[en] As one of the innovative 4th generation nuclear reactors, the sodium cooled fast reactor (SFR) has unique advantages on heat transfer efficiency and breeding ratio. Due to the design requirement of a high fuel-coolant ratio, the fuel rods are usually arranged in triangular array. Helical wire spacers are widely used in the sodium cooled fuel assemblies. Another important feature is the pool type design of SFR. The most famous demonstration of SFR is Phenix reactor, which was closed in 2009. The upgraded future design version of Phenix is ASTRID [Le Coz et al. (2011)]. Comparing with water reactors which have a long term commercial operating experience, the special features of SFR result several unique detailed thermal hydraulic issues of fuel assembly and system. The engineering interested several points are listed as below. 1. Thermal hydraulic issues of the fuel assembly a) Sweeping ow induced by the special wire spacers b) Local heat transfer due to the triangular fuel rod arrangement and spacers c) Sub-channel analysis of fuel assembly by considering the above two points 2. Thermal hydraulic issue of the reactor system a) The local detailed 3D phenomena inside the reactor pool, which is coupled with the system behaviour. The thermal hydraulic of SFR fuel assembly is mainly influenced by its helical spacers and fuel rod arrangement pattern. One of the main effects of the wire spacers is enhancing the inter-channel cross flow in the triangular arrayed rod bundles. To study the characteristic of sweeping flow rate across the gaps between the sub-channels, the computational fluid dynamics (CFD) method is used in the present study to analyze the flow field in a 19-rod bundle with wire wraps. This CFD work only considers the fluid zone of one section of the rod bundle. The height of the CFD model with a periodic inlet and outlet boundary condition is one wire pitch. The k - $\mathrm{\omega <!--\; ??\; -->}$ SST model is used as the turbulent model. The wire effect on the velocity distribution of the flow across the interface between two sub-channels is investigated in the CFD study. Based on the force balance equation between the wire wake force and the surface friction force of rods in the circumferential direction, a theoretical model for estimating the average swirling flow rate driven by one single wire is proposed. The simulation shows that the ratio of rod pitch (P) to rod diameter (D) has a strong effect on the velocity of sweeping flow. The normalized sweeping flow velocity increases with the P/D ratio. The simulation agrees well with the proposed model. In addition, the effect of the ratio of wire pitch (H) to rod diameter has also an influence on the cross flow rate: A small H/D ratio leads to higher cross flow velocity. A correlation as a function of both P/D ratio and H/D ratio is proposed to predict the sweeping flow velocity by the present study. Relating to the sweeping flow and the triangular arrangement pattern, the circumferential heat transfer of fuel rod may have an uneven distribution rather than the uniform distribution assumed in most of the sub-channel approaches. Since the local flow field in the coolant flow passage is influenced by the wires, which induce the diversion flow, the circumferential heat transfer of the rod is strongly affected by both the arrangement and the wire spacer. To study the local heat transfer behavior of wire-wrapped rod bundle, the CFD code OpenFOAM is used in present study to figure out the detailed temperature and heat flux distributions on the surface of rod and wire. In present study, two CFD models for both bare rod bundle and wire wrapped 19-rod bundle are created. For the bare rod model, a periodical boundary condition is used to simulate the heat transfer of developed flow. For the wire wrapped 19-rod bundle case, the normal inlet outlet boundary is imposed to the fluid region with a length of two wire pitches. In the CFD studies, the models are split into solid and fluid regions. The k-omega SST model is used for turbulent simulation. The simulation result shows a cosine like local heat transfer distribution in the circumferential direction of the rod. Besides the detailed investigation of the triangular arrayed fuel assembly, the sub-channel approach code MATRA is improved for the coolant of sodium and local heat transfer calculation. The thermal properties functions for sodium are implemented. The pressure drop correlation, heat transfer correlation and turbulent mixing coefficient for sodium coolant from the literatures are also implemented. An automatic nodalization function is created for generating the input deck. Besides that, the proposed local heat transfer model is implemented for local cladding temperature and heat transfer coefficient output. Finally, the modified MATRA code is used for the fuel assembly analysis of ASTRID reactor. The thermal hydraulic of SFR reactor system is mainly influenced by its pool type design. The coolant loops of the sodium cooled reactor involve a strong local 3D phenomena in hot plenum and cold plenum, which are coupling with the system behavior. In present study, the system thermal hydraulic code ATHLET and CFD code OpenFOAM are coupled to simulate the dissymmetric test of the Phenix reactor. The simulation results are compared with experimental data. The main achievements of present study includes: (1) A new physical model is proposed for the sweeping flow induced by the wire spacers; (2) A new local heat transfer correlation is proposed for the triangular arrayed fuel rods; (3) The code MATRA is improved for simulating the fuel assemblies of the sodium cooled reactor; (4) The ATHLET and OpenFOAM coupling scheme is applied to analyze the dissymmetric test of the Phenix reactor.

[en] In order to carry out the test modeling LOCA of PWR, it is necessary to develop a device simulating pipe ruptures. The opening time of the devive must be less than a few milliseconds. In this paper, the design principle of the controllable pressurized double rupture disk device and its test results are briefly described

[en] Highlights: • Mechanistic modeling of nitrogen oxide electrochemical reduction. • Fundamentals of both alternative and direct current electrolysis. • Theoretical optimal frequency in alternative current electrolysis. - Abstract: A one-dimensional symmetric model on NO electrochemical reduction in solid oxide electrolysis cell(SOEC) considering gas transport, electronic conduction, ionic conduction, and electrochemical process based on multifunctional layer electrode is developed. The simulation results agree well with the experimental results both in the direct current(DC) and alternative current(AC) electrolysis. The distributions of the NO concentration in the electrode are predicted in both DC and AC electrolysis. The effects of temperature, voltage, and O_2 concentration were investigated on NO alternative current electrolysis and direct current electrolysis processes. The modeling results show that the optimal frequency of 0.3 Hz is corresponded to the maximum NO decomposition rate in different temperatures and voltages. The NO decomposition increases with increasing temperature and decreasing O_2 concentration in most cases. At 450 °C, the NO decomposition presents first increased and then decreased trend with different voltages at the frequency of 0.3 Hz. This is similar to the effects of O_2 concentration at 450 °C and 475 °C at the same frequency

[en] The binding energies of _Λ⁵He, _Λ⁹Be, _ΛΛ⁶He and _ΛΛ¹⁰Be hypernuclei are evaluated within the framework of the microscopic α-cluster model and with special attention to the ΛN exchange force effect. The ΛN exchange force, which contributes in different ways for He and Be hypernuclei, is an important ingredient for the attempt at consistent reproduction of the four binding energies. Yet the ΛN and ΛΛ interactions which reproduce the observed B_Λ(_Λ⁵He), B_Λ(_Λ⁹Be) and B_ΛΛ(_ΛΛ⁶He) within their experimental uncertainties lead to an overbinding of B_ΛΛ(_ΛΛ¹⁰Be) and inversely B_Λ(_Λ⁵He), B_Λ(_Λ⁹Be) and B_ΛΛ(_ΛΛ¹⁰Be) to an underbinding of B_ΛΛ(_ΛΛ⁶He). (author)

[en] Binding energies B_ΛΛ of double-Λ hyperons in _ΛΛ⁵H, _ΛΛ⁷He, _ΛΛ⁸Li and _ΛΛ⁹Li are estimated on the basis of the observed B_ΛΛ values of _ΛΛ⁶He and _ΛΛ¹⁰Be. A simplified verison of the α + x + Λ + Λ model with x = (-1) - (4) is employed along with ΛN and ΛΛ interactions with three-range Gaussian form. (author)

[en] The properties of the Hubbard-Holstein model for an electron-phonon system with strong electron correlations are investigated on the basis of a new diagram technique. The equations of the main dynamical quantities and optical dispersionless phonons are established. The problem of excluding partially or completely the phonon's coordinates and the corresponding renormalization of the physical parameters of the theory is discussed

[en] In-vessel retention (IVR) of core melt is a key severe accident mitigation measure adopted by some advanced light water reactors. Decay heat removal for IVR is through the external reactor vessel cooling (ERVC) by flooding the reactor cavity during severe accident. As long as the heat flux on the vessel outer surface is lower than the critical heat flux (CHF), the reactor vessel can be sufficiently cooled to maintain the RPV integrity. Heat flux on the RPV outer surface is mainly determined by the core melt behavior, whereas CHF depends on the thermal-hydraulic parameters in the cooling channel outside the RPV. In the present studies, computational fluid dynamics (CFD) models for IVR experiments were carried out, to understand various phenomena related to molten pool and cooling channel. The UCLA experiment and SLUTAN experiment were simulated by using CFD method, the heat source driven natural convection heat transfer coefficient in UCLA experiment, and the two phase boiling phenomena in SLUTAN experiment were investigated in this paper. The CFD calculation results were compared with the experiments data

[en] The in-vessel retention is adopted by the third generation nuclear power technology as an important severe accident mitigation strategy. The integrity of reactor pressure vessel depends on the heat flux distribution of molten pool. In present study, the solidification model in open source CFD software OpenFOAM was applied to simulate solidification and natural convection which was driven by internal heat source or temperature difference. The stratified molten pool heat transfer experiment carried out by Royal Institute of Technology was analyzed in the paper, and the solidified crust, temperature and heat flux distributions were obtained. The simulation results were compared with experimental data. It is shown that this numerical method can be used in the simulation of natural convection and solidification of molten pool, and it will probably be used in the analysis of molten corium behavior in reactor lower head. (authors)

[en] A spatial axisymmetric finite element model of single-crystal silicon irradiated by a 1064 nm millisecond laser is used to investigate the thermal stress damage induced by a millisecond laser. The transient temperature field and the thermal stress field for 2 ms laser irradiation with a laser fluence of 254 J/cm² are obtained. The numerical simulation results indicate that the hoop stresses along the r axis on the front surface are compressive stress within the laser spot and convert to tensile stress outside the laser spot, while the radial stresses along the r axis on the front surface and on the z axis are compressive stress. The temperature of the irradiated center is the highest temperature obtained, yet the stress is not always highest during laser irradiation. At the end of the laser irradiation, the maximal hoop stress is located at r=0.5 mm and the maximal radial stress is located at r=0.76 mm. The temperature measurement experiments are performed by IR pyrometer. The numerical result of the temperature field is consistent with the experimental result. The damage morphologies of silicon under the action of a 254 J/cm² laser are inspected by optical microscope. The cracks are observed initiating at r=0.5 mm and extending along the radial direction.

[en] Service water system (SWS) of AP1000 nuclear power plant in pipe gallery adopts high density polyethylene (HDPE) as the pipe material, and it is for the first time to use this kind of material in the SWS system in domestic nuclear power plant. This paper analyzes the characteristic of HDPE material, takes into account the features of Haiyang pipe gallery, and then makes a comparison on butt-fusion and electro-fusion of HDPE pipe welding process. SWS, being the nuclear power plant final heat sink, requires high reliability, so butt-fusion is the preferable weld method. It is hoped to provide reference for follow-up project design and construction. (authors)