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The high intensity D-T fusion neutron generator (HINEG) has been developed at Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences/FDS Team, which will be a significant neutronic experimental platform for research and development of nuclear technology and safety, including the validation of neutronic method and software, radiation protection, material activation, and irradiation damage as well as neutronic performance of components. Its application can also be extended to nuclear medicine, radiotherapy, neutron imaging, and other nuclear technology applications. HINEG consists of two phases: the first phase, named HINEG-I, aims to have the intensity of 1012 n/s in order of magnitude, and the second phase, HINEG-II, is designed to reach a neutron yield of 1014-1015 n/s via high-power tritium target system and high-intensity ion source. HINEG-I has been completed and commissioning with the intensity produced of 1.1 × 1012 n/s, while the feasibility of HINEG-II has been preliminarily demonstrated. This paper will summarize all the latest progress and future plans for the research and development of HINEG. © 2016 The Authors. International Journal of Energy Research published by John Wiley & Sons Ltd. [ABSTRACT FROM AUTHOR]

Monte Carlo (MC) method has distinct advantages to simulate complicated nuclear systems and is envisioned as a routine method for nuclear design and analysis in the future. High-fidelity simulation with MC method coupled with multi-physics phenomena simulation has significant impact on safety, economy and sustainability of nuclear systems. However, great challenges to current MC methods and codes prevent its application in real engineering projects. SuperMC, developed by the FDS Team in China, is a CAD-based Monte Carlo program for integrated simulation of nuclear systems by making use of hybrid MC and deterministic methods and advanced computer technologies. The design objective, architecture and main methodology of SuperMC are presented in this paper. SuperMC2.1, the latest version, can perform neutron, photon and coupled neutron and photon transport calculation, geometry and physics modeling, results and process visualization. It has been developed and verified by using a series of benchmarking cases such as the fusion reactor ITER model and the fast reactor BN-600 model. SuperMC is still in its evolution process toward a general and routine tool for the simulation of nuclear systems. [ABSTRACT FROM AUTHOR]

Highlights • Established a comprehensive evaluation index system for spent fuel transportation. • Established a spent fuel transportation evaluation model based on weighted nearest neighbor method. • The application results show that the spent fuel transportation evaluation model is valid and effective. Abstract With the plan to increase nuclear generating capacity to 58 GWe and more than 30GWe under construction by 2020, China would have to confront the challenge from the transportation safety of nuclear material. In this paper, based on the analysis of transportation systems and the accident scenarios, the safety evaluation index system of spent fuel road transportation was established by using analytic hierarchy process (AHP), and the transportation safety evaluation model was constructed by weighted nearest neighbor method. By using the model, the safety evaluation of spent fuel road transportation in DaYa Bay nuclear power plant was carried out. The evaluation results showed that the whole transportation was safe, though it has safety risk caused by the weather and the junctions of the bridges and tunnels. In addition, the results from the fuzzy comprehensive evaluation method and the weighted nearest neighbor method are compared, which indicated that the weighted nearest neighbor method supplied a better decision support during the spent fuel road transportation. [ABSTRACT FROM AUTHOR]

Hydrogen is considered as the most potential energy carrier in the near future and can be produced from nuclear energy by several means. Lead-based reactor is one of the advance reactors in hydrogen production. In this paper, a novel feasible system of lead cooled reactor coupled with solid oxide electrolysis cells (SOEC) is presented, which is named China LEAd-based Reactor for Hydrogen production (CLEAR-H). By taking advantage of the chemically inert and high density of lead–bismuth eutectic (LBE), the concept of LBE-water direct-contact steam generator is proposed to supply high temperature (up to 600 °C) and low pressure (0.1–0.2 MPa) steam for SOEC stack without additional decompression devices. The results show that 47 tons hydrogen and 3000 tons oxygen-enriched air are obtained per day, and the thermal-to-hydrogen efficiency can reach about 39%. [ABSTRACT FROM AUTHOR]

The current refueling system of China Lead-based Research Reactor (CLEAR-I) is still in the developing stage and requires frequent changes of the refueling design scenarios. This paper presents a virtual simulation prototype which contributes to the complex refueling system for CLEAR-I. The proposed simulation system focuses on utilizing a desktop virtual environment to simulate the complex mechanical movement of triple crank connecting rod mechanism which consists of large rotating plug, small rotating plug and transfer machine. Main features of the proposed simulation system are: (1) automatic refueling scene modeling, (2) XML-based management for refueling scenario, (3) motion simulation and collision detection, (4) record and evaluation of refueling simulation process. The system architecture, features and capabilities of the developed virtual simulation prototype system were described. And the refueling simulation tests were carried out in order to meet the frequently updated engineering design for refueling scenarios of CLEAR-I. [ABSTRACT FROM AUTHOR]

Abstract: The preliminary studies of the activation analysis and waste management for blanket materials of the multi-functional experimental fusion–fission hybrid reactor, i.e. Multi-Functional eXperimental Fusion Driven Subcritical system named FDS-MFX, were performed. The neutron flux of the FDS-MFX blanket was calculated using VisualBUS code and Hybrid Evaluated Nuclear Data Library (HENDL) developed by FDS Team. Based on these calculated neutron fluxes, the activation properties of blanket materials were analyzed by the induced radioactivity, the decay heat and the contact dose rate for different regions of the FDS-MFX blanket. The safety and environment assessment of fusion power (SEAFP) strategy, which was developed in Europe, was applied to FDS-MFX blanket for the management of activated materials. Accordingly, the classification and management strategy of activated materials after different cooling time were proposed for FDS-MFX blanket. [Copyright &y& Elsevier]