Thermoluminescence measurements of neutron streaming through JET Torus Hall ducts

Thermoluminescence detectors (TLD) were used for dose measurements at JET. Several hundreds of LiF detectors of various types, standard LiF:Mg,Ti and highly sensitive LiF:Mg,Cu,P were produced. LiF detectors consisting of natural lithium are sensitive to slow neutrons, their response to neutrons being enhanced by ⁶Li-enriched lithium or suppressed by using lithium consisting entirely of ⁷Li. Pairs of ⁶LiF/⁷LiF detectors allow distinguishing between neutron/non-neutron components of a radiation field. For detection of neutrons of higher energy, polyethylene (PE-300) moderators were used. TLDs, located in the centre of cylindrical moderators, were installed at eleven positions in the JET hall and the hall labyrinth in July 2012, and exposure took place during the last two weeks of the experimental campaign. Measurements of the gamma dose were obtained for all positions over a range of about five orders of magnitude variation. As the TLDs were also calibrated in a thermal neutron field, the neutron fluence at the experimental position could be derived. The experimental results are compared with calculations using the MCNP code. The results confirm that the TLD technology can be usefully applied to measurements of neutron streaming through JET Torus Hall ducts.     

Dosimeter characteristics of neutron track detectors used on high-energy accelerators

Conclusions Experience gathered in recording neutrons with the aid of polymer track detectors on calibrated neutron sources and around the 70 GeV proton synchrotron has shown that detectors making use of fissile nuclides in combination with polyethylene terephthalate track detectors and spark counting of the tracks are suitable and promising for their large scale use as measuring means of dosimetric characteristics of the neutron fields in a high-energy accelerator. They can be used to build an albedo-track dosimeter in which²³⁵U is the albedo detector and²³⁷Np,²³⁸U, and²³²Th are used as the threshold detector. Also a nitrate cellulose film without a converter can be used as a threshold detector but there the measurement error is 2–3 times greater than in the detectors making use of fissile nuclides because the film sensitivity has a considerable (30%) spread even to monochromatic neutrons. In addition, a detector based on²⁰⁹Bi can be used for spectrometric measurements of high-energy neutron spectra.The results of the present work will be of interest for an experimental verification of methods used to calculate the dose-dependent sensitivity of fissile-nuclide detectors and to develop promising types of personnel dosimeters.Translated from Atomnaya Énergiya, Vol. 69, No. 3, pp. 167–171, September, 1990.     

Comparison of measured and calculated neutron and gamma ray energy spectra behind an in-line shielded duct

Integral experiments that measure the transport of 14 MeV neutrons through a 0.30-m-diameter duct having a length-to-diameter ratio of 2.83 that is partially plugged with a 0.15 m diameter, 0.51 m long shield comprised of alternating layers of stainless steel type 304 and borated polyethylene have been carried out at the Oak Ridge National Laboratory. Measured and calculated neutron and gamma ray energy spectra are compared at several locations relative to the mouth of the duct. The measured spectra were obtained using an NE-213 liquid scintillator detector with pulse shape discrimination methods used to simultaneously resolve neutron and gamma ray events. The calculated spectra were obtained using a computer code network that incorporates two radiation transport methods: discrete ordinates (with P₃ multigroup cross sections) and Monte Carlo (with continuous point cross sections). The two radiation transport methods are required to account for neutrons that singly scatter from the duct to the detectors. The calculated and measured neutron energy spectra above 850 keV agree within 5–50% depending on detector location and neutron energy. The calculated and measured gamma ray energy spectra above 750 keV are also in favorable agreement, 5–50%, depending on detector location and gamma ray energy.     

Energy Response of Neutron Area Monitor with Silicon Semiconductor Detector

Abstract

A prototype neutron area monitor with a silicon semiconductor detector has been developed which has the energy response of 1 cm dose equivalent recommended by the 1CRP-26. Boron and proton radiators are coated on the surface of the silicon semiconductor detector. The detector is set at the center of a cylindrical polyethylene moderator. This moderator is covered by a porous cadmium board which serves as the thermal neutron absorber. Neutrons are detected as α-particles generated by the nuclear reaction ¹⁰B(n, α)⁷Li and as recoil protons generated by the interaction of fast neutrons with hydrogen.

The neutron energy response of the monitor was measured using thermal neutrons and monoenergetic fast neutrons generated by an accelerator. The response was consistent with the 1cm dose equivalent response required for the monitor within ±34% in the range of 0.025–15 MeV.     

Time Dependent DD Neutrons Measurement Using a Single Crystal Chemical Vapor Deposition Diamond Detector on EAST

A single crystal chemical vapor deposition(scCVD) diamond detector has been successfully employed for neutron measurements in the EAST(Experimental Advanced Superconducting Tokamak) plasmas.The scCVD diamond detector coated with a 5 μm ~6LiF(95%~6Li enriched) layer was placed inside a polyethylene moderator to enhance the detection efficiency.The time-dependent neutron emission from deuteron plasmas during neutral beam injection(NBI)heating was obtained.The measured results are compared with that of fission chamber detectors,which always act as standard neutron flux monitors.The scCVD diamond detector exhibits good reliability,stability and the capability to withstand harsh radiation environments despite its low detection efficiency due to the small active volume.     

Development of the high-energy neutron fluence rate standard field in Japan with a peak energy of 45 MeV using the 7Li(p,n)7Be reaction at TIARA

In this study, we developed a 45 MeV neutron fluence rate standard of Japan. Quasi-monoenergetic neutrons with a peak energy of 45 MeV in the neutron standard field were produced by the ⁷Li(p,n)⁷Be reaction using a 50-MeV proton beam from an azimuthally varying field (AVF) cyclotron of the Takasaki Ion Accelerators for Advanced Radiation Application (TIARA). The neutron energy spectrum was measured using an organic liquid scintillation detector and a ⁶Li-glass scintillation detector by the time-of-flight method, and using a Bonner sphere spectrometer by the unfolding method. The absolute neutron fluence was determined using a proton recoil telescope (PRT) composed of the liquid scintillation detector and a Si(Li) detector that was newly developed in the present study. The detection efficiency of the PRT was obtained using the MCNPX code. The peak neutron production cross section for the ⁷Li(p,n)⁷Be reaction was also derived from the neutron fluence in order to confirm the neutron fluence of the TIARA high-energy neutron field. The peak neutron production cross section obtained in the present study was in good agreement with those of previous studies. The characteristics of the 45-MeV neutron field in TIARA were successfully evaluated in order to calibrate high-energy neutron detectors and high-energy neutron dosimeters.     

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.     

Standardization of 2H(d,n)3He Neutron Source by Total Absorption Method Using Water Bath

The total absorption method using water bath is applied to the case of an almost monoenergetic anisotropic neutron stream from artificial accelerator, to be used for standardizing the strength of a ²H(d, n)³He source.

A detailed neutron balance equation is derived by taking into consideration certain correction terms to account for the finite water bath and associated equipment in the source. An experimental procedure for calibrating the thermal-flux-distribution and monitor neutron detectors is established to permit determination of the total loss of thermal neutrons in the bath and of the source strength at a given reference time. This calibration of the monitor detector makes it possible to measure the source strength as a function of time during accelerator operation simply by measuring only the counting rate registered by the detector.     

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.     

Studies on new neutron-sensitive dosimeters using an optically stimulated luminescence technique

The neutron response of detectors prepared using α-Al₂O₃:C phosphor developed using a melt processing technique and mixed with neutron converters was studied in monoenergetic neutron fields. The detector pellets were arranged in two different pairs: α-Al₂O₃:C + ⁶LiF/α-Al₂O₃:C + ⁷LiF and α-Al₂O₃:C + high-density polyethylene/α-Al₂O₃:C + Teflon, for neutron dosimetry using albedo and recoil proton techniques. The optically stimulated luminescence response of the Al₂O₃:C + ^6,7LiF dosimeter to radiation from a ²⁵²Cf source was 0.21, in terms of personal dose equivalent H_p(10) and relative to radiation from a ¹³⁷Cs source. This was comparable to results obtained with similar detectors prepared using commercially available α-Al₂O₃:C phosphor. The H_p(10) response of the α-Al₂O₃:C + ^6,7LiF dosimeters was found to decrease by more than two orders of magnitude with increasing neutron energy, as expected for albedo dosimeters. The response of the α-Al₂O₃:C + high-density polyethylene/α-Al₂O₃:C + Teflon dosimeters was small, of the order of 1% to 2% in terms of Hp(10) and relative to radiation from a ¹³⁷Cs source, for neutron energies greater than 1 MeV.     

Measurement of α particle energy loss in biological tissue below 2 MeV

The energy loss of α particles crossing biological tissue at energies between 0.8 and 2.2 MeV has been measured. This energy range is very important for boron neutron capture therapy, based on the ¹⁰B(n,α)⁷Li reaction, which emits α particles with energies of 1.78 and 1.47 MeV. One of the methods used for the measurement of the boron concentration in tissue is based on the deconvolution of the α spectra obtained from neutron irradiation of thin (70 μm) tissue samples. For this technique, a knowledge of the behaviour of the energy loss of the particles in the irradiated tissue is of critical importance. In particular, the curve of the residual energy as a function of the distance travelled in the tissue must be known. In this paper, the results of an experiment carried out with an ²⁴¹Am source and a series of cryostatic sections of rat-lung tissue are presented. The experimental measurements are compared with the results of Monte Carlo calculations performed with the MCNPX code.     

Application of a 6LiF Small Neutron Detector with an Optical Fiber to Tritium Production Rate Measurement in D-T Neutron Fields

⁶LiF small neutron detectors with an optical fiber have been used to measure ⁶Li(n,α)T reaction rate distributions at thermal research reactors and accelerator facilities. In the present study, we developed an experimental method for the measurement of tritium production rate (TPR) of ⁶Li using this small detector in deuterium-tritium (D-T) neutron fields. Reaction rate measurements with the detector were conducted in the D-T neutron fields at the Fusion Neutronics Source (FNS) facility. From the results, we determined that this detector can be used to measure the TPR distribution in soft neutron spectrum fields such as in a Be assembly. It is difficult to obtain ⁶Li(n,α)T reaction rate separately in hard neutron spectrum fields such as in a Li₂O assembly, because many kinds of charged particle production reactions need to be taken into consideration. However, a time-dependent reaction rate measurement method combined with the ⁶LiF detector and the ZnS detector is effective to separate the ⁶Li(n,α)T reaction from other reactions even in a hard spectrum field, and it can be applied to the measurement of the TPR distribution accurately.     

Feasibility study of an intense D–T fusion source: “The New Sorgentina”

The international community agrees on the importance to build a large facility devoted to test and validate materials to be used in harsh neutron environments. Such a facility, proposed by ENEA, reconsiders a previous study known as “Sorgentina” but takes into account new technological development so far attained. The “New Sorgentina” Fusion Source (NSFS) project is based upon an intense D–T 14 MeV neutron source achievable with T and D ion beams impinging on 2 m radius rotating targets. NSFS produces about 1 × 10¹³ n cm⁻² s⁻¹ over about 50 cm³. Larger volumes of lower neutron flux will be available (e.g. for TBM experiments) as well as collimated channels to study some features of the ITER neutron camera. The NSFS facility will overcome problems related to the ion source and accelerating system, by means of an upgraded version of the JET–PINI ion beams. NSFS has to be intended as an European facility that may be realized in a few years, once provided a preliminary technological program devoted to the operation of the ion source in continuous mode, target heat loading/removal, target and tritium handling, inventory as well as site licensing. In this contribution, the main characteristics of NSFS project will be presented.     

High Efficiency Neutron Detector for Low Neutron Flux Measurement

A neutron flux detector composed of a bank of nine He³ detectors is developed. It is employed successfully to record neutron flux from a pulsed source as low as 5 × 10² neutrons per m², which is about 0.5 nrem per shot. It may find applications in recording low neutron flux at large distances from the source or in detecting leakage from the radiation shield, and hence may help to estimate health hazards.     

非准直中子场能谱测量用多测量面~3He探测器理论设计

利用Geant4软件模拟计算了5种具有较好对称性的多测量面结构探测器对不同方向入射中子的探测效率。计算结果表明,具有对称性的多测量面结构探测器对不同方向入射的中子具有较好的探测效率响应方向一致性。其中球形正比探测器加6只圆形半导体探测器的结构芯体在各方面表现最均衡,具有应用于非准直中子场能谱测量的潜力。     

基于4H-SiC肖特基二极管的中子探测器

针对当前极端环境下耐辐照半导体中子探测器的需求,采用耐高温、耐辐照的碳化硅宽禁带半导体材料,利用⁶Li(n, α)³H核反应原理,研制了基于4H-SiC肖特基二极管的中子探测器。利用磁控溅射完成中子转化层的制作,并用SEM表征⁶LiF膜厚。在10～600 V反向偏压下,漏电流维持在6.4 nA以下,表明探测器具备良好的半导体-金属肖特基整流接触。利用²⁴¹Am源研究探测器对5.486 MeV的α粒子的响应,测得分辨率为4.5%。同时,利用该探测器测量了由临界装置产生并经石蜡块慢化后的热中子,观测到热中子与⁶Li作用产生的α和T粒子信号的实验结果。     

14 MeV neutron yields from D-T operation of the MIT Cockcroft-Walton accelerator

As part of our fusion-product diagnostic development program, we have begun a series of experiments with 14 MeV neutrons generated in a Cockcroft-Walton accelerator. Two different detectors have been used to measure the neutron yield: a silicon SBD and a Cu foil. The energy of the emitted neutrons has been determined by using two spectrometers: the SBD and a³He proportional counter. The reaction rate is monitored, with about ±5% accuracy, by detecting the particles from D + T n +. The neutron yields obtained from the Si detector and the Cu activation had associated uncertainties of about ±15% and agreed well with the predicted values from measurements.     

Cold fusion experiment at Department of Nuclear Engineering,National Tsing-Hua University

We have repeated the so-called cold fusion experiment by electrolyzing heavy water, with 0.1M LiOD, with palladium rod as the cathode and platinum wire as the anode. The purpose of our experiment is to dectect the neutrons that are produced from the fusion process of deuterium if fusion does occur. We use one³He detector and one BF₃ detector to detect the thermal neutrons coming out of the water bath that surrounds the heavy water cell. Possible neutron bursts are detected by the³He detector during a period of about 7 h after electrolyzing for 11 days.     

Design of an epi-thermal neutron flux intensity monitor with GaN wafer for boron neutron capture therapy

Boron neutron capture therapy (BNCT) is a promising cancer therapy. Epi-thermal neutron (0.5 eV < E_n < 10 keV) flux intensity is one of the basic characteristics for modern BNCT. In this work, based on the ⁷¹Ga(n,γ)⁷²Ga reaction, a new simple monitor with gallium nitride (GaN) wafer as activation material was designed by Monte Carlo simulations to precisely measure the absolute integral flux intensity of epi-thermal neutrons especially for practical BNCT. In the monitor, a GaN wafer was positioned in the center of a polyethylene sphere as neutron moderator covered with cadmium (Cd) layer as thermal neutron absorber outside. The simulation results and related analysis indicated that the epi-thermal neutron flux intensity could be precisely measured by the presently designed monitor.