Neutron transport benchmark problem proposal for fast critical assembly without homogenizations

In the present paper, we propose a neutron transport benchmark problem for fast critical assembly without homogenizations. With this problem, we can validate applicability of neutron transport codes when employed in highly heterogeneous fast critical assembly analyses. In addition, this benchmark problem can be used to validate homogenization procedures for slab lattices.     

Integral transport in a three region sphere with associated benchmark problems

We derive, by the method of characteristics, an integral form of the transport equation with isotropic scattering for spherical symmetry in which the spatial dependence of the cross-sections is arbitrary. The sources of particles (photons or neutrons) can be either incident on the surface with an arbitrary angular distribution or distributed within the body. A specific example is given for the case of a three region sphere in the form of a central spherical core surrounded by a concentric shell followed by an outer shell. By allowing the total cross-section to go to zero in any one of the regions we may evaluate the effect of voids. Some numerical results are given which enable the limitations of void representation in a standard computer code to be highlighted. We also consider the case of a multiplying system and calculate the effect of an annular void on the multiplication factor.     

Neutron transport in a reactor system with temperature feedback

When the effect of temperature feedback in a reactor system is considered the neutron transport equation for the neutron density is supplemented by a temperature equation which is a partial differential equation of parabolic type if heat conduction is taken into consideration. This consideration leads to a coupled system of nonlinear partial integro-differential equations. The aim of this paper is to present an iterative scheme for the determination of the solution of the nonlinear coupled system and to establish some qualitative property of the solution. The iterative scheme consists of two monotone sequences which converge monotonically from above and below, respectively, to a unique solution. The qualitative aspect includes the existence and uniqueness of a positive solution, upper and lower bounds of the solution and stability of a steady-state solution.     

Neutron transport for pure-triplet scattering in finite planar media with reflective boundaries

Pure-triplet scattering in neutron transport through a finite plane-parallel medium with internal source of energy is considered. The medium is assumed to have specular- and diffusely-reflecting boundaries. The neutron partial heat fluxes for this problem are computed in terms of the albedos of the source-free problem. Pomraning–Eddington approximation is used to solve the source free problem. A weight function is introduced to force the boundary conditions to be fulfilled.     

Neutron Activation Analysis with Low Neutron Flux

Procedures are described for identifying and determining concentrations of trace impurities in minerals by means of neutron activation analysis with low neutron flux and standard gamma measurement equipment. Techniques developed in this work may be of use to those involved with the ever increasing need for mineral assays and for pollutant measurements, but who do not have the latest and most sophisticated equipment. This is especially true for institutions without formal nuclear programs and for developing countries desiring to be self-sustaining.     

考虑6组缓发中子效应的中子倍增公式

导出了 6组缓发中子效应的中子倍增公式     

Neutron transport through semi-infinite continuous stochastic media using Gaussian statistics

The stationary solution of the one-speed neutron transport equation in a semi-infinite stochastic medium with linear anisotropic scattering is considered. The cross-section function of the medium is assumed to be a continuous random function of position with fluctuations about the mean taken as Gaussian distributed. The joint probability distribution function of these Gaussian random variables is used to calculate the ensemble-averaged quantities, such as radiant neutron energy and net neutron flux, for an arbitrary correlation function. The problem is solved at first in the deterministic case, then the solution is averaged using Gaussian joint probability distribution function. A modified Gaussian probability distribution function is also used to average the solution. Numerical results are given for the sake of comparison.     

Validation of neutron transport calculations in benchmark facilities for improved damage fluence predictions

Neutron transport calculations play a vital role in the determination of the amount of radiation damage accumulated by reactor pressure vessels, and hence are an important factor in establishing reactor lifetime. Benchmark experiments are required to evaluate the validity and limitations of transport calculations. A comprehensive benchmark program is currently supported by the US NRC and several European organizations to test various aspects of the overall fluence calculational procedure. This paper summarizes the needs and requirements of the overall benchmark program. Results and conclusions of some current benchmark experiments are reviewed. Finally, some approximations in transport calculations which need validation, but which are not addressed in current benchmark efforts, are discussed.     

MTR benchmark static calculations with MCNP5 code

With the sustained development in computer technology, the use of more powerful computational tools becomes mandatory. The challenge today is to revisit safety features of the existing nuclear research reactors using new generation of computer tools. The objective is to verify that the safety requirements still met and when necessary to introduce some amendments coming from the new attainments. In the current paper the IAEA safety-related nuclear research reactors (RR) benchmark problem is reconsidered. The idea consists in performing static calculations of the benchmark using the last version of the MCNP5 code. This later offers updated code models and cross-section library. The results are afterwards compared with previous calculations and discussed.     

Two-dimensional whole-core transport calculationsof the oecd benchmark problem C5G7 MOX by the CRX code

The OECD 2-D benchmark problem C5G7 MOX was calculated by the CRX code which is based on the method of characteristics. A modular ray tracing scheme and parallel computation in angular domain are implemented to reduce computer memory and computation time. In this paper, we present results of calculations, including sensitivity studies performed with varying values for some calculation parameters.     

Assessment of Neutron Cross Sections for Titanium with Neutron Spectrum in Titanium Pile

The measurement of angular neutron spectrum in a quasi-spherical pile of Ti was carried out by the linac time-of-flight method for the assessment of neutron cross sections for Ti in the energy range from a few keV to a few MeV. The measured spectrum in the pile is generally in good agreement with the calculated one from ENDF/B-IV (MAT = 1,286 for Ti) except in the energy range from about 60 keV to a few 100 keV, where the calculation gives considerably lower neutron flux than the measurement.

In order to investigate the cause of this discrepancy between the measured and calculated spectra, the total cross sections for Ti were measured by the transmission method. The results give larger values of total cross sections for Ti by about 30% than ENDF/B-IV below 200 keV, and smaller values by about 10% above 200 keV. These results were ensured at 55 and 147 keV by the measurement using a Si-filtered neutron beam. The calculation based on the measured cross sections shows better agreement with the measured spectrum than that based on ENDF/B-IV. The discrepancy is still observed around 100 keV.

The sensitivity analysis shows the importance of cross sections above 1 MeV and elastic cross sections in the resonance energy region to solve the disagreement between the measured spectrum and the calculated one.     

小型中子源高能中子照相装置准直屏蔽系统设计

准直屏蔽系统是中子照相装置的必要设备。本文采用蒙特卡洛中子输运程序(MCNP)等软件对高能中子准直屏蔽系统进行理论设计,初步确定了其材料构成和外观尺寸,从理论上确定了装置包括成像处注量率、成像距离及相应视场等关键参数。     

Neutron Dosimetry with Boron-Doped CR-39 Plastic

As a highly sensitive, simple and non-radioactive neutron dosimeter, CR-39 plastic plates doped with a boron compound: ortho-carborane were prepared. After thermal neutron irradiation, the plates were etched in an aqueous solution of 30%KOH, at 60°C for 2 to 16 h. The etch-pits generated by ¹⁰B(n, α)⁷Li reactions were then counted using an optical microscope or an automatic track counting system. The density of the etch-pits on an irradiated plate increases with the etching time. When the etching time is kept constant, the etch-pit density is proportional to the irradiated thermal neutron fluence. The proportional constant is termed “sensitivity”, which is 4.2 x 10^?4 for a plate containing ortho-carborane at a concentration of 0.5% by weight and for etching time of 16 h. By considering background counts, a thermal neutron dose of 0.025 mSv can be measured with this plate. The plates are insensitive to visible-, UV-, X-, β- and γ-rays and are easy to handle because the detector and converter are incorporated. There is no possibility of underestimating the dose equivalent due to fading. Furthermore, the isotopes of boron are not radioactive and thus are radio- logically safe.