Energy deposition distributions caused by fast neutrons in converter type silicon detectors

Polyethylene converter silicon detectors have been developed as part of a dosimetry system for the application in neutron fields. Signals have been investigated which are mainly caused by reactions occurring when neutrons hit the PE converter or the silicon diode directly. In this work, neutron interactions in the components of a PE converter silicon sensor have been calculated to determine energy deposition distributions for neutron energies from 1.2 v MeV to 14.8 v MeV. Experiments in quasi-monoenergetic neutron fields have been performed. For the simulation of the neutron interactions with the detector layers, the GSF neutron interaction code NISD has been applied. The transport of ions has been calculated separately by means of the program TRIM. It has been shown that by the use of these concepts and models, a rather good agreement of measured and computed pulse height distributions can be obtained and, consequently, the response of converter type detectors to neutron radiation can be determined.     

Modeling the efficiency of neutron detection scintillators in the energy range of 5–70 MeV

The instrumental functions for neutron godoscope detectors used in studying the nd-breakup reaction are obtained. The effect on light output of contributions from different processes that occur in a scintillator under the effect of neutrons is examined at different energies of the incident neutrons. The relationship between the detector??s efficiency and neutron energies ranging from 5 to 70 MeV is obtained for different thresholds.     

Measurement of cross sections for inelastic cold-neutron scattering in metals and polymers by the method of (<Emphasis Type="Italic">n, γ</Emphasis>) analysis

The results obtained by measuring the cross sections for the inelastic scattering of very cold neutrons for a number of metals and polymers by the method of a neutron-irradiation analysis are presented. The method is based on simultaneously measuring events of inelastic scattering and neutron capture in the sample under investigation via recording gamma radiation with a semiconductor germanium detector. Neutron capture by a nucleus of the sample is accompanied by the prompt radiation of gamma rays having a known spectrum. Upon inelastic scattering, a neutron acquires thermal energy. Upon leaving the sample, this neutron is absorbed in a special converter that contains the isotope ¹⁰B. The capture of the neutron by a ¹⁰B nucleus is followed by the emission of a 477-keV gamma ray. The probabilities of capture and inelastic scattering are proportional to the respective neutron-interaction cross sections, and the ratio of the recorded detector counts corresponding to events of the two types does not depend on the spectrum of the incident flux of very cold neutrons or on the trajectory of neutron motion in the sample. The sought inelastic-scattering cross section at a fixed sample temperature is calculated by using this ratio and the known cross section for neutron capture by the sample isotope having a known gamma-radiation spectrum.     

Investigation of Fe+3 and Fe+2 properties in calcium aluminoborate glasses

Two types of silicon detectors are reported (ion implantation and surface barrier detector) for the detection of neutrons in high gamma-background. The purpose is to find an optimal design of silicon detector combined with the neutron converter, which is efficient enough to neutron flux and with low sensitivity to %-rays. The experiments show that for fabrication of effective Si detectors for neutrons (using the (n,!) reactions) in the presence of an intensive %-field one must optimise the resistivity of the Si substrate and the active area of the detector, at which the detector has a maximum efficiency to the charged particles and is relatively insensitive to the %-rays.     

Simulation study using GEANT4 Monte Carlo code for a Gd-coated resistive plate chamber as a thermal neutron detector

The Resistive Plate Chamber (RPC) has been developed in many application areas ever since its introduction, from high energy physics experiments to positron emission tomography. Such detectors can be coated with a Gd layer that enables them to detect thermal neutrons. Consequently these RPCs can be utilized for industrial and medical purposes. Here, we present the configuration of a resistive plate chamber which is utilized to detect thermal neutrons by employing GEANT4 Monte Carlo code. The response of the RPC was evaluated as a function of neutron energy in the GEANT4 Monte Carlo code. The simulation results are taken for incident neutron energy in the energy range from 25 meV to 100 meV. The detection efficiency was found to be between 10% and 20%, depending on the detector configuration, for incident thermal neutrons of 25 meV energy.     

Neutron dosimetric measurements in shuttle and MIR

Detector packages consisting of thermoluminescence detectors (TLD), nuclear emulsions and plastic track detectors were exposed at identical positions inside MIR space station and on shuttle flights inside Spacelab and Spacehab during different phases of the solar cycle. The objectives of the investigations are to provide data on charge and energy spectra of heavy ions, and the contribution of events with low-energy deposit (protons, electrons, gamma, etc.) to the dose, as well as the contribution of secondaries, such as nuclear disintegration stars and neutrons. For neutron dosimetry 6LiF (TLD600) and 7LiF (TLD700) chips were used both of which have almost the same response to gamma rays but different response to neutrons. Neutrons in space are produced mainly in evaporation and knock-on processes with energies mainly of 1-10 MeV and up to several 100 MeV, respectively. The energy spectrum undergoes continuous changes toward greater depth in the attenuating material until an equilibrium is reached. In equilibrium, the spectrum is a wide continuum extending down to thermal energies to which the 6LiF is sensitive. Based on the difference of absorbed doses in the 6LiF and 7LiF chips, thermal neutron fluxes from 1 to 2.3 cm-2 s-1 are calculated using the assumption that the maximum induced dose in TLD600 for 1 neutron cm-2 is 1.6 x 10(-10) Gy (Horowitz and Freeman, Nucl. Instr. and Meth. 157 (1978) 393). It is assumed that the flux of high-energy neutrons is at least of that quantity. Tissue doses were calculated taking as a mean ambient absorbed dose per neutron 6 x10(-12) Gy cm2 (for a10 MeV neutron). The neutron equivalent doses for the above-mentioned fluxes are 52 micro Gy d-1 and 120 micro Gy d-1. In recent experiments, a personal neutron dosimeter was integrated into the dosimeter packages. First results of this dosimeter which is based on nuclear track detectors with converter foils are reported. For future measurements, a scintillator counter with anticoincidence logic is under development.     

Method for Monitoring of Neutron Fields near High-Energy Accelerators

The monitoring of neutron radiation from high-energy accelerators cannot fully rely on the standard dosimeters and radiometers manufactured in Russia, since these are sensitive only to neutrons with energies below some 10 MeV. This is because neutrons of higher energies can significantly contribute to the personnel doses both close to the accelerator shield and in the neutron multiscattered field around the shield. In this paper, we propose to measure the ambient neutron dose in energy range 10^–2 MeV to 1 GeV with a device consisting of two polyethylene balls with diameters of 3 and 10 in. housing slow-neutron detectors. The larger ball also comprises a lead converter (10'' + Pb). This device can be implemented in zonal radiation monitoring in the near-accelerator area.     

Neutron scattering from Bismuth,Strontium and Sodium at 14.8 MeV

Differential cross sections for the elastic and inelastic scattering of 14.8 MeV neutrons from Bi, Sr and Na have been measured with four neutron detectors participating in a time-of-flight system which had an energy resolution of 900 keV for the incident neutrons. The whole procedure was checked by scattering from carbon. For the elastic scattering of Bi, Sr and Na satisfactory optical-model fits were obtained. The absolute cross sections for excitation of collective states are compared to the distorted-wave theory for direct reactions, and a reasonable agreement has been found. The nuclear deformation parameters are extracted and compared to the values from other experiments.     

Recent enhancements to the understanding of the response of the NRPB neutron personal dosemeter

The response of the NRPB neutron personal dosemeter has been determined for intermediate energy neutrons using the code MCNP. These calculations fill the energy gap for which it is difficult to obtain calibration fields, and thereby enable the response function to cover the energy range from thermal to 15 MeV in an effectively continuous fashion. Recent free-in-air irradiations of the dosemeter around its fast neutron threshold have enabled the energy dependence of response in the 100 keV–1.2 MeV energy range to be determined with high resolution. The effect of performing these irradiations free-in-air has been investigated using monoenergetic irradiations and MCNP calculations.     

介质型脉冲高能中子探测器

首次报道一种基于反冲质子法和高速带电粒子在物质表面引起二次电子发射的高能中子探测方法,研制成功的新一代强流脉冲高能中子探测器,即介质型脉冲高能中子探测器.与传统的强流脉冲中子探测器相比,该探测器对中子的探测及信号的传输过程是在介质中完成的,因而不需要真空和高压也可以正常工作,并具有多项优异的物理性能.该探测器的输出信号来自于高能中子在聚乙烯转换靶产生的高能反冲质子和这些质子穿越电荷收集极表面时产生的部分二次电子.选择适当的卡阈吸收片和电荷收集极材料,二次电子在探测灵敏度中的份额及其随能量的变化可由实验测定.     

Slow and fast neutron detection using bare CN-85 cellulose nitrate detectors

We report on the application of CN-85 cellulose nitrate track detectors to neutron dosimetry. CN-85 has the ability to register proton tracks and therefore can be used for neutron detection via neutron-induced recoil events within the detector matrix. The CN-85 detectors were used to study the distribution of the weakly moderated spallation neutrons, produced in the irradiation of an extended Pb-target with 1 GeV protons. The results obtained are compared with Monte Carlo simulations using the MCNPX-2.5.c code. The thermal and epithermal neutrons can also be detected by CN-85 detectors without the need for a neutron converter via the ¹⁴N(n, p)¹⁴C reaction, which has a positive Q-value of 626 keV.     

厚闪烁体内次级中子对快中子图像质量的影响研究

利用编制的快中子照相数值模拟程序(FNRSC)模拟计算了入射中子能量为14 MeV时,厚度5—300 mm闪烁体内次级中子对快中子图像质量的影响,结果表明闪烁体厚度d<50mm时,次级中子对图像的影响强烈依赖于闪烁体厚度,而当d>50 mm时,次级中子对图像的影响趋于饱和.将文献中利用蒙特卡罗中子-光子输运程序(MCNP)计算的次级中子对图像影响和文中计算结果进行了对比,给出了二者存在差异的主要原因：次级中子分布对入射中子空间分布的强烈依赖性;能量沉积和荧光输出这两种计算方法对 关键词： 14 MeV中子 快中子照相 次级中子 Monte Carlo模拟     

Spallation reactions studied with 4π-detector arrays

Recently here has been a renewed interest in the study of spallation reactions in basic nuclear physics as well as in potential applications. Spallation reactions induced by light projectiles (protons, antiprotons, pions, etc.) in the GeV range allow the formation of hot nuclei which do not suffer the collective excitations (compression, rotation, deformation) unavoidable when using massive projectiles. Such nuclei provide an ideal testbench for probing their decay as a function of excitation energy. In these investigations, 4π-detector arrays for charged particles and neutrons play a major role in the event-by-event sorting according to the excitation energy of the nucleus. Spallation reactions induced on heavy nuclei allow the conversion of the incident GeV proton into several tens of evaporated neutrons. The neutron production in thick targets has been investigated in great detail thanks to the use of high efficiency neutron detector arrays. When scattered on samples of inert or biological materials, these neutrons can be used to study details of the material structure. They could also be utilized for the transmutation of long-lived nuclear wastes or for the feeding of sub-critical nuclear reactors. The role of different types of multi-detector arrays is highlighted in this paper. Several references are also given for different uses of high efficiency neutron detectors in other contexts.     

Measurement of the neutron spectrum of a Pu-C source with a liquid scintillator

The neutron response function for a BC501A liquid scintillator (LS) has been measured using a series of monoenergetic neutrons produced by the p-T reaction. The proton energies were chosen such as to produce neutrons in the energy range of 1 to 20 MeV. The principles of the technique of unfolding a neutron energy spectrum by using the measured neutron response function and the measured Pulse Height (PH) spectrum is briefly described. The PH spectrum of neutrons from the Pu-C source, which will be used for the calibration of the reactor antineutrino detectors for the Daya Bay neutrino experiment, was measured and analyzed to get the neutron energy spectrum. Simultaneously the neutron energy spectrum of an Am-Be source was measured and compared with other measurements as a check of the result for the Pu-C source. Finally, an error analysis and a discussion of the results are given.     

强光一号装置单氘丝中子发射实验研究

 介绍了利用飞行时间技术测量Z箍缩装置单氘丝中子发射。实验发现，Z箍缩中子发射过程中,伴随产生很强的硬X射线，利用铝丝阵产生硬X射线的实验，建立较好的硬X射线屏蔽,大大减弱了它的辐射强度。所采用的双闪烁探测器中子探测系统，进行了3个发次的单氘丝负载实验，确认产生了聚变氘氘中子。4172发次实验测量结果表明，中子产额1.5×10⁹，中子能量为(2.45±0.26) MeV。实验结合分幅照相数据分析认为，峰电流和负载未能很好匹配，等离子体存在不稳定现象。     

Investigation of fast neutron interaction with235U

The angular distribution of neutrons emitted by elastic, inelastic and fission processes on²³⁵U were measured at the incident neutron energies of 1.5, 1.9, 2.3, 4.0, 4.5, 5.0 and 5.5 MeV using nanosecond time-of-flight technique. The differential elastic scattering cross sections and their angular distributions at all the seven energies are presented. The total elastic scattering cross sections, angle and energy integrated cross sections for the inelastically scattered neutrons in energy bands of 200 keV, fission cross sections and the angular distributions of fission neutrons were extracted at 1.5, 1.9 and 2.3 MeV incident neutron energies. The energy distributions of the prompt fission neutrons and of the inelastically scattered neutrons are given at the incoming neutron energies of 1.5, 1.9 and 2.3 MeV; and the average fission neutron energies and the inelastic neutron evaporation temperatures were also evaluated at these energies.     

Study of bubble damage detectors for neutron detection

Bubble damage detectors have been prepared by using polyacrylamide as detector solid and freon as detector liquid. Tests show that the prepared detectors are sensitive to fast neutrons and have proportionality between bubble number and neutron fluence within a certain range of neutron fluence. Therefore, it can be used as a fast neutron detector and a dosimeter. Our experiments also show that the sensitivity of the detector increases with temperature.     

Franco-Russian comparison of mixed neutron and gamma radiation field dosimeters at the Silène reactor

This paper gives the results of dosimetry measurements carried out in the Silène reactor at Valduc (France) with neutron and photon dosimeters in mixed neutron and gamma radiation fields, in the frame of a Franco-Russian comparison of dosimeters. Neutron dosimetry was supplied by passive semiconductors, activation detectors and nuclear track detectors. For photon dosimetry, thermoluminescent and passive semiconductor detectors were used. The experiments were located at 3 m from the reactor core, in free air and also at the front and back of a tissue-equivalent phantom. The pulse operating mode of the reactor was used to simulate a criticality accident with solid fissile material, while the free evolution mode simulated a criticality accident in a fissile solution. The photon absorbed dose showed a slight increase on entering the phantom compared to measurements in free air, probably due to backscattering by the phantom. At the rear of the phantom, the neutron kerma was four times lower than on the front, whereas the photon dose was only two times lower. The heterogeneity of dose inside the phantom was far greater for neutrons than for photons.     

Investigation of neutron emission in the interaction of relativistic protons and deuterons with lead targets

Measurements of neutron double-differential cross sections and yields in time-of-flight experiments with various lead targets and beams of protons and deuterons with an energy of about 2 GeV are discussed. Space and energy distribution of neutrons in an extended lead target is studied by threshold detectors in the proton energy range of 1–3.7 GeV. The average multiplicity of neutrons, neutron energy, and neutron multiplication in lead as functions of target dimension, type, and energy of the beam particle are analyzed.     

Modeling a neutron-rich nuclei source

The deuteron break-up process in a suitable converter gives rise to intense neutron beams. A source of neutron-rich nuclei based on the neutron-induced fission can be realised using these beams. A theoretical optimization of such a facility as a function of the incident deuteron energy is reported. The model used to determine the fission products takes into account the excitation energy of the target nucleus and the evaporation of prompt neutrons. Results are presented in connection with a converter-target specific geometry. Received: 1 December 2000 / Accepted: 12 May 2001