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.     

Comparative experimental and theoretical investigations of the DM neutron moisture probe

Theoretical and experimental investigations of the Danish produced DM subsurface moisture probe have been carried out at the Research Establishment Risö, and the results obtained are presented in this paper. The DM probe contains an Am-Be fast neutron source and has a glass scintillator containing ⁶Li as neutron detector. Through modification of the probe it was possible to vary the source-detector distance in steps. By the use of a three-group diffusion model and a semi-empirical calibration procedure good agreement between measured and calculated calibration curves was obtained. Further the model has been used to study the effect of variations in the dry soil density and in the boron content of soil. Finally the applicability of the model in connection with probe design has been considered.     

CR-39 detector based thermal neutron flux measurements, in the photo neutron project

PhoNeS (photo neutron source) is a project aimed at the production and moderation of neutrons by exploiting high energy linear accelerators, currently used in radiotherapy. A feasibility study has been carried out with the scope in mind to use the high energy photon beams from these accelerators for the production of neutrons suitable for boron neutron capture therapy (BNCT). Within these investigations, it was necessary to carry out preliminary measurements of the thermal neutron component of neutron spectra, produced by the photo-conversion of X-ray radiotherapy beams supplied by three LinAcs: 15 MV, 18 MV and 23 MV. To this end, a simple passive thermal neutron detector has been used which consists of a CR-39 track detector facing a new type of boron-loaded radiator. Once calibrated, this passive detector has been used for the measurement of both the thermal neutron component and the cadmium ratio of different neutron spectra. In addition, bubble detectors with a response highly sensitive to thermal neutrons have also been used. Both thermal neutron detectors are simple to use, very compact and totally insensitive to low-ionizing radiation such as electrons and X-rays. The resultant thermal neutron flux was above 10⁶ n/cm²s and the cadmium ratio was no greater than 15 for the first attempt of photo-conversion of X-ray radiotherapy beams.     

加速器中子源的中子注量测量方法

在用静电加速器中子源标定探测器的中子灵敏度实验中,采用“BF3长计数管 定标器”系统过渡,用^197Au中子活化分析方法达到了对中子注量在线、绝对监测的目的。这种方法给出与加速器束流不同角度、不同距离处的中子注量。介绍了这种中子注量测量方法。     

Direct measurement of mechanical momentum of a neutron flux

Translated from Atomnaya Énergiya, Vol. 66, No. 1, pp. 48–49, January, 1989.     

能谱变化对239Pu裂变室测量快中子注量的影响

热中子和共振区的中子在快中子临界装置中所占的份额很小,但是由于其相对大的截面,在慢化物存在的情况下,热中子和共振中子份额的微小变化,对^239Pu裂变室测量中子注量的结果影响很大。通过测量^239Pu裂变电离室在包镉和包硼、周围有无慢化物等情况下的反应率,Au、In活化片的镉比,S活化片在能谱变化下与^239。Pu的反应率比等,分析了快中子临界装置中热中子和共振区中子的分布,讨论了中子能谱变化对^239Pu裂变室测量快中子注量的影响及解决办法。     

The design of the J-type neutron flux distribution measurement system

This paper introduces the design of a J-type aeroball system that the tube penetrates the lateral wall of reactor pressurized vessel (RPV), then immediately goes down to the vessel bottom and then goes up through the lower core support plate into the reactor core. Some experimental results related to gas flowing within a thin tube are presented in the paper, such as the gas friction drag coefficient on the ball’s way and etc. From theoretical and experimental viewpoints, the feasibility of the system is proved in pneumatic holding-up and β measurement aspects. In order to ensure an enough ratio of signal to noise, the maximum distance between two measuring points in water reactors is given. The paper gives out the measuring number per two-assemblies width N_i=Int(n_i/2+0.51), which is the accuracy relation between the number of fuel assembly and the minimum one of measuring point to only reconstruct 2D neutron flux distribution completely by the measured data.     

利用金属钆测量热中子通量的实验研究

研究了一种测量热中子通量的新方法,利用金属钆与热中子反应产生的次级γ射线来确定热中子通量。使用两个NaI探测器进行符合测量,设计了对伽玛射线和散射中子的良好屏蔽体。经过对本底和钆样品的多次测量,对钆与中子反应产生的伽玛能谱进行了分析,计算出了中子与钆反应产生的计数率(样品净计数率),其最大值达到2.74 Hz,从而证实了这种方法的可行性。     

The fast neutron energy spectrum and the thermal neutron flux distribution in the experimental channel of the VVR reactor

Using threshold detectors of aluminum, indium, iron, titanium, zinc, silicon, magnesium, mercury, and sulfur, we have measured the fast-neutron energy distribution in the experimental vertical channel of the VVR-2 reactor, traveling in the immediate vicinity of the active zone. We have obtained the spectral composition of the fast neutrons in the center of the channel at the level of the middle of the active zone, the thermal neutron flux and its distribution in the channel. The activity of all detectors except the sulfur and copper detectors was determined from the -radiation. The obtained experimental data are compared with the results of the theoretical calculation.     

FPGA-based position reconstruction method for neutron beam flux spatial distribution measurement in BNCT

A new measurement method for the spatial distribution of neutron beam flux in boron neutron capture therapy(BNCT)is being developed based on the two-dimensional Micromegas detector.To address the issue of long processing times in traditional offline position reconstruction methods,this paper proposes a field programmable gate array based online posi-tion reconstruction method utilizing the micro-time projection chamber principle.This method encapsulates key technical aspects:a self-adaptive serial link technique built upon the dynamical adjustment of the delay chain length,fast sorting,a coordinate-matching technique based on the mapping between signal timestamps and random access memory(RAM)addresses,and a precise start point-merging technique utilizing a circular combined RAM.The performance test of the self-adaptive serial link shows that the bit error rate of the link is better than 10-12 at a confidence level of 99％,ensuring reliable data transmission.The experiment utilizing the readout electronics and Micromegas detector shows a spatial resolution of approximately 1.4 mm,surpassing the current method's resolution level of 5 mm.The beam experiment confirms that the readout electronics system can obtain the flux spatial distribution of neutron beams online,thus validating the feasibility of the position reconstruction method.The online position reconstruction method avoids traditional methods,such as bubble sorting and traversal searching,simplifies the design of the logic firmware,and reduces the time complexity from O(n2)to O(n).This study contributes to the advancement in measuring neutron beam flux for BNCT.     

On comparison of experimental and calculated neutron energy flux spectra at miniature neutron source reactor (MNSR)

Neutron energy spectrum in Miniature Neutron Source Reactor (MNSR), called Pakistan Research Reactor (PARR-2), is measured employing threshold neutron activation detectors. The calculated neutron spectrum was obtained through modeling the core in detail in three-dimensions employing the transport theory based code WIMS-D/4 and the diffusion theory based code CITATION which was also used as pre-information in the adjustment procedure. A Number of threshold detectors in the form of thin foils are used for spectrum measurements. Gamma activity of irradiated foils was measured with the help of a gamma spectroscopic system consisting of a high efficiency HPGe detector and 8000 channels PC based multi-channel analyzer. STAYNL computer code supplied by International Atomic Energy Agency (IAEA) was used for neutron spectrum adjustment. The group cross-section values and their covariance matrices were derived from the data given in preprocessed cross section libraries in ENDF–6 format of IRDF-90/NMF-G. The comparison between theoretical and experimental work shows good agreement.     

用于CSNS反角白光中子能谱及注量率测量的快裂变室

即将建成的中国散裂中子源(China Spallation Neutron Source,CSNS)反角白光中子束线可为核数据测量提供高注量率的脉冲白光中子束流,填补我国核数据测量用白光中子源的空白,提高我国核数据测量水平,满足核能、核技术及基础核物理研究对核数据的需求。该束线建成后,其中子能谱及注量率的精确测量将是开展其它物理实验的基础,快裂变电离室因其独特优点被选为中子能谱和注量率测量探测器。通过实验研究了快裂变电离室的粒子分辨性能、时间分辨性能;确定阴、阳极的合理间距为10 mm,据此测得电离室的时间分辨约15 ns;利用235U样品量计算的探测效率与利用伴随粒子法给出的探测效率在不确定度范围内符合,因此可以标定快裂变室的探测效率。通过这些工作,完成了满足反角白光中子束能谱及注量率测量需求的快裂变室的物理设计。     

Attenuation of pile neutron flux in polyethylene

The relaxation lengthsofresearch reactor neutrons were measured in polyethyIene. The findings are in excellent accord with data computed theoretically using the method of moments. The relaxation lengths are 15% less than in the case of water.Translated from Atomnaya Énergiya, Vol. 15, No. 1, pp. 17–20, July, 1963     

Cold neutron interaction with a classical electric field: Some basic theoretical and experimental considerations

We explore the feasibility of performing an experiment to measure the interaction of cold neutrons with a given classical electric field. Bound and scattering states could be detected by means of an approximate Aharonov-Casher configuration. The theoretical background is presented and then some primary elements for building a neutron detector of this nature are proposed.     

Evolution of the neutron sensor characteristics in the RBMK-1500 reactor neutron flux

Variation of characteristics of the RBMK-1500 reactor radial neutron flux sensors with the HfO₂ emitter during long-term maintenance was investigated. The influence of nuclear fuel enrichment and burnable erbium admixtures on the energy neutron spectrum, neutron absorption, and hafnium isotopic composition variation was considered. The dependences of corrective factors of the neutron sensor signal on the nuclear fuel burnup depth and the integral current accumulated by the sensor for different enrichment nuclear fuel are presented in the work. The experimental verification of the calculated dependence of the sensor corrective factor on the accumulated integral current was performed.     

Polynomial approximation of neutron flux in first-collision probability method

Conclusions The generalized first-collision method may prove useful in calculating channels with complicated geometry — e.g., RBMK channels. Then polynomials of different order may be used in different zones, taking into account, where necessary, the curvature of the neutron distribution by means of quadratic terms.Translated from Atomnaya Énergiya, Vol. 48, No. 2, pp. 80–84, February, 1980.