Design of a permanent epithermal neutron irradiation site in the Syrian MNSR

A permanent epithermal neutron irradiation site was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using cadmium as a thermal neutron shielding material. This site was designed by Cd-shielding the internal surface of the outer aluminum tube of the FOIS (First Outer Irradiation Site) in the MNSR. The MCNP-4C calculations showed that, to have a permanent epithermal neutron irradiation site for the ENAA using the cadmium, it is necessary to add the top beryllium shims of the reactor to compensate for the reactivity losses due to the neutrons absorption in the cylindrical cadmium shell. The activation detectors were used to measure the thermal and epithermal neutron fluxes in the FOIS. Distribution of the thermal neutron flux along the vertical direction of the outer irradiation capsule used in the FOIS has been found using MCNP-4C code, and experimentally by irradiating five copper wires. Good agreements were obtained between the calculated and the measured results.     

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample.     

Tokamak fusion neutron spectrometer based on PXI bus

In order to realize on-line real-time measurement of dynamic and time-sharing neutron spectrum of HL-2A,a tokamak fusion neutron spectrometer based on PXI bus was developed.It consists of electronics system and eight thermal neutron detectors,namely SP9 3He proportional counter,embedded in eight polyethylene spheres in different diameters.Response function of the eight polyethylene spheres was the key to calculate the neutron spectrum accurately.In this paper,response function of the eight polyethylene spheres is simulated by adopting Geant4 code,and neutron counts from an 241Am-Be neutron source are measured by the eight detectors.The calculated spectrum of the Am-Be neutron is accurate in 0-2 MeV region,and is similar to the theoretical spectrum.The tokamak fusion neutron spectrometer was used in HL-2A device to monitor the dynamic neutron spectrum of HL-2A on-line and real-time.     

Characterization of neutron field in the experimental fast reactor JOYO for fuel and structural material irradiation test

The experimental fast reactor JOYO has been operated as an irradiation test facility for fast reactor fuel and structural material since 1983 with its MK-II core. During this time, an extensive study was conducted to characterize the neutron field in order to assure the accuracy and reliability of neutron fluence. Neutron flux for a given irradiation test was calculated using a core management code system based on three-dimensional diffusion theory. It was then corrected with the adjusted neutron spectrum by means of the multiple foil activation method. The neutron fluence calculation accuracy in the fuel region was evaluated within a 5% error by comparing the burn-up of spent fuel with the measured values, which had been obtained from their post-irradiation examination. At positions away from the fuel region, the neutron flux distribution was calculated using a two-dimensional transport code. A Monte Carlo code was also used to analyze the detailed neutron flux distribution within an irradiation test subassembly that had a heterogeneous internal structure. With the neutron flux results various irradiation parameters, such as displacement per atom (dpa) and helium production, could be evaluated. A helium accumulation fluence monitor has been developed to measure not only neutron fluence but also helium production. Neutron flux and fluence obtained from the core management calculations were compiled as a database for users’ convenience together with related irradiation information and fuel subassembly material compositions. These data are expected to be widely used in the post-irradiation analysis of fuel and structural material.     

Characteristics of Fabricated Neutron Detectors Based on a SiC Semiconductor

Wide-band-gap semiconductors such as SiC, AlN, and GaN are promising materials for harsh environment applications due to their high-temperature operation capability. Two types of PIN-type semiconductor neutron detectors based on SiC were designed and fabricated for nuclear power plant (NPP) applications such an in-core reactor neutron flux monitoring and safeguarding nuclear materials. One is for fast neutron detection and the other, which was evaporated with ⁶LiF, is for thermal neutron detection. In this study, preliminary tests, such as the determination of I-V and alpha responses, were performed. Reaction probabilities with respect to neutron energies were also calculated by using an MCNPX code for comparison with the experimental results. Responses of the neutrons were measured at the Ex-core Neutron irradiation Facility (ENF) of the High-flux Advanced Neutron Application Reactor (HANARO) research reactor at the Korea Atomic Energy Research Institute (KAERI). Pulse height spectra and count rates were measured with respect to the neutron fluxes from 1:6 × 10⁶ n/cm²·s to 1:9 × 10⁷ n/cm²·s. Also, a 0.99 root-mean-square value of linearity against the fluxes to the count rates was obtained with the fabricated neutron detectors. For a thermal neutron detector, a 3.3% detection efficiency was obtained.     

Monte Carlo simulation of the Greek Research Reactor neutron irradiation facilities

A Monte Carlo simulation of the Greek Research Reactor was carried out using MCNP-4C2 code and continuous energy cross-section data from ENDF/B-VI library. A detailed model of the reactor core was employed including standard and control fuel assemblies, reflectors and irradiation devices. The model predicted neutron flux distributions within the core in good agreement with calculations performed using the deterministic code CITATION and measurements using activation foils. The model is used for the prediction of the neutron field characteristics at the reactor irradiation devices and enables the design and evaluation of experiments involving material irradiations.     

Four energy group neutron flux distributions in the Syrian Miniature Neutron Source Reactor using the WIMSD4 and CITATION codes

A 3-D (R, θ, Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the pointwise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation sites with relative differences less than 7% and 5%, respectively.     

中子积分输运动态方程

通过理论分析给出了中子积分输运动态方程 ,发展了中子积分输运理论 ,使中子积分输运理论不仅可以用来分析反应堆栅格非均匀效应和计算反应堆参数等稳态问题 ,而且还可以处理反应堆动态问题。中子积分输运动态方程是一个多群多点 (一个空间分区为一点 )中子动态方程 ,在单群情况下就是多点反应堆动态方程。多点动态方程可以用来分析与空间有关的反应堆动态问题。介绍了中子积分输运动态方程的应用个例 ,通过中子积分输运动态方程分析了中国先进研究堆中子代时间的构成 (刚性和柔性中子代时间 )问题。     

基于广义最小二乘法原理的中子能谱解谱程序开发及验证

中子能谱是反应堆的一项重要参数,为验证理论计算,常使用阈探测器活化法对中子能谱进行实际测量。为解决由于活化片数量小于能群数量而引起的解谱问题并获得较高的解谱精度,基于广义最小二乘法原理开发了解谱程序NSAGLS。使用IAEA给出的例题进行了验证,结果表明该程序能够得到正确的解谱结果。该程序可用于反应堆中子能谱的解谱,并能够考虑输入谱、核反应截面及测量活度不确定度信息对解谱结果的影响。     

A new unfolding code combining maximum entropy and maximum likelihood for neutron spectrum measurement

We present a new spectrum unfolding code, the Maximum Entropy and Maximum Likelihood Unfolding Code (MEALU), based on the maximum likelihood method combined with the maximum entropy method, which can determine a neutron spectrum without requiring an initial guess spectrum. The Normal or Poisson distributions can be used for the statistical distribution. MEALU can treat full covariance data for a measured detector response and response function. The algorithm was verified through an analysis of mock-up data and its performance was checked by applying it to measured data. The results for measured data from the Joyo experimental fast reactor were also compared with those obtained by the conventional J-log method for neutron spectrum adjustment. It was found that MEALU has potential advantages over conventional methods with regard to preparation of a priori information and uncertainty estimation.     

The ITER radial neutron camera: An updated neutronic analysis

The radial neutron camera (RNC) will provide the spatial distribution and the total strength of the ITER neutron source (emissivity profile and fusion power) by means of collimated neutron measurements. Line-integrated neutron spectral measurements can also provide information on the ion temperature profile. The present design of the RNC consists of two collimating structures for a full coverage of the plasma: 36 collimated lines of sight (LOS) distributed in three different planes view the plasma core (ex-port system) and nine collimated LOS view the plasma edge (in-port system).The RNC design is based on the combined use of the MCNP Monte Carlo code and a software tool performing asymmetric Abel inversion of simulated measured neutron signals (MSST). Neutron and γ-ray transport calculations are performed with MCNP using a 3D RNC model to determine the signal/noise for each RNC channel and the spectra at the detectors. The MSST code is used to check the RNC compliance with the ITER measurement requirements for the neutron emissivity profile.In the present paper the improvement of the hard variance reduction technique applied to the MCNP neutron source (consisting in sampling neutrons only from plasma regions contributing to the detector signal) is presented and the following issues are analyzed: the possibility of reducing the length of the ex-port collimators (resulting in a significant reduction of the overall RNC dimension and weight); options for the reduction of the dose due to the neutron streaming through the RNC cut-outs in the blanket shielding module; the integration of a γ-ray detection system in the RNC by partially filling the collimators with a neutron absorbing material (LiH).     

原型微堆辐照座物理特性参数模拟测定

文章给出了原型微堆辐照座同的某些物理特性参数；相对中子通量密度分布，绝对中子通量密度，能谱能数（镉比、超热指标和中子温度），某些样品在辐照座内对反应性的影响以及各辐照座之间的相互关系，实验研究在原型微堆的零功率实验装置上完成。     

Comparative assessment of theoretical and experimental measurement of photo neutron flux level in GHARR-1 after shutdown

Direct photo-neutron source strength was evaluated for the Miniature Neutron Source Reactor (MNSR) in subcritical condition in the GHARR-1 facility. Two different static methods were applied for comparison. A theoretical method based on the use of MCNP code and an experimental method based on foil activation technique. The latter has been found to be most convenient method for neutron flux measurement. The method depends only on the activity of a bare and cadmium covered foil if the irradiation positions are known. Photo-neutron flux level was determined theoretically using MNCP after measuring neutron flux at shutdown; and experimentally using Neutron Activation Analysis (NAA) technique also at shutdown with great care. The values obtained from the theoretical and experimental measurements are tabulated in Table 2. The results recorded were validated using biological peach leave and a geological rock sample. The results after validation for Mn concentration in the samples were 87 ± 1 μg/g and 432 ± 23 μg/g, respectively. Results for the two methods were in good agreement. Realization of photo-neutron source existence due to beryllium reflector was also experienced.     

Spark counting of particle damage tracks in neutron studies

Conclusion As shown by trial experiments, neutron detectors based on fissile nuclides and thin Lavsan foils can be used successfully in combination with a spark counter of tracks in a variety of neutron studies: measurements of the spatial distributions of the spectral characteristics of neutron fields of various kinds; measurements of integral and differential fission cross sections; in personal neutron dosimetry; etc. This method ensures that information from a large number of neutron detectors is obtained rapidly.Translated from Atomnaya Énergiya, Vol. 61, No. 1, pp. 35–40, July, 1986.     

Optimization of moderator assembly for neutron flux measurement: experimental and theoretical approaches

A moderator of paraffin wax assembly has been demonstrated where its thickness can be optimized to thermalize fast neutrons. The assembly is used for measuring fast neutron flux of a neutron probe at different neutron energies, using BF03(U10and 200) and3He(U0.500)neutron detectors. The paraffin wax thickness was optimized at 6 cm for the neutron probe which contains an Am–Be neutron source. The experimental data are compared with Monte Carlo simulation results using MCNP5 version 1.4. Neutron flux comparison and neutron activation techniques are used for measuring neutron flux of the neutron probe to validate the optimum paraffin moderator thickness in the assembly. The neutron fluxes are measured at(1.17 ± 0.09) 9 105 and(1.19 ± 0.1) 9 105n/s, being in agreement with the simulated values. The moderator assembly can easily be utilized for essential requirements of neutron flux measurements.     

Some Studies on the Time Resolved Energy Spectrum Reconstruction for Pulsed Neutron Source

The purpose of this article is to present some new numerical results concerning the time resolved energy spectrum reconstruction of a short pulsed neutron source, like one created in plasma focus devices. An MCNP code is used to simulate a time dependent neutron source, plastic scintillator detectors, and their recorded signals. A Monte Carlo program reconstructs spectrum using the signals produced by MCNP code. By the numerical computations we determined the optimum number of four detectors which are needed to be placed at 0, 10, 18 and 30 m from the source, respectively. A time resolution about 12 ns and an energy resolution of 40 keV are obtained. Neutron scattering in the air is considered, and it is found that the intensity of spectrum is increased by 6%.     

多方向加权方法在D-T聚变中子能谱测量中的应用

中子能谱是研究和诊断核反应过程特性最重要的特征量之一。建立了一种新的多方向加权方法用于D-T聚变中子能谱测量:在反冲质子出射方向的多个不同角度上,同时布置探测器,最终的中子能谱由各方向所获取的反冲质子能谱结合相应的权重值确定。Geant4模拟结果显示,多方向加权方法可以提升探测效率和求解结果精度。利用多方向加权方法对高斯分布中子源以及实际的D-T聚变中子能谱进行测量模拟与求解分析,分析结果验证了该方法的可行性和有效性。     

Availability of MCNP & MATLAB for reconstructing the water-vapor two-phase flow pattern in neutron radiography

The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008. In this paper, we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the tube were obtained using the MCNP code without influence of γ-ray and electronic-noise. The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated. The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI. The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques. And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.     

中国实验快堆中子能谱测量实验研究

中子能谱是反应堆的一项重要参数,在快堆中,中子能谱直接决定其增殖与嬗变性能。中国实验快堆是我国第一座钠冷快中子堆,需测量其中子能谱。本文利用活化法在堆芯两个位置进行辐照实验,利用解谱程序处理得到这两个位置的中子能谱。实验结果表明,两个位置的中子能谱与理论计算值基本一致。     

单球中子谱仪的能量响应模拟及解谱方法研究

本文阐述了单球中子谱仪的原理,介绍了基于单慢化球和19对⁶Li-⁷Li闪烁体探测器构成的单球中子谱仪的结构及解谱方法,使用蒙特卡罗中子输运程序模拟了单球中子谱仪的中子响应函数。计算结果表明,该谱仪具有较好的空间对称性,能根据谱仪中各探测器的计数对源的大致方位进行判断;模拟了单球谱仪在²⁴¹Am-Be源照射下各探测器的计数,使用Unfolding with Maxed and Gravel (UMG)解谱程序在不同解谱算法以及初始谱的情况下对模拟数据进行解谱计算,在使用最大熵散发以及与源项相同的预置谱的情况下,解谱结果最为准确,验证了响应函数的准确性。