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The Generation IV International Forum (GIF) Experts Group tasked the Risk and Safety Working Group (RSWG) with assessing the high-level safety design characteristics of all six GIF systems. The objective is to review and identify the main safety advantages and possible challenges of the individual technologies, with a view to assess the current status of safety- related research & development (R&D) activities and identify future R&D needs for each system. In this context, the RSWG requested that the GIF Lead-cooled Fast Reactor (LFR) provisional System Steering Committee (pSSC) prepare a system safety assessment (SSA) document summarizing the safety advantages and challenges of LFR concepts under consideration. The present report is structured according to the “table of contents” below as proposed by the RSWG. In preparing this assessment, the LFR pSSC has placed emphasis on the outcomes of safety- related R&D for the GIF LFR Reference Systems and the review of LFR safety aspects by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN)

Nowadays, there are many reconstruction methods base on the ex-core detectors to monitor the state of the core in real-time, but none of these methods can calculate the neutron spectrum which has always been an essential parameter in advanced reactors design, performance diagnoses and safe operation. Thus, by utilizing ex-core measurements, a reconstruction neutron field with energy distributions method based on artificial neural networks was proposed in this paper. Furthermore, in this study, in order to amplify the contribution of low-energy neutrons to the ex-core detectors, the idea of truncation plane is used to calculate the response matrix library. To validate the method, neutron field of pressure vessel regions and core regions in a homogenized assembly-level VENUS-3 reactor model were obtained through transport calculation and reconstruction calculation. The comparison of results demonstrated that the reconstruction method was effective and available to reconstruct the neutron spectrum in reactors.

To further investigate the fusion neutron source based on a gas dynamic trap (GDT), characteristics of the GDT were analyzed and physics analyses were made for a fusion neutron source based on the GDT concept. The prior design of a GDT-based fusion neutron source was optimized based on a refreshed understanding of GDT operation. A two-step progressive development route of a GDT-based fusion neutron source was suggested. Potential applications of GDT are discussed. Preliminary analyses show that a fusion neutron source based on the GDT concept is suitable for plasma-material interaction research, fusion material and subcomponent testing, and capable of driving a proof-of-principle fusion fission hybrid experimental facility.

The corrosion of Liquid lead or lead-bismuth eutectic (LBE) on the structural steels is a serious problem for their application in fast reactors or accelerator-driven transmutation systems. This paper focuses on the modeling and experimental study on LBE corrosion in the spent fuel transmutation systems, as well as the effects of corrosion products on heat transfer and flow. By studying the corrosion and deposit of the oxidation products on the pipe wall, the empirical correlations of thermal resistance and fouling factor can be theoretically developed and experimentally verified. The comparison of fouling factors obtained with each oxygen level is presented, the relationship between fouling factors and oxygen concentrations is correlated, and the effects of oxidation on heat transfer are demonstrated qualitatively by wetting conditions of the samples.

The $^{192}$Ir sources are widely used for high dose rate (HDR) brachytherapy treatments. The aim of this study is to simulate $^{192}$Ir MicroSelectron v2 HDR brachytherapy source and calculate the air kerma strength, dose rate constant, radial dose function and anisotropy function established in the updated AAPM Task Group 43 protocol. The EGSnrc Monte Carlo (MC) code package is used to calculate these dosimetric parameters, including dose contribution from secondary electron source and also contribution of bremsstrahlung photons to air kerma strength. The Air kerma strength, dose rate constant and radial dose function while anisotropy functions for the distance greater than 0.5 cm away from the source center are in good agreement with previous published studies. Obtained value from MC simulation for air kerma strength is $9.762\times 10^{-8} \textrm{UBq}^{-1}$and dose rate constant is $1.108\pm 0.13\%\textrm{cGyh}^{-1} \textrm{U}^{-1}$., Comment: 7 pages

This paper studied the adequacy of the World and China lithium resources, considering the most promising uses in the future, involving nuclear fusion and electric-vehicles. The lithium recycle model for D-T fusion power plant and electric-vehicles, and the logistic growth prediction model of the primary energy for the World and China were constructed. Based on these models, preliminary evaluation of lithium resources adequacy of the World and China for D-T fusion reactors was presented under certain assumptions. Results show that: a. The world terrestrial reserves of lithium seems too limited to support a significant D-T power program, but the lithium reserves of China are relatively abundant, compared with the world case. b. The lithium resources contained in the oceans can be called the "permanent" energy. c. The change in 6Li enrichment has no obvious effect on the availability period of the lithium resources using FDS-II (Liquid Pb-17Li breeder blanket) type of reactors, but it has a stronger effect when PPCS-B (Solid Li4 SiO4 ceramics breeder blanket) is used.

Intensity modulated radiation therapy (IMRT) requires the determination of the appropriate multileaf collimator settings to deliver an intensity map. The purpose of this work was to attempt to regulate the shape between adjacent multileaf collimator apertures by a leaf sequencing algorithm. To qualify and validate this algorithm, the integral test for the segment of the multileaf collimator of ARTS was performed with clinical intensity map experiments. By comparisons and analyses of the total number of monitor units and number of segments with benchmark results, the proposed algorithm performed well while the segment shape constraint produced segments with more compact shapes when delivering the planned intensity maps, which may help to reduce the multileaf collimator's specific effects.

To determine the electron energy spectra for medical accelerator effectively, we investigate a nonlinear programming model with several nonlinear regression algorithms, including Levenberg?Marquardt, Quasi?Newton, Gradient, Conjugate Gradient, Newton, Principal-Axis and NMinimize algorithms. The local relaxation-bound method is also developed to increase the calculation accuracy. The testing results demonstrate that the above methods could reconstruct the electron energy spectra effectively. Especially, further with the local relaxation-bound method the Levenberg?Marquardt, Newton and NMinimize algorithms could precisely obtain both the electron energy spectra and the photon contamination. Further study shows that ignoring about 4% photon contamination would increase error greatly, and it also inaccurately makes the electron energy spectra 'drift' to the low energy.

As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system.

Radiation damage to structural material of fusion facilities is of high concern for safety. The superconducting tokamak EAST will conduct D-D plasma experiments with the neutron production of 1015 neutrons per second. To evaluate the material radiation damage a programme system has been devised with the Monte Carlo transport code MCNP-4C, the inventory code FISPACT99, a specific interface, and the fusion evaluated nuclear data library FENDL-2. The key nuclear responses, i.e. fast neutron flux, displacement per atom, and the helium and hydrogen production, are calculated for the structural material SS-316L of the first wall, and the vacuum vessel, using this programme. The results demonstrate that the radiation damage to the structural material is so little that it will not lead to any significant change of material properties according to the reference design. This indicates that there is a large potential space for EAST to test advanced operation regime from the viewpoint of structural material safety.

The multi-group working nuclear data library HENDL1.0/MG is numerically tested with a series of existent benchmark spherical shell experiments (Si, Cr, Fe, Cu, Zr and Nb) by calculations using the multi-functional neutronics code VisualBUS. The ratio of calculated/measured neutron leakage rates and the neutron leakage spectra are presented in tabular and figural forms. The results from the calculations with the code ANISN and IAEA data library FENDL2.0/MG were also included for comparison, where the origination of the data used is different from that of HENDL1.0/MG.

HENDL1.0/MG, a multi-group working library of the Hybrid Evaluated Nuclear Data Library, was home-developed by the FDS Team of ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences) on the basis of several national data libraries. To validate and qualify the process of producing HENDL1.0/MG, simulating calculations of a series of existent spherical shell benchmark experiments (Al, Mo, Co, Ti, Mn, W, Be and V) have been performed with HENDL1.0/MG and the multifunctional neutronics code system named VisualBUS home-developed also by FDS Team.

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. 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 an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis.

The Fusion-Driven Subcritical System (FDS) is a subcritical nuclear energy system drived by fusion neutron source. In this paper, an advanced plasma configuration for FDS system has been proposed, which aims at high beta, high bootstrap current and good confinement. A fixed-boundary equilibrium code has been used to obtain ideal equilibrium configuration. In order to determine the feasibility of FDS operation, a two-dimensional time-dependent free boundary simulation code has been adopted to simulate time-scale evolution of plasma current profile and boundary position. By analyses, the Reversed Shear mode as the most attractive one has been recommended for the FDS equilibrium configuration design.

Tritium self-sustainment, which will meet the fuel requirement of fusion reactor, is one of the key issues of fusion power development. The tritium breeding performances of various tritium-breeding materials are compared based on a series of neutronics calculations using three-dimensional Monte Carlo neutron-photon transport code MCNP/4C with the IAEA FENDL-2 data library. The effects of the dimensions of the tritium-breeding zone and the enrichment of 6Li on Tritium Breeding Ratio (TBR) are analyzed. The effects of Be as a neutron multiplier on TBR are also calculated.

Effect of various spatial and energy distributions of fusion neutron source on the calculation of neutron wall loading of Tokamak D-D fusion device has been investigated by means of the 3-D Monte Carlo code MCNP. A realistic Monte Carlo source model was developed based on the accurate representation of the spatial distribution and energy spectrum of fusion neutrons to solve the complicated problem of tokamak fusion neutron source modelling. The results show that those simplified source models will introduce significant uncertainties. For accurate estimation of the key nuclear responses of the tokamak design and analyses, the use of the realistic source is recommended. In addition, the accumulation of tritium produced during D-D plasma operation should be carefully considered.

Preliminary analysis and calculation of liquid metal Li17Pb83 magnetohydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket. To decrease the liquid metal MHD pressure drop, Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts. The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed. Finally, the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.

The concept of the liquid Li17Pb83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR >1.05) and annual output of 100 kg or more fissile 239Pu (FBR >0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimizated calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (BR = TBR + FBR) is listed corresponding to different cases.

This paper proposes a conceptual structure of segmental water-cooled Center Conductor Post (CCP) to be flexible in installment and replacement. Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100 MW fusion power. Numerical simulation by using a commercial code PHOENICS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.

Neutronics optimization calculations have been performed for the tritium breeding blankets with solid ceramic breeder Li2O and liquid eutectic breeder Li17Pb83, respectively, based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport code MCNP/4B. The effects of beryllium, 6Li enrichment and various structural materials on Tritium Breeding Ratio have been systematically analyzed.

This paper proposes a sub-critical nuclear energy system driven by fusion neutron source, FDS, which can be used to transmute long-lived radioactive wastes and to produce fissile nuclear fuel as a way for early application of fusion technology. The necessity and feasibility to develop that system in China are illustrated on the basis of prediction of the demand of energy source in the first half of the 21th century, the status of current fission energy supply and the progress in fusion technology in the world. The characteristics of fusion neutron driver and the potential for transmutation of long-lived nuclear wastes and breeding of fissile nuclear fuel in a blanket are analyzed. A scenario of development steps is proposed.

The self-cooling molten-salt Flibe blanket of FFHR is numerically analyzed, resulting the optimum first wall to be as thin as 5mm and the heat flux up to 0.25MW/m 2 to be feasible with adopting V-4Cr-4Ti as the structural material. An alternative concept of free surface using a capillary force is shown to be feasible even in helical systems, where a spiral flow is formed and drastically enhances the heat transfer efficiency. The nuclear property of Flibe blanket is modified with increasing Be amount and adopting carbon reflector, resulting the local TBR of 1.3. As an optional technique, 50% enrichment of Li-6 gives the maximum TBR of 1.4.

A multi-objective hybrid genetic based optimization algorithm is proposed according to the multi-objective character of inverse planning. It is based on hybrid adaptive genetic algorithm, which combines the simulated annealing, uses adaptive crossover and mutation, and adopts niched tournament selection. The result of the test calculation demonstrates that excellent converge speed can be achieved using this approach. Key words- Inverse Planning Multi-objective optimization Genetic algorithm Hybrid.

Liquid metal, mainly lead and lead-bismuth eutectic (LBE) has been widely applied in the advanced nuclear reactors and accelerator-driven transmutation (ADS) systems for a series of excellent physical and chemical properties, but LBE is very corrosive to the structural materials such as iron and stainless steel in medium and high temperature circumstance which limits its development seriously. Research works of anticorrosion for LBE has been carried out all over the world. In this paper, A kinetic model for corrosion and corrosion products precipitation in non-isothermal LBE circuits is developed to predict corrosion and precipitation behaviors, the mass transfer convection-diffusion equation is established and the corrosion/corrosion products precipitation rate along the circuit is predicted considering the effects of temperature profile of the entire circuit which are different from local parameters, the overall estimate of the corrosion/precipitation situation is obtained, at the end, the solid-phase oxygen control technique is discussed, a program developed for dissolution rate of the PbO particles in the mass exchanger is proposed.Copyright © 2013 by ASME

Probabilistic Safety Assessment (PSA) has been widely used in nuclear safety and reliability field in recent years. PSA is playing a positive role in these fields and the analysis programs are also in rapid development, but the model created by one program cannot be used in another, so the results are difficult to be cross verified. A universal method is develped based on Open-PSA platform to solve such problem and is also used to check RiskA calculation engine. Results proves that RiskA is valid and reliable.

Risk monitor is a plant specific real-time analysis tool used to determine the instantaneous risk based on the actual status of the systems and components. And the tool can also be used to optimize the maintenance plan. So how to keep real-time is one of the important problems to be solved. It may be lack of computing resources when there are many users or large-scale maintenance plans because the risk monitor's server is a single computer and has high concurrency. Cloud computing can meet the requirement of risk monitor. In this paper the cloud platform is built, and the maintenance plan computing is taken as an example to verify the advantages of using cloud computing.

Fusion reactor is considered as one of the solutions for the sustaining development of nuclear energy. International Thermonuclear Experimental Reactor (ITER) is the biggest fusion reactor research plan in the world. High-intensity D-T fusion neutron generator can generate 14 MeV neutrons, and it matches the neutrons generated in ITER and be competently used for imitating the neutron environment in nuclear fusion reactor, which is important for the relevant experimental researches of blanket materials of fusion reactors. It can also be used for validating the correctness and reliability of the simulations and analyses in fusion basic studies, and can guide the subsequent material improvement and innovation of calculation methodology. A rotating tritium target system for D-T fusion neutron generator with a neutron yield of 1012 n·-1, i.e., a high intensity D-T fusion neutron generator, is proposed in this paper and the design, main parameters, technical difficulties and heat transfer enhancement method are introduced. The key and innovative technology of this rotating target system is the integration of the sprayed water cooling, mechanical seal and magnetic fluid seal technologies, which focuses on the heat transfer of the high heat power density in the target system. The most important technical index is that the maximum temperature on the target should not be above 200 ℃ as the tritium ions run away heavily from the tritium target when the target temperature is bigger than 200 ℃. To investigate the heat transfer characteristics of this rotating target system, the effects of water layer thickness, water flow rate and rotating speed on the heat transfer of this rotating target system are analyzed by computational fluid dynamics method. And the heat transfer processes of the target system under different heat power densities are also simulated and studied. The analysis results show that big water layer thickness, big water flow rate and high rotating speed are good for the heat transfer enhancement of the rotating target system, but the effects of the changes of the water layer thickness and water flow rate on the heat transfer process are both very small. Due to the design index, the heat power density on the target should be under a limit value, which is about 12 kW·cm-2 in the calculation results of this paper.

Currently, the conventional empirical or semi-empirical dose calculation algorithms for radiation treatment planning system have been selected as dose engines. The accuracy of these dose calculation algorithms is limited. The main problem is that they fail to adequately consider the lateral transport of radiation or loss of electronic equilibrium near interface between two heterogeneous mediums. Monte Carlo method provides a kind of precise and general solution to photon and electron dose calculation problem, but the puzzle of the longer CPU time affects its clinical application finally. The Monte Carlo code system (EGSnrc) has been used to investigate the different voxel sizes effect on the accuracy of dose distributions and computing time in the present paper. Based on a patient head and neck CT data case studied, it may be concluded that less than 4mm voxels should be used for Monte Carlo dose calculations to insure the superior accuracy of dose distribution, higher image resolution and shorter CPU time.

The decrease of accuracy in patients' repeated positioning in radiotherapy due to body's transformation is considered and a positioning error measurement method using dynamic template matching technique is brought forward in this paper. This method is an improvement of the matching method to correct positioning error based on the CT digitally reconstructed radiograph(DRR), which uses the video reconstructed 3D surface model that can reflect the body's transformation as the matching template instead of DRR. This method can not only improve the positioning accuracy but also eliminate the adverse influence brought by body's transformation.

In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R&D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. c