effectively improves the predicted concentration distribution, and the temporal profiles of source release rate and plume rise height are also successfully reconstructed. Moreover, the time lag in the response of ensemble Kalman filter is shortened. The method proposed here can be a useful tool not only in the nuclear power plant accident emergency management but also in other similar situation where hazardous material is released into the atmosphere

[en] A spherical tokamak fusion-fission hybrid system has been proposed for transmutation of long-lived actinides and fission products and production of nuclear energy. The preliminary conceptual design of the system has been presented. (authors)

[en] A preliminary conceptual design of dual-cooled long-lived radioactive waste transmutation blanket for the Fusion-Driven sub-critical hybrid System (FDS) is presented on the basis of feasible plasma physics and technology level i.e. the neutron wall loading is assumed to 0.5 MW/m² with availability of 50%. The concept has the attractive advantages e.g. tritium is self-sustainable, plutonium for the purpose of neutron multiplication is self-sustainable and the thermal power output is stable. The one-dimensional neutronic transport calculation using the discrete ordinate method and burnup calculation using the direct numerical method are carried out to assess the performance of the blanket

[en] The low aspect ratio tokamak is proposed for the driver of a transmutation reactor in this paper. The main parameters of the reactor core, neutronic analysis of the blanket are given. Our previous analysis showed that the neutron wall loading can be lowered to the magnitude order of 1 MW/m² for the adequate transmutation capacity and efficiency. The recent analysis puts the conclusion forward. The neutron wall loading can be lowered even to the magnitude order of 0.5 MW/m² which is much easier to reach in the near future. It is also shown that the transmutation efficiency (fission/absorption ratio) is higher than the previous one. The blanket power density is about 200 MW/m³ which is not difficult to deal with. The key components such as divertor and center conductor post of the low-A transmutation reactor are also analyzed and compared with conventional tokamak. Finally, by the comparison with the other drivers such as FBR, PWR and Accelerator etc and pure fusion anticipation, the low aspect ratio transmutation reactor is one of near future applications of fusion energy. 6 refs., 5 figs., 12 tabs

[en] This paper summarizes recent achievements in the characterization of candidate vanadium alloys obtained for fusion in the framework of the Japan-China Core University Program. National Institute for Fusion Science (NIFS) has a program of fabricating high-purity V-4Cr-4Ti alloys. The resulting products (NIFS-HEAT-1,2), were characterized by various research groups in the world including Chinese partners. South Western Institute of Physics (SWIP) fabricated a new V-4Cr-4Ti alloy (SWIP-heat), and carried out a comparative evaluation of hydrogen embrittlement of NIFS-HEATs and SWIP-Heat. The tensile test of hydrogen-doped alloys showed that the NIFS-HEAT maintained the ductility to relatively high hydrogen levels. The comparison of the data with those of previous studies suggested that the reduced oxygen level in the NIFS-HEATs should be responsible for the increased resistance to hydrogen embrittlement. Based on the chemical analysis data of NIFS-HEATs and SWIP-Heats, neutron-induced activation was analyzed in Institute of Plasma Physics (IPP-CAS) as a function of cooling time after the use in the fusion first wall. The results showed that the low level of Co dominates the activity up to 50 years followed by a domination of Nb or Nb and Al in the respective alloys. It was suggested that reduction of Co and Nb, both of which are thought to have been introduced via cross-contamination into the alloys from the molds used should be crucial for reducing further the activation. (authors)

[en] Material corrosion in lead or Lead-Bismuth Eutectic (LBE) is the important limiting factor to keep integrity of internal components of lead-based reactor. In order to verify engineering feasibility and screen corrosion-resistant material in LBE environment, the compatibility of structural materials with LBE at 500 °C and 550 °C was evaluated. First, T91 and 15-15Ti steels were selected to be tested in stagnant LBE with different oxygen concentrations to investigate the influence of oxygen concentrations on corrosion behaviour of typical martensitic and austenitic steels. For the two types of steels, the formation of protective oxide layer is sensitive to the oxygen concentration. Second, long-term corrosion tests were carried out in large-scaled KYLIN-II material corrosion loop to evaluate the corrosion performance of the candidate structure materials under service condition of CLEAR-I. It is found that the growth kinetics curves of oxide layers for T91, 15-15Ti, CLAM and 316L steels follow a parabolic rule (Δx²=K_pt), and the rate constant for 15-15Ti steel is lowest. Thirdly, new Si-contained stainless steel and ODS-9Cr steel have been developed in INEST and compatibility evaluation revealed that the corrosion resistances of the above steels have attained considerable improvement. (author)

[en] A low aspect ratio tokamak transmutation system is proposed as an alternative application of fusion energy on the basis of a review of previous studies. This system includes: (1) a low aspect ratio tokamak as fusion neutron driver, (2) a radioactivity-clean nuclear power system as blanket, and (3) a novel concept of liquid metal centre conductor post as part of the toroidal field coils. In the conceptual design, a driver of 100 MW fusion power under 1 MW/m² neutron wall loading can transmute the amount of high level waste (including minor actinides and fission products) produced by ten standard pressurized water reactors of 1 GW electrical power output. Meanwhile, the system can produce tritium on a self-sustaining basis and an output of about 2 GW of electrical energy. After 30 years of operation, the biological hazard potential level of the whole system will decrease by two orders of magnitude. (author)

[en] An advanced tokamak transmutation system is proposed as an alternative application of fusion energy based on a review of the previous studies. This system includes: (1) a low aspect ratio tokamak as fusion neutron driver, (2) a Radioactivity Clean Nuclear Power System (RCNPS) as blanket, (3) a novel concept of liquid metal center conductor post (CCP) as part of toroidal field coils. A preliminary feasibility study has been carried out for the system, which included the aspects of core plasma physics, blanket neutronics and design. A driver of 100MW fusion power under 1MW/m² neutron wall loading can transmute the amount of High Level Waste (including minor actinides and fission products) produced by 10 standard pressurized water reactors of 1 GW electrical power output. Meanwhile, the system can produce tritium and output about 2 GW of electrical energy. After 30 years of operation , the biological hazard potential (BHP) level of the whole system will decrease by 2 orders of magnitude. (author)

[en] Dual-functional lithium lead-test blanket module (DFLL-TBM) is designed to demonstrate and validate related technologies for the high-power density liquid metal blanket of fusion commercial demo reactors (e.g. FDS-II) and the He/LiPb dual-cooled multi-functional blanket of a fusion-driven sub-critical system (i.e. FDS-I). It is important to assess the reliability of DFLL-TBM due to its complexity. A new approach is proposed by combining the probabilistic fracture mechanics (PFM) method and the probabilistic safety assessment (PSA) method. A reliability analysis code for blanket modules is developed. One of the failure modes of DFLL-TBM, leakage failure, is analyzed as an example to demonstrate the capability of the combinational approach. Some key aspects that affect the leakage reliability of DFLL-TBM are identified