Virtual PIXE and RBS laboratory

An interactive software package for a complete simulation of Particle Induced X-ray Emission (PIXE) and Backscattering Spectra (BS) is being developed. The user is in a position to define all experimental parameters such as incident ions (protons, deuterons or He ions), their energy, excitation and detection geometry, detector parameters and filters as well as sample composition and thicknesses of a number of layers. This is all done through an easy to operate interactive user interface. Simulated PIXE and RBS spectra are immediately displayed on the screen and can be saved either as bitmaps and/or files suitable for further processing. Each output comes with a complete set of experimental parameters, absolute and relative line intensities (including all major K and L lines), peak to background ratio and detection limits for all elements of interest.

The program has also a number of utility routines to calculate various fundamental parameters such as photon cross sections, K and L shell ionization and production cross sections, energy loss, and detector efficiency. All these routines use the state-of-the art data base sources.

The program operates on a personal computer under the MS Windows operating system. The simulation is fast and the program is easy to operate. The package will be useful in many ways. Firstly, it will be an excellent teaching tool for researchers/students without experience with PIXE/RBS. Secondly, it will be of immense help in planning and optimizing PIXE and/or RBS experiments. The user can ‘run’ a simulated experiment for any sample/experimental parameters and test various settings/scenarios to achieve optimal results without actually carrying out the experiment, thereby saving a lot of expensive machine time that would otherwise have been spent on trial and error experiments.     

ADS加速器束流瞬变分析程序开发

加速器驱动的反应堆系统(ADS)中次临界堆芯的功率水平依靠强流质子轰击散裂靶产生的中子源来维持,质子束流的不稳定性将对次临界堆的功率水平产生影响,进而对ADS的安全性产生影响.本文研究了ADS系统束流瞬变事故特性,建立了相应的物理数学模型,设计开发出具有较强针对性的用于ADS系统束流瞬变事故仿真软件--SIMULINK-ADS.并选取了典型的束流瞬变工况进行分析,通过与OECD/NEA和FZK Karlsruhe研究成果进行比较,验证了SIMULINK-ADS程序能够有效地计算和分析ADS束流瞬变次临界反应堆堆芯物理及热工响应.     

Automated metadata,provenance cataloging and navigable interfaces: Ensuring the usefulness of extreme-scale data

For scientific research, it is not the mere existence of experimental or simulation data that is important, but the ability to make use of it. This paper presents the results of research to create a data model, infrastructure, and a set of tools that support data tracking, cataloging, and integration across a broad scientific domain. The system is intended to document workflow and data provenance in the widest sense. Combining research on integrated metadata, provenance, and ontology information with research on user interfaces has allowed the construction of early prototype. While using fusion science as a test bed, the system's framework and data model is quite general.     

Enhanced tritium production for fusion reactors via3He(n,p)3 H in the heavy water moderator of a CANDU reactor

There is an increasing requirement for tritium to supply the fuel needs of current experimental fusion devices and in the initial startup of future power generating reactors. Tritium is produced in heavy water reactors through deuterium activation, but the total production capacity of Canadian operated CANDUs will fall short of future demands, during the period before and for some time after self-sufficient reactors become available. Consequently, methods of enhancing tritium generating rates warrant investigation. Herein we provide the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the heavy water moderator system of a hypothetical 500–600 MWe CANDU reactor. The approach adopted involves simulation of the temporal evolution of the tritium activities, originating from²H(n,)³H and³He(n, p)³H, as described by a simple first order kinetic model. The results suggest that the frequent addition of helium-3 to the moderator water will enhance tritium production inventories. The enhancement factor is highly dependent upon the rate at which helium-3 irretrievably escapes to the moderator cover gas. However, the direct activation of helium-3, contained in a closed loop such as the annulus gas system, for example, would be essentially complete within a few weeks without any significant loss.     

k-? modeling using modified nodal integral method

The simulation of turbulent flows is an ongoing challenge. This is especially true for the flows in the nuclear reactors. In order to save computational time and resource, accurate numerical schemes are required for such simulations. The encouraging results from the laminar flow simulations using modified nodal integral method (MNIM), serves as a motivation to use the method for turbulent flow simulations. The k-? model in this work has been implemented using the MNIM formulation. Two models, one for low Reynolds number and one for high Reynolds number, are implemented. The application of the model to relatively simple problems shows that results are good and similar to what one would expect from the k-? model implementation with any other numerical scheme. The results are compared with the DNS data from various sources in the literature. The difference between the DNS data and current implementation arises mainly from the assumption made in the k-? model rather than the choice of the numerical scheme in the present work. It is seen that very coarse grids can be used away from the walls for the present simulation. This is especially true for low Reynolds number model. Hence, MNIM formulation for the k-? model promises to reduce the over all computational cost.     

Generation of Radiation Transport Geometry from Computer-Generated Design Data

Monte Carlo simulation is a valuable tool for analyzing radiation effects in three-dimensional objects. Generating the geometry data necessary for describing solids to the simulation programs can be tedious and time consuming, and it is extremely error-prone. A faster, more accurate method of defining geometry data would speed up the radiation transport analysis process sufficiently for the analysis results to influence design. For computer-generated design data, manually defining the geometry data for radiation transport programs is an unnecessary step if software to translate the data into an appropriate format for the radiation transport calculations is available. This paper will present work on the translation of a particular type of computer-generated design data to a format suitable for the MORSE radiation transport code and will discuss applications for its use in radiation transport calculations.     

Large eddy simulation in pebble bed gas cooled core reactors

A high temperature gas-cooled reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concept is currently under consideration and development worldwide. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) was performed using the large eddy simulation. This would help in understanding the highly three-dimensional, complex flow phenomena caused by flow curvature in the pebble bed. Resolving all the scales of a turbulent flow is too costly, while employing highly empirical turbulence models to complex problems could give inaccurate simulation results. The large eddy simulation (LES) method would overcome these shortcomings. An attempt to obtain experimental velocity flow patterns using particle image velocimetry technique combined with matched refractive index liquid was pursued.     

Radioactive Transportation,Clarity and Consensus: IBC Conference, 5–6 October 2006, London,UK

Abstract

It is believed that there will be shipments of radioactive material and waste, as the nuclear industry moves into maturity, which may not be well accommodated within the current transport regulations. Although these shipments could be made under Special Arrangement approvals, the regulatory system would be better served if more formal requirements and criteria were included in the regulations and the shipments were considered in accordance with the regulations. A Special Arrangement approval is now defined as authorising transport of a shipment which does not satisfy all the applicable requirements of the regulations. This paper proposes the Transport System approach to regulating these types of shipments, where operational restrictions or other packaging provisions could compensate for the absence or inadequacy of packaging or other associated requirements. These shipments would require Competent Authority approval, and acceptance criteria would be included in terms of limits on probability, consequences and risk. The process would be limited to those types of shipments where the package system does not work well. The advantages of including Transport System approval within the regulations include reduction in the time required to obtain an approval, greater efficiency of decontamination and decommissioning operations, and assurance of an equivalent level of safety.     

NURESIM - A European simulation platform for nuclear reactor safety: Multi-scale and multi-physics calculations, sensitivity and uncertainty analysis

Starting in 2005 with the NURESIM Integrated Project (FP6), a European Reference Simulation Platform for Nuclear Reactors called NURESIM is being developed. This development follows a roadmap which is consistent with the SRA (Strategic Research Agenda) of the European SNETP (Sustainable Nuclear Energy Technology Platform). After delivery of two successive versions during the course of the NURESIM project, the numerical simulation platform is presently being developed in the frame of the NURISP European Collaborative Project (FP7), which includes 22 organizations from 14 European countries.NURESIM intends to be a reference platform providing high quality software tools, physical models, generic functions and assessment results.The NURESIM platform provides an accurate representation of the physical phenomena by promoting and incorporating the latest advances in core physics, two-phase thermal-hydraulics and fuel modelling. It includes multi-scale and multi-physics features, especially for coupling core physics and thermal-hydraulics models for reactor safety. Easy coupling of the different codes and solvers is provided through the use of a common data structure and generic functions (e.g., for interpolation between nonconforming meshes).More generally, the platform includes generic pre-processing, post-processing and supervision functions through the open-source SALOME software, in order to make the codes more user-friendly.The platform also provides the informatics environment for testing and comparing different codes. For this purpose, it is essential to permit connection of the codes in a standardized way. The standards are being progressively built, concurrently with the process of developing the platform.The NURESIM platform and the individual models, solvers and codes are being validated through challenging applications corresponding to nuclear reactor situations, and including reference calculations, experiments and plant data. Quantitative deterministic and statistical sensitivity and uncertainty analyses tools are also developed and provided through the platform.A Users’ Group of European and non-European countries, including vendors, utilities, TSOs, and additional research organizations (beyond the current partners) has also been established in order to enhance the role of the simulation platform in meeting the needs of the nuclear industry, as applied to current and future nuclear reactors.This presentation summarizes the achievements and ongoing developments of the simulation platform in core physics, thermal-hydraulics, multi-physics, uncertainties and code integration.     

Experiment archive,analysis, and visualization at the National Ignition Facility

The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory is the world's most energetic laser, providing a scientific research center to study inertial confinement fusion and matter at extreme energy densities and pressures. A target shot involves over 30 specialized diagnostics measuring critical x-ray, optical and nuclear phenomena to quantify ignition results for comparison with computational models. The Shot Analysis and Visualization System (SAVI) acquires and analyzes target diagnostic data for display within a time-budget of 30 min. Laser and target diagnostic data are automatically loaded into the NIF archive database through clustered software data collection agents. The SAVI Analysis Engine distributes signal and image processing tasks to a Linux cluster where computation is performed. Intermediate results are archived at each step of the analysis pipeline. Data is archived with metadata and pedigree. Experiment results are visualized through a web-based user interface in interactive dashboards tailored to single or multiple shot perspectives. The SAVI system integrates open-source software, commercial workflow tools, relational database and messaging technologies into a service-oriented and distributed software architecture that is highly parallel, scalable, and flexible. The architecture and functionality of the SAVI system will be presented along with examples.     

Natural convection analyses and verification for LMFBR cores

Analytical model requirements for core natural convection analyses are reviewed. Then results from current modeling on intra-assembly flow and heat redistribution are compared with several sources of experimental data. Also, data are described on low flow rod bundle hydraulic characteristics. Numerous sensitivity studies are also presented which show the effect and importance of various parameters on core temperatures during natural circulation, including inter-assembly flow redistribution, pump flow coastdown, rod size and fuel type, control system scram worth and shutdown power level. A system of codes for making the natural circulation predictions is also described, i.e., a plant-wide dynamic code, a whole-core system dynamic code and a hot channel dynamic analysis code. The overall approach of verifying the core related codes is presented, along with the interaction and linkage between all the codes. Confirmation of this system of three codes will bee through prototypic data obtained from EBR-II and FFTF natural circulation experiments.     

The Automatic Data Processing System for 2X II

An automatic data system is being developed to record transient data that will replace the present method of photographing scope traces with a Polaroid camera. The advantages of the new system are that it will increase data accuracy, eliminate long hours of manual reduction of data, provide instant processing of data to aid in optimizing the experimental parameters, and provide rapid comparison of current data with data taken from previous runs for statistical analysis. The 2X II automatic data system is composed of an analog disc connected to a small computer system. The multichannel analog disc is the principal intermediate means of recording the data, which are then digitized and stored permanently on magnetic tape. Reading of the data is accomplished by scanning all of the disc channels with a single analog-to-digital converter and transferring the data into the computer's magnetic core memory. This scanning of the analog disc is done automatically by the computer's software program, but it can also be done either manually from the disc control panel or remotely from a teletype terminal. Currently the digitized data are fed into a mini-computer for on-line data reduction and display. Using teletype control, one can view any individual channel from 0 to 30 ms with 10 ?s resolution, or one can retrieve processed data (for example, the number of charge exchange neutrals vs energy for any selected time up to 30 ms).     

Plan of ITER remote experimentation center

Plan of ITER remote experimentation center (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is described. Objectives of REC activity are (1) to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, (2) to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, and (3) to test the functions of REC and demonstrate the total system by using JT-60SA and existing other facilities in EU. Preliminary identified items to be developed are (1) Functions of the remote experiment system, such as setting of experiment parameters, shot scheduling, real time data streaming, communication by video-conference between the remote-site and on-site, (2) Effective data transfer system that is capable of fast transfer of the huge amount of data between on-site and off-site and the network connecting the REC system, (3) Storage system that can store/access the huge amount of data, including database management, (4) Data analysis software for the data viewing of the diagnostic data on the storage system, (5) Numerical simulation for preparation and estimation of the shot performance and the analysis of the plasma shot. Detailed specifications of the above items will be discussed and the system will be made in these four years in collaboration with tokamak facilities of JT-60SA and EU tokamak, experts of informatics, activities of plasma simulation and ITER. Finally, the function of REC will be tested and the total system will be demonstrated by the middle of 2017.     

Impact of Nuclear Data Uncertainties on Innovative Fast Reactors and Required Target Accuracies

It is widely accepted that the current status of neutronics calculations for fast reactor design is such that the present uncertainties on nuclear data should still be significantly reduced, in order to get the full benefit from advances in modeling and simulation. Only a parallel effort in advanced simulation, high-accuracy validation experiments, and nuclear data improvement will provide designers with more general and wellvalidated calculation tools to meet tight design target accuracies to further improve safety and economics. The present paper presents very recent results related to nuclear data uncertainty impact assessment and target accuracy requirements for advanced reactor systems.     

Development,computer simulation and performance testing in sodium of an eddy current flowmeter

Sodium is used as a coolant in Liquid Metal Fast Breeder Reactor (LMFBR). Sodium flow measurement is of prime importance both from the operational and safety aspects of a fast reactor. Various types of flowmeters namely permanent magnet, saddle type and eddy current flowmeters are used in FBRs. From the safety point of view flow through the core should be assured under all operating conditions. This requires a flow sensor which can withstand the high temperature sodium environment and can meet the dimensional constraints and be amenable to maintenance. Eddy current flowmeter (ECFM) is one such device which meets these requirements. It is meant for measuring flow in PFBR primary pump and also at the outlets of the fuel sub-assemblies to detect flow blockage. A simulation model of ECFM was made and output of ECFM was predicted for various flowrates and temperatures. The simulation model was validated by testing in a sodium loop. This paper deals with the design, simulation and tests conducted in sodium for the eddy current flowmeter for use in the Prototype Fast Breeder Reactor (PFBR).     

Paralleling Inverters Study of EAST Vertical Stability Coil Power Supply

Experimental advanced superconducting tokamak vertical stability (VS) coil power supply is a large capacity single phase inverter power supply. To meet the requirement of large current and fast response, multi-inverters in parallel is presented, which based on carrier phase-shifted modulation technology. In parallel inverter system, the disperse circuit parameters and phase-shift carriers between parallel inverter units will cause circulating current, which contains fundamental component and a large number of harmonic components. In this paper, the model of circulating current is analyzed when VS coil power supply is working in voltage given mode, and an instantaneous current sharing control strategy is proposed based on the combination of current sharing inductor and instantaneous circulating current feedback control. Parallel inverter units are connected together through current sharing inductors which can change the impedance characteristic of the circulating impedance and well restrain the high-frequency circulating current. Then, the real part of the circulating impedance will be increased and the ability to restrain the low-frequency circulating current will be advanced by introducing virtual resistance, which is realized in the instantaneous circulating current feedback routine. The designations of the current sharing inductor and the virtual resistance are provided. The results of simulation and experiment verify that this current-sharing strategy is available and efficient.     

Mesh generation technology for nuclear reactor simulation; barriers and opportunities

Mesh generation in support of nuclear reactor simulation has much in common with the requirements of other application areas, such as computational fluid dynamics (CFD). Indeed, fluid dynamics analysis of the coolant behavior inside the reactor core is an internal flow problem that requires the resolution of spatial and temporal variations in the flow caused by complex component configurations, fluids/structure interaction, turbulence, and thermal heating of the coolant. Typical concerns of meshing complex geometries; the use of hexahedral vs. tetrahedral elements, element geometric quality, mesh smoothness, use of anisotropic elements in the thermal boundary layer, etc., are all considerations important to the reactor meshing problem.Reactor meshing begins to become more specialized as the need to employ reactor simulation as a predictive design and safety analysis capability grows in importance. First, a predictive capability will require more precise physical models to be included, and these models will need to be supported by a computational science framework that will allow them to be accurately approximated both spatially and temporally during the reactor core analysis. Both the multiphysical nature of the composite reactor model and details of the physics algorithms themselves will place new requirements on the meshing process needed to support multidimensional reactor simulation. This article discusses the current state of meshing technology applied to reactor simulation and examines a set of issues that are important in the generation of high-quality reactor meshes today and in the future.     

Nuclear plant’s virtual simulation for on-line radioactive environment monitoring and dose assessment for personnel

This paper reports the use of nuclear plant’s simulation for online dose rate monitoring and dose assessment for personnel, using virtual reality technology. The platform used for virtual simulation was adapted from a low cost game engine, taking advantage of all its image rendering capabilities, as well as the physics for movement and collision, and networking capabilities for multi-user interactive navigation. A real nuclear plant was virtually modeled and simulated, so that a number of users can navigate simultaneously in this virtual environment in first or third person view, each one receiving visual information about both the radiation dose rate in each actual position, and the radiation dose received. Currently, this research and development activity has been extended to consider also on-line measurements collected from radiation monitors installed in the real plant that feed the simulation platform with dose rate data, through a TCP/IP network. Results are shown and commented, and other improvements are discussed, as the execution of a more detailed dose rate mapping campaign.     

The design of the Tokamak Physics Experiment (TPX)

The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.     

Stabilization and termination of severe accidents in LWRs

The last 20 years of research on severe accident safety for light water reactors (LWRs) has resolved a number of issues. However, the issue of melt/debris coolability is still unresolved. At stake is the stabilization and termination of a severe accident, if ever it would occur. The stabilization and termination can be established only through the coolability of the melt or the particulate debris, which are found in-vessel, or ex-vessel, depending upon the extent of the progression of a postulated accident.This paper will review the state of the art of coolability during a severe accident for the current light water reactors (LWRs). It will also review whether the accident management actions will be effective in terminating a postulated severe accident. The attention paid to the stabilization and coolability in future LWRs will be discussed and the design solutions will be evaluated.