pMOS剂量计的温度效应及其补偿探讨

利用恒流注入法和Ｉ－Ｖ测试方法研究了用于ｐＭＯＳ剂量计的国产ｐＭＯＳ ＲＡＤＦＥＴ的温度效应，结合ＭＯＳ晶体管阈值的电流－电压方程的理论公式推导分析了温度对阈电压，沟道载流子迁移率的影响，对“零温度系数点”从理论上进行了讨论，探讨了辐射对“零温度系数点”的影响。     

Effects of package geometry, materials, and die design on energydependence of pMOS dosimeters

This paper presents the results of further studies of dose enhancement in dual and single-dielectric pMOSFET dosimeters for various package and die designs. Eight different MOSFET designs and package types were investigated over a photon energy range from 14 to 1250 keV. Seven X-ray effective energies and two radioactive sources of cesium and cobalt provided the radiation. As in a previous study, Rutherford back-scattered electrons were primarily responsible for the dose enhancement factors which achieved values as high as 20. Packages filled with silicon grease, aluminum oxide, or paraffin eliminated the contribution of back-scatter to the enhanced dose. These modifications allowed measurements of the usual dose enhancement at the aluminum or polysilicon gate-silicon nitride (dual dielectric devices), or silicon dioxide interfaces (single dielectric parts), and at the silicon nitride-silicon dioxide interface. In addition to the primary peak in the DEF (dose enhancement factor) curve versus energy at 45.7 keV, there is a second peak at about 215 keV. This peak might be due to enhancements at the interfaces of a MOSFET. These interface effects were small in the single-insulator parts in standard ceramic packages, and significantly larger in the dual-insulator devices. The effects were reduced by filling the packages with the materials as previously described. The geometry of the package, for example, the size of the air gap between the die's surface, and the lid of the package impacts the value of the DEF     

pMOSFET多管级联结构辐照响应特性研究

ｐＭＯＳＦＥＴ剂量计多管级联结构电离辐射响应持性比单管能明显提高辐射响应灵敏度。多管共衬底级联与不共衬底级联灵敏度提高倍数较大。比较了多管共衬底级联结构在两种辐照偏置条件下的辐照响应差异,实验结果表明,辐照时,保持恒流注入条件,其响应灵敏度、线性度和稳定性均高于零偏置的结果。同时,研究了多管级联结构完全退火后的二次辐照响应,结果表明,响应灵敏度与线性度均高于第１次辐照的。     

A finite transform approach to the solution of radiation transport equation

Starting from the integral form of transport equation and defining a new transform G(x,μ,μ_m), the integration over space is carried out analytically resulting in an integral equation for F in μ. Use of Legendre polynomial for the solution of this equation is explored. The method is applicable to practical problems of radiation transport in multiregion, energy dependent systems with arbitrary degree of anisotropy. Its simplification for idealised systems and comparison with earlier work are indicated.     

A constraint-based approach to intelligent support of nuclear reactor design

Constraint is a powerful representation to formulate and solve problems in design; a constraint-based approach to intelligent support of nuclear reactor design will be proposed in this paper. We will first discuss the features of the approach, and then present the architecture of a nuclear reactor design support system under development. In this design support system, the knowledge base contains constraints useful to structure the design space as object class definitions, and several types of constraint resolvers are provided as design support subsystems. The adopted methods of constraint resolution will be explained next in detail. The usefulness of the approach will be demonstrated using two design problems: design window search and multiobjective optimization in nuclear reactor design.     

光子剂量计对14.5MeV中子的相对灵敏度(K_U)的测定

在n-γ混合场中,普遍采用双探测器测量生物组织的中于吸收剂量。当用组织等效电离室(T)和光子剂量计(U)组合,测量中子吸收剂量(D_n)和伴随的γ剂量(D_γ)时,将有如下方程组:     

Thermo-mechanical loadings of primary structures in accidental conditions, and an approach to solution by limit design

We first describe the static thermo-mechanical loadings of some important structures due to accidental situations. We calculate the behaviour of these structures by several methods; elastic and plastic analysis, and limit design. We show the interest of this last method.     

Fender: A finite element code for the solution of the diffusion equation in shield design applications

Diffusion theory remains an important method of calculation for shield design. Using adjusted coefficients the method provides an inexpensive solution of adequate accuracy for survey and optimization studies in two- and three-dimensional geometries. Solved in the adjoint mode, the method provides an estimate of the importance function which may be used for the acceleration of generalized-geometry Monte Carlo calculations.A number of computer codes exist to solve the diffusion equation by a finite difference approximation in one-, two- and three-dimensions. The mesh systems used in such codes usually impose restrictions on the accuracy of representation of shields with complicated geometries.The computer code FENDER solves the diffusion equation for neutron or gamma transport using the finite element technique. At present the code is written for a two-dimensional problem in which the geometry is specified as an array of triangular or rectangular elements. This permits a good representation to be made of shields containing curved surfaces. The variation of the calculated particle fluxes within an element is assumed to be quadratic.FENDER may take details of the element structure from an external mesh generating package but also contains a semi-automatic mesh generating routine for use as a stand-alone code. Multigroup diffusion parameters may be either input directly or generated from material compositions. The code is capable of handling problems with at least 1000 elements which is roughly equivalent in size and attenuation to 10,000 finite difference meshes. A variety of boundary conditions may be specified.The paper includes an example of application to demonstrate the potential usefulness of the method and the code. The case chosen is the calculation of neutron fluxes in a stylized fast reactor.     

A solution for the integration of finite element analysis in a ship design environment

The DEMAIN system for the pre- and post-processing of finite element analyses of ship structures is presented. It is shown that this new modelling concept, although being self-contained and specialized, has features which relate it to computer-aided design applications of a more general nature. Thus, compared to other finite element pre/post-processors, it allows a more natural occurrence of the structural analysis task in the design flow and can be considered a major step towards an integrated design and analysis system.     

A new approach to the finite-element formulation and solution of a class of problems in coupled thermoelastoviscoplasticity of crystalline solids

The application of the general thermoviscoelastic-plastic theory to the analysis of the coupled thermomechanical response of a class of crystalline solids is considered. After a brief review of the essential features of the theory, the manner in which finite-element models of the generalized plastic behaviour can be formulated is discussed. Available experimental data are used to solve a representative example problem in transient, nonlinear, coupled thermoviscoelastoplasticity of a three-dimensional solid of aluminum.     

Fast reactor decay heat removal: approach to the safety system design in Japan and Europe

In the framework of the cooperation on fast reactor between the European and Japanese electrical utilities, the design companies responsible for the demonstration fast breeder reactor (DFBR) in Japan and the European fast reactor (EFR) have performed a comparative evaluation of the safety qualified decay heat removal systems of the two reactor designs. At the level of overall safety and concept design there is an obvious similarity between the two DHR systems. In both cases heat is removed directly from the reactor vessel primary sodium by systems designed according to a similar deterministic methodology, with a probabilistic assessment performed to demonstrate achievement of the required reliability. Nevertheless, the evaluation revealed a number of differences resulting from different national practices. These include the application of diversity and redundancy philosophy, the extent of passivity taken into account, the consequences of postulated maintenance outage on the design and the decay heat curve.     

A spatially adaptive grid-refinement approach for the finite element solution of the even-parity Boltzmann transport equation

A spatially adaptive grid-refinement approach has been investigated to solve the even-parity Boltzmann transport equation. A residual based a posteriori error estimation scheme has been utilized for checking the approximate solutions for various finite element grids. The local particle balance has been considered as an error assessment criterion. To implement the adaptive approach, a computer program ADAFENT (adaptive finite elements for neutron transport) has been developed to solve the second order even-parity Boltzmann transport equation using K⁺ variational principle for slab geometry. The program has a core K⁺ module which employs Lagrange polynomials as spatial basis functions for the finite element formulation and Legendre polynomials for the directional dependence of the solution. The core module is called in by the adaptive grid generator to determine local gradients and residuals to explore the possibility of grid refinements in appropriate regions of the problem. The a posteriori error estimation scheme has been implemented in the outer grid refining iteration module. Numerical experiments indicate that local errors are large in regions where the flux gradients are large. A comparison of the spatially adaptive grid-refinement approach with that of uniform meshing approach for various benchmark cases confirms its superiority in greatly enhancing the accuracy of the solution without increasing the number of unknown coefficients. A reduction in the local errors of the order of 10² has been achieved using the new approach in some cases.     

A new paradigm for core design aimed at the sustainability of nuclear energy: The solution of the extended equilibrium state

The future expansion of nuclear energy, a technology identified as one of the main candidates for reducing the world’s dependence on fossil fuels, requires a thorough analysis of the sustainability of this energy source for long-term supply. Generation-IV nuclear systems could represent a turning point for energy production by minimizing the environmental footprint of the fuel cycle. A new paradigm is thus required for reactor design, focusing, at the core design level, on both the closure of the fuel cycle and the effective utilization of natural resources.     

Nuclear data requirements for the ADS conceptual design EFIT: Uncertainty and sensitivity study

In this paper, we assess the impact of activation cross-section uncertainties on relevant fuel cycle parameters for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) with a “double strata” fuel cycle. Next, the nuclear data requirements are evaluated so that the parameters can meet the assigned design target accuracies. Different discharge burn-up levels are considered: a low burn-up, corresponding to the equilibrium cycle, and a high burn-up level, simulating the effects on the fuel of the multi-recycling scenario.     

A systematic approach to the radiation damage problem

To assess the suitability of a material for use as a core component in a fast reactor or for the first wall in a fusion reactor, it is necessary to know the irradiation damage behaviour of the material outside the usual materials testing data domain. In the present paper we propose a strategy based on a closely co-ordinated programme of experimental and theoretical research. The aim of this strategy is the systematic construction of a physically based model of the evolving damage structures. This would then allow both the necessary extrapolations of the data to the desired conditions to be achieved in a reliable fashion and provide a rational basis for the development of low-swelling alloys for the two nuclear systems.     

Safety approach in the EU test blanket systems design

This paper reveals the safety strategy and approach developed and followed in the design of the two EU TBS describing its objectives, components and implementation. Addressing the safety in the early stage of the conceptual design of nuclear facilities is a well recognized international practice and industrial project-level requirement for the successful completion of the licensing process within expected project cost and schedule. The impact of the early development of the safety approach, its implementation and monitoring in the design of nuclear device like the TBS is not limited to the safety assessment and licensing activities only. Safety approach plays indispensible role in reducing the overall project risk. It infiltrates the entire design process through the unavoidable interfaces between the design features and its safety level. In reality the entire process of the TBS development, design, technological demonstration and implementation is affected by the project team safety culture.