Effect of a body-tie structure fabricated by partial trench isolation on the suppression of floating body effect induced soft errors in SOI SRAM investigated using nuclear probes

Soft errors induced by proton, helium and oxygen ion irradiations were measured as a function of distance between a body electrode under partial trench isolation and a metal pad connected to a tungsten via for the first metal layer of a silicon-on-insulator (SOI) static random access memory. Abnormal drain charges induced by ion irradiations with various distances in the SOI metal oxide semiconductor field effect transistor were simulated to be compared with the experimental results. The soft errors were found to depend on the distance between the body electrode and the metal pad in the case of the abnormal drain charge, which is induced by incident ions, lower than the critical charge of the SRAM cells. The soft errors did not depend on the distance for the abnormal drain charges higher than the critical charge.     

Effectiveness and failure modes of error correcting code in industrial 65 nm CMOS SRAMs exposed to heavy ions

Single event upsets （SEUs） induced by heavy ions were observed in 65 nm SRAMs to quantitatively evaluate the applicability and effectiveness of single-bit error correcting code （ECC） utilizing Hamming Code. The results show that the ECC did improve the performance dramatically, with the SEU cross sections of SRAMs with ECC being at the order of 10-11 cm2/bit, two orders of magnitude higher than that without ECC （at the order of 10-9 cm2/bit）. Also, ineffectiveness of ECC module, including 1-, 2- and 3-bits errors in single word （not Multiple Bit Upsets）, was detected. The ECC modules in SRAMs utilizing （12, 8） Hamming code would lose work when 2-bits upset accumulates in one codeword. Finally, the probabilities of failure modes involving 1-, 2- and 3-bits errors, were calcaulated at 39.39%, 37.88% and 22.73%, respectively, which agree well with the experimental results.     

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k-ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.     

Simulating a 100 keV cascade in a simplified nuclear glass by a method combining ballistic and classical molecular dynamics

The CoMoD method (combined molecular dynamics) extends the scope of classical molecular dynamics to allow the investigation of energy levels corresponding to actual recoil nuclei with shorter computation times. The results obtained at 100 keV confirm those obtained by classical molecular dynamics at lower energies. After the acceleration of the initial projectile, the glass passes through a depolymerization phase that reaches a maximum before structure recovery.

The 100 keV cascade simulated here shows a persistent final depolymerization of about 150 chemical bonds. We also observe a decrease in the atomic density along the primary or secondary projectile path within a radius of 15 Å.

There does not seem to be any reason that would prevent applying the CoMoD method to other glass or crystalline matrices, although each matrix will require reconfiguration.     

Comparison of thermo-hydraulic performances of large scale vortex flow (LSVF) and small scale vortex flow (SSVF) mixing vanes in 17 × 17 nuclear rod bundle

Spacer grids in the nuclear fuel rod assembly maintain a constant distance between rods, secure flow passage and prevent the damage of the rod bundle from flow-induced vibration. The mixing vanes attached to the spacer grids generate vortex flows in the subchannels and enhance the heat transfer performance of the rod bundle. Various types of mixing vanes have been developed to produce cross flows between subchannels as well as vortex flows in the subchannels.The shapes of the mixing vane have been improved to generate larger turbulence and cross flow mixing. In the present study, two types of large scale vortex flow (LSVF) mixing vanes and two types of small scale vortex flow (SSVF) mixing vanes are examined. SSVF-single is conventional split type and SSVF-couple is split type with different arraying method. LSVF mixing vane has different geometry and arraying method to make large scale vortex. 17 × 17 rod bundle with eight spans of mixing vanes is simulated using the IBM 690 supercomputer. The FLUENT code and IBM supercomputer is employed to calculate the flow field and heat transfer in the subchannels.Turbulence intensities, maximum surface temperatures of the rod bundle, heat transfer coefficients and pressure drops of the four kinds of mixing vanes are compared. LSVF mixing vanes produced higher turbulence intensity and heat transfer coefficient than SSVF mixing vanes. Consequently, LSVF mixing vane increases the thermal efficiency and safety of the rod bundle.     

Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies

The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.     

First principles calculation of the elastic constants and phonon modes of UO2 using GGA + U with orbital occupancy control

The actinide oxide UO₂ is studied using the GGA + U method combined with monitoring of the orbital occupation matrices to avoid erroneous metastable states. The need to ensure strict convergence of the ground state energy with basis set energy cut-off is demonstrated, and a ground state is determined with a highly isotropic unit cell. Using this ground state the elastic constants and phonon modes of UO₂ are calculated and found to be in excellent agreement with experimentally determined values. Peaks observed in the experimental far-infrared response spectra are reproduced, and related to the vibrational modes of the heavy atoms in a static oxygen sub-lattice.     

Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid–structure-interaction coupling

A homogenisation method is presented and validated in order to perform the dynamic analysis of a nuclear pressure vessel with a “reduced” numerical model accounting for inertial fluid–structure coupling and describing the geometrical details of the internal structures, periodically embedded within the nuclear reactor. Homogenisation techniques have been widely used in nuclear engineering to model confinement effects in reactor cores or tubes bundles. Application of such techniques to rector internals is investigated in the present paper. The theory bases of the method are first recalled. Adaptation of the homogenisation approach to the case of rector internals is then exposed: it is shown that in such case, confinement effects can de modelled by a suitable modification of classical fluid–structure symmetric formulation. The method is then validated by comparison of 3D and 2D calculations. In the latter, a “reduced” model with homogenised fluid is used, whereas in the former, a full finite element model of the nuclear pressure vessel with internal structures is elaborated. The homogenisation approach is proved to be efficient from the numerical point of view and accurate from the physical point of view. Confinement effects in the industrial case can then be highlighted.     

A general survey of the potential and the main issues associated with the sulfur–iodine thermochemical cycle for hydrogen production using nuclear heat

The thermochemical sulfur–iodine cycle is studied by CEA with the objective of massive hydrogen production using nuclear heat at high temperature. The challenge is to acquire by the end of 2008 the necessary decision elements, based on a scientific and validated approach, to choose the most promising way to produce hydrogen using a generation IV nuclear reactor. Amongst the thermochemical cycles, the sulfur–iodine process remains a very promising solution in matter of efficiency and cost, versus its main competitor, conventional electrolysis. The sulfur–iodine cycle is a very versatile process, which allows lot of variants for each section which can be adjusted in synergy in order to optimise the whole process. The main part of CEA's program is devoted to the study of the basic processes: new thermodynamics data acquisition, optimisation of water and iodine quantity, optimisation of temperature and pressure in each unit of the flow-sheet and survey of innovative solutions (membrane separations for instance). This program also includes optimisation of a detailed flow-sheet and studies for a hydrogen production plant (design, scale, first evaluations of safety issues and technico-economic questions). This program interacts strongly with other teams, in the framework of international collaborations (Europe, USA for instance).     

Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.     

Stopping power of 11B in Si and TiO2 measured with a bulk sample method and Bayesian inference data analysis

The popular stopping power interpolative schemes require experimental data to be developed. Where the data bases are sparse, with few experiments available, interpolations can be more inaccurate. This is the case for the stopping of heavy ions, where even for important targets such as Si there is a need for more measurements. For compounds, the situation is even worse with very few measurements available. In particular, the stopping in oxides and nitrides often deviates significantly from what would be expected using the Bragg’s rule. We apply a method that uses bulk or thick film samples to determine the stopping power of ¹¹B in Si and TiO₂. The method, which relies on Bayesian inference analysis of RBS spectra obtained at different energies, has been previously validated by verifying the results obtained in the well-known system ⁴He in Si.     

Design and fabrication aspects of a plutonium-incineration experiment using inert matrices in a “Once-Through-Then-Out” mode

In the plutonium incineration experiment, named ‘Once-Through-Then-Out’ (OTTO), that is being prepared by JAERI, PSI and NRG, the use of highly stable inert matrices will be examined. The inert matrices MgAl₂O₄ spinel and ZrO₂ are insoluble in nitric acid and are considered as good storage media for final disposal. These inert matrices will be used in this experiment, which is representative for an OTTO scenario. A total of 7 Pu-containing targets were prepared for an irradiation in the High Flux Reactor in Petten. The objective of the irradiation is to reach a very high Pu-burnup. The main parameters to be studied are stability under irradiation, swelling, fission gas release and chemical interactions in the fuel. Four targets will be equipped with thermocouples for on-line monitoring of central temperature. Four of the targets contain MgAl₂O₄ as an inert matrix, 2 targets contain ZrO₂ and one target contains mixed-oxide (MOX) fuel for reference purposes. The fissile plutonium concentration is 0.32–0.44 g cm⁻³. Both particle-dispersed fuel and homogeneous dispersions were fabricated in order to test the effect of the size of the fissile inclusions. The design of the experiment and the fabrication of the samples are discussed.     

Substantiation of the parameters and layout solutions for an energy conversion unit with a gas-turbine cycle in a nuclear power plant with HTGR

Light is shed on questions concerning the application of a direct closed gas-turbine cycle for generating electricity in a nuclear power plant with HTGR. This makes it possible to use such reactor systems for generating electricity and(or) producing hydrogen from water and combine high electricity-generation efficiency of 47–50% with high safety. It is shown that it is advantageous to use a recovery gas cycle with recovery efficiency of at least 95% and intermediate cooling of helium in a compressor. The choice of the configuration of the reactor system is substantiated. The heat-exchange equipment of the gas-turbine cycle should have unique characteristics when placed in vessels with limited size and for operation at high temperatures (up to 600°C) and with a large pressure drop (up to 5 MPa). Approaches to solving the problems studied are elucidated on the basis of a modular helium reactor with a GT-MGR gas turbine.I. I. Afrikantov Special Office Design for Machine Engineering.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 24–36, January, 2005.     

Determination of the minimal number and optimal sensor location in a nuclear system with fixed incore detectors

The determination of the minimal number of sensors and the optimal sensor location in a nuclear system with fixed incore detectors, which is represented by a linear stochastic distributed parameter system, was studied in this work. The partial differential equation representing nuclear reactor dynamics was approximated to the finite dimensional ordinary differential equation by the modal expansion. A scalar measure of the covariance matrix error in the optimal filter was minimized with respect to the sensor locations. The necessary conditions for optimal sensor location were derived using the matrix minimum principle, thus making the calculations computationally more attractive. The locations of sensors were guessed initially through sensitivity analysis to reach solutions of the optimal location quickly. A method to determine the minimum number of sensors is suggested based on the observability and admissible error bound. Several numerical simulations are performed to determine the minimal number and optimal sensor location for a one-dimensional slab reactor and a two-dimensional ABB Combustion Engineering type reactor with fixed incore detectors. Through the simulations the possibility of practical implementation and the rapid convergence of the algorithm are verified.     

Evaluation of the conduction shape factor with a CFD code for a liquid–metal heat transfer in heated triangular rod bundles

A heat transfer due to conduction through a coolant itself is not negligible in a liquid–metal cooled reactor (LMR). This portion of a heat transfer is frequently described with a conduction shape factor during the thermal-hydraulic design of an LMR. The conduction shape factor, which is highly dependent on a pitch-to-diameter (P/D) ratio, is defined as the ratio of the local conduction heat flux at a gap between two subchannels to the reference heat flux calculated by the averaged subchannel temperatures. The shape factors in heated triangular rod arrays for three different pitch-to-diameter ratios are generated through CFX calculations in the present study. The flow paths of 1.0–2.0 m in length are meshed into 180,000–360,000 volumes depending on the flow velocities. The SSG Reynolds stress model is used as a turbulent model in the calculations. The evaluated data fell between the heated-rod data and the plane-source data obtained by theoretical investigations. The conduction shape factors were found to be independent of the heating pattern of the rod arrays. Based on the evaluated data, a correlation for a liquid sodium coolant is suggested, which will improve the accuracy of the subchannel analysis codes for the thermal-hydraulic design of an LMR. When it is compared with the existing correlations, the suggested correlation is expected to enhance the reliability of the conduction shape factor because the data is evaluated by a more realistic numerical experiment.