Development of neutron imaging beamline for NDT applications at Dhruva reactor, India

Science.gov (United States)

Shukla, Mayank; Roy, Tushar; Kashyap, Yogesh; Shukla, Shefali; Singh, Prashant; Ravi, Baribaddala; Patel, Tarun; Gadkari, S. C.

2018-05-01

Thermal neutron imaging techniques such as radiography or tomography are very useful tool for various scientific investigations and industrial applications. Neutron radiography is complementary to X-ray radiography, as neutrons interact with nucleus as compared to X-ray interaction with orbital electrons. We present here design and development of a neutron imaging beamline at 100 MW Dhruva research reactor for neutron imaging applications such as radiography, tomography and phase contrast imaging. Combinations of sapphire and bismuth single crystals have been used as thermal neutron filter/gamma absorber at the input of a specially designed collimator to maximize thermal neutron to gamma ratio. The maximum beam size of neutrons has been restricted to ∼120 mm diameter at the sample position. A cadmium ratio of ∼250 with L / D ratio of 160 and thermal neutron flux of ∼ 4 × 107 n/cm2 s at the sample position has been measured. In this paper, different aspects of the beamline design such as collimator, shielding, sample manipulator, digital imaging system are described. Nondestructive radiography/tomography experiments on hydrogen concentration in Zr-alloy, aluminium foam, ceramic metal seals etc. are also presented.

Gamma–neutron imaging system utilizing pulse shape discrimination with CLYC

International Nuclear Information System (INIS)

Whitney, Chad M.; Soundara-Pandian, Lakshmi; Johnson, Erik B.; Vogel, Sam; Vinci, Bob; Squillante, Michael; Glodo, Jarek; Christian, James F.

2015-01-01

Recently, RMD has investigated the use of CLYC (Cs 2 LiYCl 6 :Ce), a new and emerging scintillation material, in a gamma–neutron coded aperture imaging system based on RMD's commercial RadCam TM instrument. CLYC offers efficient thermal neutron detection, fast neutron detection capabilities, excellent pulse shape discrimination (PSD), and gamma-ray energy resolution as good as 4% at 662 keV. PSD improves the isolation of higher energy gammas from thermal neutron interactions (>3 MeV electron equivalent peak), compared to conventional pulse height techniques. The scintillation emission time in CLYC provides the basis for PSD; where neutron interactions result in a slower emission rise and decay components while gamma interactions result in a faster emission components. By creating a population plot based on the ratio of the decay tail compared to the total integral amplitude (PSD ratio), discrimination of gammas, thermal neutrons, and fast neutrons is possible. Previously, we characterized the CLYC-based RadCam system for imaging gammas and neutrons using a layered W-Cd coded aperture mask and employing only pulse height discrimination. In this paper, we present the latest results which investigate gamma-neutron imaging capabilities using PSD. An FPGA system is used to acquire the CLYC–PSPMT last dynode signals, determine a PSD ratio for each event, and compare it to a calibrated PSD cutoff. Each event is assigned either a gamma (low) or neutron (high) flag signal which is then correlated with the imaging information for each event. - Highlights: • The latest results are presented for our CLYC RadCam-2 system which investigate gamma–neutron imaging using pulse shape discrimination. • CLYC RadCam-2 system successfully discriminates gammas, thermal neutrons, and fast neutrons by employing a fully integrated, FPGA-based PSD system. • Imaging of our 252 Cf source was possible using both pulse height and pulse shape discrimination with CLYC. • Imaging

A study of television imaging system for fast neutron radiography

International Nuclear Information System (INIS)

Yoshii, Koji

1992-01-01

The neutron radiography with fast neutron beam is a very useful imaging technique for thicker objects, especially those composed of hydrogen-rich materials which are sometimes difficult to image by thermal neutron radiography. The fast neutron radiography has not been studied so much as the thermal neutron radiography. The fast neutron radiography has been studied at the fast neutron source reactor 'Yayoi' of the University of Tokyo built in Tokai-mura. The average neutron energy of the Yayoi is about 1 MeV, and the peak neutron flux at the core center is 0.8 x 10 12 at the maximum operating power of 2 kW. In the experiment on fast neutron radiography, a CR39 nuclear track detector has been used successfully. But in the Yayoi radiography procedure, about 24 hours were required for obtaining an imaging result. To get a prompt imaging result and a real-time imaging result, it is necessary to develop a fast neutron television system, and in this paper, a new fast neutron TV system is proposed. The main difference is the converter material sensitive to fast neutrons. The study on the fast neutron TV system was carried out by using the Baby Cyclotron of Japan Steel Works, and the good images were realized. (K.I.)

Neutron beam imaging with GEM detectors

International Nuclear Information System (INIS)

Albani, G.; Cazzaniga, C.; Rebai, M.; Gorini, G.; Croci, G.; Muraro, A.; Cippo, E. Perelli; Tardocchi, M.; Cavenago, M.; Murtas, F.; Claps, G.; Pasqualotto, R.

2015-01-01

Neutron GEM-based detectors represent a new frontier of devices in neutron physics applications where a very high neutron flux must be measured such as future fusion experiments (e.g. ITER Neutral beam Injector) and spallation sources (e.g. the European Spallation source). This kind of detectors can be properly adapted to be used both as beam monitors but also as neutron diffraction detectors that could represent a valid alternative for the 3 He detectors replacement. Fast neutron GEM detectors (nGEM) feature a cathode composed by one layer of polyethylene and one of aluminium (neutron scattering on hydrogen generates protons that are detected in the gas) while thermal neutron GEM detectors (bGEM) are equipped with a borated aluminium cathode (charged particles are generated through the 10 B(n,α) 7 Li reaction). GEM detectors can be realized in large area (1 m 2 ) and their readout can be pixelated. Three different prototypes of nGEM and one prototype of bGEM detectors of different areas and equipped with different types of readout have been built and tested. All the detectors have been used to measure the fast and thermal neutron 2D beam image at the ISIS-VESUVIO beamline. The different kinds of readout patterns (different areas of the pixels) have been compared in similar conditions. All the detectors measured a width of the beam profile consitent with the expected one. The imaging property of each detector was then tested by inserting samples of different material and shape in the beam. All the samples were correctly reconstructed and the definition of the reconstruction depends on the type of readout anode. The fast neutron beam profile reconstruction was then compared to the one obtained by diamond detectors positioned on the same beamline while the thermal neutron one was compared to the imaged obtained by cadmium-coupled x-rays films. Also efficiency and the gamma background rejection have been determined. These prototypes represent the first step towards

Non-destructive studies of fuel pellets by neutron resonance absorption radiography and thermal neutron radiography

Energy Technology Data Exchange (ETDEWEB)

Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California, Berkeley, CA 94720 (United States); Vogel, S.C.; Mocko, M.; Bourke, M.A.M.; Yuan, V.; Nelson, R.O.; Brown, D.W. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

2013-09-15

Many isotopes in nuclear materials exhibit strong peaks in neutron absorption cross sections in the epithermal energy range (1–1000 eV). These peaks (often referred to as resonances) occur at energies specific to particular isotopes, providing a means of isotope identification and concentration measurements. The high penetration of epithermal neutrons through most materials is very useful for studies where samples consist of heavy-Z elements opaque to X-rays and sometimes to thermal neutrons as well. The characterization of nuclear fuel elements in their cladding can benefit from the development of high resolution neutron resonance absorption imaging (NRAI), enabled by recently developed spatially-resolved neutron time-of-flight detectors. In this technique the neutron transmission of the sample is measured as a function of spatial location and of neutron energy. In the region of the spectra that borders the resonance energy for a particular isotope, the reduction in transmission can be used to acquire an image revealing the 2-dimensional distribution of that isotope within the sample. Provided that the energy of each transmitted neutron is measured by the neutron detector used and the irradiated sample possesses neutron absorption resonances, then isotope-specific location maps can be acquired simultaneously for several isotopes. This can be done even in the case where samples are opaque or have very similar transmission for thermal neutrons and X-rays or where only low concentrations of particular isotopes are present (<0.1 atom% in some cases). Ultimately, such radiographs of isotope location can be utilized to measure isotope concentration, and can even be combined to produce three-dimensional distributions using tomographic methods. In this paper we present the proof-of-principle of NRAI and transmission Bragg edge imaging performed at Flight Path 5 (FP5) at the LANSCE pulsed, moderated neutron source of Los Alamos National Laboratory. A set of urania mockup

Applications of image plates in neutron radiography and neutron diffraction at BARC, Trombay

International Nuclear Information System (INIS)

Shaikh, A.M.

2013-01-01

Neutron radiography techniques based on Gd, Dy and In metallic foils and X-ray film have been used at this centre since early seventies for various NDT and R and D work in nuclear, defence and aerospace industries. In recent years use of photostimulated luminescence based phosphor imaging plate has been introduced in our work. This has enabled to achieve higher sensitivities and dynamic ranges of recording radiographs with acceptable spatial resolution. It also provides digital image information which is more convenient for quantitative evaluations. Neutron image plates have been used in variety of radiography techniques such as conventional neutron radiography (NR), neutron induced beta radiography (NIBR), hydrogen sensitive epithermal neutron radiography (HYSEN) and for neutron powder diffractometry using Apsara, CIRUS and Dhruva reactors as neutron sources. Recently the image plates have also been used for characterization of thermalized neutron beam from a plasma focus neutron source and recording neutron radiographs. Prior to the utilization image plates have been characterised for their performance. Details of the measurements and applications will be presented. (author)

Performance test of Si PIN photodiode line scanner for thermal neutron detection

Energy Technology Data Exchange (ETDEWEB)

Totsuka, Daisuke, E-mail: totsuka@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Nihon Kessho Kogaku Co., Ltd., 810-5 Nobe-cho Tatebayashi, Gunma 374-0047 (Japan); Yanagida, Takayuki [New Industry Creation Hatchery Center (NICHe) 6-6-10 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi 980-8579 (Japan); Fukuda, Kentaro; Kawaguchi, Noriaki [Tokuyama Corp., 3 Shibuya Shibuya-ku, Tokyo 150-8383 (Japan); Fujimoto, Yutaka [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Pejchal, Jan [Institute of Physics AS CR, Cukrovarnicka 10, Prague 6, 162-53 (Czech Republic); Yokota, Yuui [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Yoshikawa, Akira [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); New Industry Creation Hatchery Center (NICHe) 6-6-10 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi 980-8579 (Japan)

2011-12-11

Thermal neutron imaging using Si PIN photodiode line scanner and Eu-doped LiCaAlF{sub 6} crystal scintillator has been developed. The pixel dimensions of photodiode are 1.18 mm (width) Multiplication-Sign 3.8 mm (length) with 0.4 mm gap and the module has 192 channels in linear array. The emission peaks of Eu-doped LiCaAlF{sub 6} after thermal neutron excitation are placed at 370 and 590 nm, and the corresponding photon sensitivities of photodiode are 0.04 and 0.34 A/W, respectively. Polished scintillator blocks with a size of 1.18 mm (width) Multiplication-Sign 3.8 mm (length) Multiplication-Sign 5.0 mm (thickness) were wrapped by several layers of Teflon tapes as a reflector and optically coupled to the photodiodes by silicone grease. JRR-3 MUSASI beam line emitting 13.5 meV thermal neutrons with the flux of 8 Multiplication-Sign 10{sup 5} n/cm{sup 2} s was used for the imaging test. As a subject for imaging, a Cd plate was moved at the speed of 50 mm/s perpendicular to the thermal neutron beam. Analog integration time was set to be 416.6 {mu}s, then signals were converted by a delta-sigma A/D converter. After the image processing, we successfully obtained moving Cd plate image under thermal neutron irradiation using PIN photodiode line scanner coupled with Eu-doped LiCaAlF{sub 6} scintillator.

Performance test of Si PIN photodiode line scanner for thermal neutron detection

International Nuclear Information System (INIS)

Totsuka, Daisuke; Yanagida, Takayuki; Fukuda, Kentaro; Kawaguchi, Noriaki; Fujimoto, Yutaka; Pejchal, Jan; Yokota, Yuui; Yoshikawa, Akira

2011-01-01

Thermal neutron imaging using Si PIN photodiode line scanner and Eu-doped LiCaAlF 6 crystal scintillator has been developed. The pixel dimensions of photodiode are 1.18 mm (width)×3.8 mm (length) with 0.4 mm gap and the module has 192 channels in linear array. The emission peaks of Eu-doped LiCaAlF 6 after thermal neutron excitation are placed at 370 and 590 nm, and the corresponding photon sensitivities of photodiode are 0.04 and 0.34 A/W, respectively. Polished scintillator blocks with a size of 1.18 mm (width)×3.8 mm (length)×5.0 mm (thickness) were wrapped by several layers of Teflon tapes as a reflector and optically coupled to the photodiodes by silicone grease. JRR-3 MUSASI beam line emitting 13.5 meV thermal neutrons with the flux of 8×10 5 n/cm 2 s was used for the imaging test. As a subject for imaging, a Cd plate was moved at the speed of 50 mm/s perpendicular to the thermal neutron beam. Analog integration time was set to be 416.6 μs, then signals were converted by a delta-sigma A/D converter. After the image processing, we successfully obtained moving Cd plate image under thermal neutron irradiation using PIN photodiode line scanner coupled with Eu-doped LiCaAlF 6 scintillator.

Monte Carlo simulation of a coded-aperture thermal neutron camera

International Nuclear Information System (INIS)

Dioszegi, I.; Salwen, C.; Forman, L.

2011-01-01

We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)

Application of imaging plate neutron detector to neutron radiography

CERN Document Server

Fujine, S; Kamata, M; Etoh, M

1999-01-01

As an imaging plate neutron detector (IP-ND) has been available for thermal neutron radiography (TNR) which has high resolution, high sensitivity and wide range, some basic characteristics of the IP-ND system were measured at the E-2 facility of the KUR. After basic performances of the IP were studied, images with high quality were obtained at a neutron fluence of 2 to 7x10 sup 8 n cm sup - sup 2. It was found that the IP-ND system with Gd sub 2 O sub 3 as a neutron converter material has a higher sensitivity to gamma-ray than that of a conventional film method. As a successful example, clear radiographs of the flat view for the fuel side plates with boron burnable poison were obtained. An application of the IP-ND system to neutron radiography (NR) is presented in this paper.

Design of hyper-thermal neutron irradiation fields for neutron capture therapy in KUR-heavy water neutron irradiation facility. Mounting of hyper-thermal neutron converter in therapeutic collimator

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2001-01-01

Neutron capture therapy (NCP) using thermal neutron needs to improve of depth dose distribution in a living body. Epi-thermal neutron following moderation of fast neutron is usually used for improving of the depth dose distribution. The moderation method of fast neutron, however, gets mixed some of high energy neutron which give some of serious effects to a living body, and involves the difficulty for collimation of thermal neutron to the diseased part. Hyper-thermal neutrons, which are in an energy range of 0.1-3 eV at high temperature side of thermal neutron, are under consideration for application to the NCP. The hyper-thermal neutrons can be produced by up-scattering of thermal neutron in a high temperature material. Fast neutron components in collimator for the NCP reduce on application of the up-scattering method. Graphite at high temperature (>1000k) is used as a hyper-thermal neutron converter. The hyper-thermal neutron converter is planted to mount on therapeutic collimator which is located at the nearest side of patient for the NCP. Total neutron flux, ratio of hyper-thermal neutron to total neutron, and ratio of gamma-ray dose to neutron flux are calculated as a function of thickness of the graphite converter using monte carlo code MCNP-V4B. (M. Suetake)

Hyper-thermal neutron irradiation field for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1994-01-01

The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

International Nuclear Information System (INIS)

Thiagu Supramaniam

2007-01-01

The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

Moisture imaging of a camphor tree by neutron beam

International Nuclear Information System (INIS)

Nakanishi, Tomoko M.; Karakama, Isamu; Sakura, Tsuguo; Matsubayashi, Masashi

1998-01-01

Moisture distribution of a camphor tree was presented. A 23 year old camphor tree was downed at university forest and a wood disk, about 1 cm in width, was lumbered out from the breast height of the tree. The wood disk as well as a newly developing branch of the tree were irradiated with thermal neutrons at an atomic reactor installed at Japan Atomic Energy Research Institute. The total flux of thermal neutron was 3.0 x 10 9 n/cm 2 . Water specific images of the disk and a branch were presented with high resolution, which was estimated to be about 16 μm. In the case of wood disk, moisture decreasing manner while drying was also shown through neutron image. Neutron images showed that the moisture decreasing rate in sapwood was similar to that of heartwood. (author)

Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

Science.gov (United States)

Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

2018-01-01

Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

Computed tomography with thermal neutrons and gaseous position sensitive detector

International Nuclear Information System (INIS)

Souza, Maria Ines Silvani

2001-12-01

A third generation tomographic system using a parallel thermal neutron beam and gaseous position sensitive detector has been developed along three discrete phases. At the first one, X-ray tomographic images of several objects, using a position sensitive detector designed and constructed for this purpose have been obtained. The second phase involved the conversion of that detector for thermal neutron detection, by using materials capable to convert neutrons into detectable charged particles, testing afterwards its performance in a tomographic system by evaluation the quality of the image arising from several test-objects containing materials applicable in the engineering field. High enriched 3 He, replacing the argon-methane otherwise used as filling gas for the X-ray detection, as well as, a gadolinium foil, have been utilized as converters. Besides the pure enriched 3 He, its mixture with argon-methane and later on with propane, have been also tested, in order to evaluate the detector efficiency and resolution. After each gas change, the overall performance of the tomographic system using the modified detector, has been analyzed through measurements of the related parameters. This was done by analyzing the images produced by test-objects containing several materials having well known attenuation coefficients for both thermal neutrons and X-rays. In order to compare the performance of the position sensitive detector as modified to detect thermal neutrons, with that of a conventional BF 3 detector, additional tomographs have been conducted using the last one. The results have been compared in terms of advantages, handicaps and complementary aspects for different kinds of radiation and materials. (author)

Non-destructive assay of mechanical components using gamma-rays and thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Souza, Erica Silvani; Avelino, Mila R. [PPG-EM/UERJ, R. Sao Francisco Xavier, 524, Maracana - Rio de Janeiro - RJ (Brazil); Almeida, Gevaldo L. de; Souza, Maria Ines S. [IEN/CNEN, Rua Helio de Almeida, 75, Ilha do Fundao, Rio de Janeiro - RJ (Brazil)

2013-05-06

This work presents the results obtained in the inspection of several mechanical components through neutron and gamma-ray transmission radiography. The 4.46 Multiplication-Sign 10{sup 5} n.cm{sup -2}.s{sup -1} thermal neutron flux available at the main port of the Argonauta research reactor in Instituto de Engenharia Nuclear has been used as source for the neutron radiographic imaging. The 412 keV {gamma}-ray emitted by {sup 198}Au, also produced in that reactor, has been used as interrogation agent for the gamma radiography. Imaging Plates - IP specifically designed to operate with thermal neutrons or with X-rays have been employed as detectors and storage devices for each of these radiations.

Applying thermal neutron radiography to non-destructive assays of dynamic systems

International Nuclear Information System (INIS)

Silvani, Maria I.; Almeida, Gevaldo L. de; Goncalves, Marcelo J.; Lopes, Ricardo T.

2008-01-01

Dynamic processes or systems frequently can not have their behavior directly analyzed due to safety reasons or because they require destructive assays, which can not be always afforded when high-cost equipment, devices and components are involved. Under these circumstances, some kind of non-destructive technique should be applied to preserve the safety of the personnel performing the assay, as well as the integrity of the piece being inspected. Thermal neutrons are specially suited as a tool for this purpose, thanks to their capability to pass through metallic materials, which could be utterly opaque to X-rays. This paper describes the accomplishments achieved at the Instituto de Engenharia Nuclear / CNEN, Brazil, aiming at the development of an Image Acquisition System capable to perform non-destructive assays using thermal neutrons. It is comprised of a thermal neutron source provided by the Argonauta research reactor, a converter-scintillating screen, and a CCD-based video camera optically coupled to the screen through a dark chamber equipped with a mirror. The developed system has been used to acquire 2D neutron radiographic images of static devices to reveal their inner structure, as well as movies of running systems and working devices to verify its functioning and soundness. Radiographic images of objects taken at different angles would be later on used as projections to retrieve - through a proper unfolding software - their 3D images expressed as attenuation coefficients for thermal neutrons. A quantitative performance of the system has been assessed through its Modulation Transfer Function - MTF. In order to determine this curve, unique collimators designed to simulate different spatial frequencies have been manufactured. Besides that, images of some objects have been acquired with the system being developed as well as using the conventional radiographic film, allowing thus a qualitative comparison between them. (author)

Pulsed thermal neutron source at the fast neutron generator.

Science.gov (United States)

Tracz, Grzegorz; Drozdowicz, Krzysztof; Gabańska, Barbara; Krynicka, Ewa

2009-06-01

A small pulsed thermal neutron source has been designed based on results of the MCNP simulations of the thermalization of 14 MeV neutrons in a cluster-moderator which consists of small moderating cells decoupled by an absorber. Optimum dimensions of the single cell and of the whole cluster have been selected, considering the thermal neutron intensity and the short decay time of the thermal neutron flux. The source has been built and the test experiments have been performed. To ensure the response is not due to the choice of target for the experiments, calculations have been done to demonstrate the response is valid regardless of the thermalization properties of the target.

Fail-safe neutron shutter used for thermal neutron radiography

International Nuclear Information System (INIS)

Sachs, R.D.; Morris, R.A.

1976-11-01

A fail-safe, reliable, easy-to-use neutron shutter was designed, built, and put into operation at the Omega West Reactor, Los Alamos Scientific Laboratory. The neutron shutter will be used primarily to perform thermal neutron radiography, but is also available for a highly collimated source of thermal neutrons [neutron flux = 3.876 x 10 6 (neutrons)/(cm 2 .s)]. Neutron collimator sizes of either 10.16 by 10.16 cm or 10.16 by 30.48 cm are available

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

Energy Technology Data Exchange (ETDEWEB)

Soltes, Jaroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague, (Czech Republic); Viererbl, Ladislav; Lahodova, Zdena; Koleska, Michal; Vins, Miroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic)

2015-07-01

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values

A study on the utilization of hyper-thermal neutrons for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1993-01-01

The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwellian distribution of a higher temperature than the room temperature of 300 K, was studied in order to improve the thermal neutron flux distribution at the deeper part in a living body for neutron capture therapy. Simulation calculations were carried out using MCNP-V3 in order to confirm the characteristics of hyper-thermal neutrons, i.e., (1) depth dependence of neutron energy spectrum, and (2) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that the hyper-thermal neutron irradiation can improve the thermal neutron flux distribution in the deeper and wider area in a living body compared with the thermal neutron irradiation. Practically, by the incidence of the hyper-thermal neutrons with a 3000 K Maxwellian distribution, the thermal neutron flux at 5 cm depth can be given about four times larger than by the incidence of the thermal neutrons of 300 K. (author)

Development of high flux thermal neutron generator for neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Vainionpaa, Jaakko H., E-mail: hannes@adelphitech.com [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K. [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Jones, Glenn [G& J Jones Enterprice, 7486 Brighton Ct, Dublin, CA 94568 (United States); Pantell, Richard H. [Department of Electrical Engineering, Stanford University, Stanford, CA (United States)

2015-05-01

The new model DD110MB neutron generator from Adelphi Technology produces thermal (<0.5 eV) neutron flux that is normally achieved in a nuclear reactor or larger accelerator based systems. Thermal neutron fluxes of 3–5 · 10{sup 7} n/cm{sup 2}/s are measured. This flux is achieved using four ion beams arranged concentrically around a target chamber containing a compact moderator with a central sample cylinder. Fast neutron yield of ∼2 · 10{sup 10} n/s is created at the titanium surface of the target chamber. The thickness and material of the moderator is selected to maximize the thermal neutron flux at the center. The 2.5 MeV neutrons are quickly thermalized to energies below 0.5 eV and concentrated at the sample cylinder. The maximum flux of thermal neutrons at the target is achieved when approximately half of the neutrons at the sample area are thermalized. In this paper we present simulation results used to characterize performance of the neutron generator. The neutron flux can be used for neutron activation analysis (NAA) prompt gamma neutron activation analysis (PGNAA) for determining the concentrations of elements in many materials. Another envisioned use of the generator is production of radioactive isotopes. DD110MB is small enough for modest-sized laboratories and universities. Compared to nuclear reactors the DD110MB produces comparable thermal flux but provides reduced administrative and safety requirements and it can be run in pulsed mode, which is beneficial in many neutron activation techniques.

Neutron radiography using neutron imaging plate

International Nuclear Information System (INIS)

Chankow, Nares; Wonglee, Sarinrat

2008-01-01

Full text: The aims of this research are to study properties of neutron imaging plate, to obtain a suitable condition for neutron radiography and to use the neutron imaging plate for testing of materials nondestructively. The experiments were carried out by using a neutron beam from the Thai Research Reactor TRR-1/M1 at a power of 1.2 MW. A BAS-ND 2040 FUJI neutron imaging plate and a MX125 Kodak X-ray film/Gadolinium neutron converter screen combination were tested for comparison. It was found that the photostimulated light (PSL) read out of the imaging plate was directly proportional to the exposure time. It was also found that radiography with neutron using the imaging plate was approximately 40 times faster than the conventional neutron radiography using x-ray film/Gd converter screen combination. The sensitivity of the imaging plate to gamma-rays was investigated by using gamma-rays from an 192 Ir and a 60 Co radiographic sources. The imaging plate was found to be 5-6 times less sensitive to gamma-rays than a FUJI BAS-MS 2040 gamma-ray imaging plate. Finally, some specimens were selected to be radiographed with neutrons using the imaging plate and the x-ray film/Gd converter screen combination in comparison to x-rays. Parts containing light elements could be clearly observed by the two neutron radiographic techniques. It could be concluded that the image quality from the neutron imaging plate was comparable to the conventional x-ray film/Gd converter screen combination but the exposure time could be approximately reduced by a factor of 40

Probing the potential of neutron imaging for biomedical and biological applications

International Nuclear Information System (INIS)

Watkin, Kenneth L.; Bilheux, Hassina Z.; Ankner, John Francis

2009-01-01

Neutron imaging of biological specimens began soon after the discovery of the neutron by Chadwick in 1932. The first samples included tumors in tissues, internal organs in rats, and bones. These studies mainly employed thermal neutrons and were often compared with X-ray images of the same or equivalent samples. Although neutron scattering is widely used in biological studies, neutron imaging has yet to be exploited to its full capability in this area. This chapter summarizes past and current research efforts to apply neutron radiography to the study of biological specimens, in the expectation that clinical and medical research, as well as forensic science, may benefit from it.

Pilot experimental study on continual spectrum thermal neutron in-line phase contrast radiography

International Nuclear Information System (INIS)

Tang Bin; Huo Heyong; Wu Yang

2009-01-01

The in-line phase contrast radiography is one of phase contrast imaging methods. The neutron in-line phase contrast is developed with X-rays phase contrast radiography. In the paper, the principle of in-line phase contrast is introduced briefly and the experimental result of thermal neutron in-line contrast at SPRR-300 is analysed. It shows that thermal neutron can be used as in-line phase contrast radiography and enhances the edge of some sample in radiography and complements the disadvantage of conventional neutron radiography. (authors)

Characterization of a neutron imaging setup at the INES facility

Science.gov (United States)

Durisi, E. A.; Visca, L.; Albertin, F.; Brancaccio, R.; Corsi, J.; Dughera, G.; Ferrarese, W.; Giovagnoli, A.; Grassi, N.; Grazzi, F.; Lo Giudice, A.; Mila, G.; Nervo, M.; Pastrone, N.; Prino, F.; Ramello, L.; Re, A.; Romero, A.; Sacchi, R.; Salvemini, F.; Scherillo, A.; Staiano, A.

2013-10-01

The Italian Neutron Experimental Station (INES) located at the ISIS pulsed neutron source (Didcot, United Kingdom) provides a thermal neutron beam mainly used for diffraction analysis. A neutron transmission imaging system was also developed for beam monitoring and for aligning the sample under investigation. Although the time-of-flight neutron diffraction is a consolidated technique, the neutron imaging setup is not yet completely characterized and optimized. In this paper the performance for neutron radiography and tomography at INES of two scintillator screens read out by two different commercial CCD cameras is compared in terms of linearity, signal-to-noise ratio, effective dynamic range and spatial resolution. In addition, the results of neutron radiographies and a tomography of metal alloy test structures are presented to better characterize the INES imaging capabilities of metal artifacts in the cultural heritage field.

Thermal neutron imaging with rare-earth-ion-doped LiCaAlF6 scintillators and a sealed 252Cf source

International Nuclear Information System (INIS)

Kawaguchi, Noriaki; Yanagida, Takayuki; Fujimoto, Yutaka; Yokota, Yuui; Kamada, Kei; Fukuda, Kentaro; Suyama, Toshihisa; Watanabe, Kenichi; Yamazaki, Atsushi; Chani, Valery; Yoshikawa, Akira

2011-01-01

Thermal neutron imaging with Ce-doped LiCaAlF 6 crystals has been performed. The prototype of the neutron imager using a Ce-doped LiCaAlF 6 scintillating crystal and a position sensitive photomultiplier tube (PSPMT) which had 64 multi-channel anode was developed. The Ce-doped LiCaAlF 6 single crystal was grown by the Czochralski method. A plate with dimensions of a diameter of 50x2 mm 2 was cut from the grown crystal, polished, and optically coupled to PSPMT by silicone grease. The 252 Cf source ( 6 .

Thermal neutron polarisation

International Nuclear Information System (INIS)

Satya Murthy, N.S.; Madhava Rao, L.

1984-01-01

The basic principle for the production of polarised thermal neutrons is discussed and the choice of various crystal monochromators surveyed. Brief mention of broad-spectrum polarisers is made. The application of polarised neutrons to the study of magnetisation density distributions in magnetic crystals, the dynamic concept of polarisation, principle and use of polarisation analysis, the neutron spin-echo technique are discussed. (author)

Study on the dose distribution of the mixed field with thermal and epi-thermal neutrons for neutron capture therapy

International Nuclear Information System (INIS)

Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

1994-01-01

Simulation calculations using DOT 3.5 were carried out in order to confirm the characteristics of depth-dependent dose distribution in water phantom dependent on incident neutron energy. The epithermal neutrons mixed to thermal neutron field is effective improving the thermal neutron depth-dose distribution for neutron capture therapy. A feasibility study on the neutron energy spectrum shifter was performed using ANISN-JR for the KUR Heavy Water Facility. The design of the neutron spectrum shifter is feasible, without reducing the performance as a thermal neutron irradiation field. (author)

Imaging with cold neutrons

International Nuclear Information System (INIS)

Lehmann, E.H.; Kaestner, A.; Josic, L.; Hartmann, S.; Mannes, D.

2011-01-01

Neutrons for imaging purposes are provided mainly from thermal beam lines at suitable facilities around the world. The access to cold neutrons is presently limited to very few places only. However, many challenging options for imaging with cold neutrons have been found out, given by the interaction behavior of the observed materials with neutrons in the cold energy range (3-10 A). For absorbing materials, the interaction probability increases proportionally with the wavelength with the consequence of more contrast but less transmission with cold neutrons. Many materials are predominantly scattering neutrons, in particular most of crystalline structural materials. In these cases, cold neutrons play an important role by covering the energy range of the most important Bragg edges given by the lattice planes of the crystallites. This particular behavior can be used for at least two important aspects-choosing the right energy of the initial beam enables to have a material more or less transparent, and a direct macroscopic visualization of the crystalline structure and its change in a manufacturing process. Since 2006, PSI operates its second beam line for neutron imaging, where cold neutrons are provided from a liquid deuterium cold source (operated at 25 K). It has been designed to cover the most current aspects in neutron imaging research with the help of high flexibility. This has been done with changeable inlet apertures, a turbine based velocity selector, two beam positions and variable detector systems, satisfying the demands of the individual investigation. The most important detection system was found to be a micro-tomography system that enables studies in the presently best spatial resolution. In this case, the high contrast from the sample interaction process and the high detection probability for the cold neutrons combines in an ideal combination for the best possible performance. Recently, it was found out that the energy selective studies might become a

Preliminary energy-filtering neutron imaging with time-of-flight method on PKUNIFTY: A compact accelerator based neutron imaging facility at Peking University

Energy Technology Data Exchange (ETDEWEB)

Wang, Hu; Zou, Yubin, E-mail: zouyubin@pku.edu.cn; Wen, Weiwei; Lu, Yuanrong; Guo, Zhiyu

2016-07-01

Peking University Neutron Imaging Facility (PKUNIFTY) works on an accelerator–based neutron source with a repetition period of 10 ms and pulse duration of 0.4 ms, which has a rather low Cd ratio. To improve the effective Cd ratio and thus improve the detection capability of the facility, energy-filtering neutron imaging was realized with the intensified CCD camera and time-of-flight (TOF) method. Time structure of the pulsed neutron source was firstly simulated with Geant4, and the simulation result was evaluated with experiment. Both simulation and experiment results indicated that fast neutrons and epithermal neutrons were concentrated in the first 0.8 ms of each pulse period; meanwhile in the period of 0.8–2.0 ms only thermal neutrons existed. Based on this result, neutron images with and without energy filtering were acquired respectively, and it showed that detection capability of PKUNIFTY was improved with setting the exposure interval as 0.8–2.0 ms, especially for materials with strong moderating capability.

Studsvik thermal neutron facility

International Nuclear Information System (INIS)

Pettersson, O.A.; Larsson, B.; Grusell, E.; Svensson, P.

1992-01-01

The Studsvik thermal neutron facility at the R2-0 reactor originally designed for neutron capture radiography has been modified to permit irradiation of living cells and animals. A hole was drilled in the concrete shielding to provide a cylindrical channel with diameter of 25.3 cm. A shielding water tank serves as an entry holder for cells and animals. The advantage of this modification is that cells and animals can be irradiated at a constant thermal neutron fluence rate of approximately 10 9 n cm -2 s -1 (at 100 kW) without stopping and restarting the reactor. Topographic analysis of boron done by neutron capture autoradiography (NCR) can be irradiated under the same conditions as previously

Television imaging system for fast neutron radiography using baby cyclotron

International Nuclear Information System (INIS)

Yoshii, Koji; Miya, Kenzo; Katoh, Norihiko.

1993-01-01

A television imaging system for fast neutron radiography (FNR-TV) developed using the fast neutron source reactor YAYOI was applied to the baby-cyclotron based fast neutron source to get images of thick objects quickly. In the system the same technique as a current television imaging system of thermal neutron radiography was applied, while the luminescent converter was used to detect fast neutrons. Using the CR39 track etch method it took about 7 h to get an image, while the FNR-TV only 20 s enough for taking the same object. However the FNR-TV imaging result of the simulation model of a large explosive device for the space launch vehicle of H-2 type was not so good as the image taken with the CR39 track etch method. The reason was that the luminescence intensity of the FNR-TV converter was a quarter of that in the YAYOI. (author)

Neutron imaging system based on a video camera

International Nuclear Information System (INIS)

Dinca, M.

2004-01-01

The non-destructive testing with cold, thermal, epithermal or fast neutrons is nowadays more and more useful because the world-wide level of industrial development requires considerably higher standards of quality of manufactured products and reliability of technological processes especially where any deviation from standards could result in large-scale catastrophic consequences or human loses. Thanks to their properties, easily obtained and very good discrimination of the materials that penetrate, the thermal neutrons are the most used probe. The methods involved for this technique have advanced from neutron radiography based on converter screens and radiological films to neutron radioscopy based on video cameras, that is, from static images to dynamic images. Many neutron radioscopy systems have been used in the past with various levels of success. The quality of an image depends on the quality of the neutron beam and the type of the neutron imaging system. For real time investigations there are involved tube type cameras, CCD cameras and recently CID cameras that capture the image from an appropriate scintillator through the agency of a mirror. The analog signal of the camera is then converted into digital signal by the signal processing technology included into the camera. The image acquisition card or frame grabber from a PC converts the digital signal into an image. The image is formatted and processed by image analysis software. The scanning position of the object is controlled by the computer that commands the electrical motors that move horizontally, vertically and rotate the table of the object. Based on this system, a lot of static image acquisitions, real time non-destructive investigations of dynamic processes and finally, tomographic investigations of the small objects are done in a short time. A system based on a CID camera is presented. Fundamental differences between CCD and CID cameras lie in their pixel readout structure and technique. CIDs

Non-Destructive Spent Fuel Characterization with Semiconducting Gallium Arsinde Neutron Imaging Arrays

International Nuclear Information System (INIS)

McGregor, Douglas S.; Gersch, Holly K.; Sanders, Jeffrey D.; Lee, John C.; Hammig, Mark D.; Hartman, Michael R.; Yong Hong Yang; Klann, Raymond T.; Elzen, Brian Van Der; Lindsay, John T.; Simpson, Philip A.

2002-01-01

High resistivity bulk grown GaAs has been used to produce thermal neutron imaging devices for use in neutron radiography and characterizing burnup in spent fuel. The basic scheme utilizes a portable Sb/Be source for monoenergetic (24 keV) neutron radiation source coupled to an Fe filter with a radiation hard B-coated pixellated GaAs detector array as the primary neutron detector. The coated neutron detectors have been tested for efficiency and radiation hardness in order to determine their fitness for the harsh environments imposed by spent fuel. Theoretical and experimental results are presented, showing detector radiation hardness, expected detection efficiency and the spatial resolution from such a scheme. A variety of advanced neutron detector designs have been explored, with experimental results achieving 13% thermal neutron detection efficiency while projecting the possibility of over 30% thermal neutron detection efficiency

Preliminary examination of the applicability of imaging plates to fast neutron radiography

International Nuclear Information System (INIS)

Matsubayashi, Masahito; Hibiki, Takashi; Mishima, Kaichiro; Yoshii, Koji; Okamoto, Koji

2001-01-01

Fast neutron radiography is an attractive non-destructive inspection technique because of the excellent penetration characteristics of fast neutrons in matter. However, the difficulty of detecting fast neutrons reduces this attractive feature. As an experiment to overcome the difficulty, imaging plates were applied to fast neutron radiography. A simple combination of two sheets of imaging plates and a sheet of polyethylene as a proton emitter was examined with the (fast neutron, thermal neutron and gamma ray) FTG discriminator proposed by Yoneda et al. . The experimental results showed that the method could be applicable to fast neutron radiography with effective discrimination of γ-rays

LISe pixel detector for neutron imaging

Energy Technology Data Exchange (ETDEWEB)

Herrera, Elan; Hamm, Daniel [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN (United States); Wiggins, Brenden [Technology Development, Y-12 National Security Complex, Oak Ridge, TN (United States); Department of Physics and Astronomy, Vanderbilt University, Nashville, TN (United States); Milburn, Rob [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN (United States); Burger, Arnold [Department of Physics and Astronomy, Vanderbilt University, Nashville, TN (United States); Department of Life and Physical Sciences, Fisk University, Nashville, TN (United States); Bilheux, Hassina [Chemical and Engineering Materials Division, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Santodonato, Louis [Instrument and Source Division, Oak Ridge National Laboratory, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Chvala, Ondrej [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN (United States); Stowe, Ashley [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN (United States); Technology Development, Y-12 National Security Complex, Oak Ridge, TN (United States); Department of Physics and Astronomy, Vanderbilt University, Nashville, TN (United States); Lukosi, Eric, E-mail: elukosi@utk.edu [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN (United States)

2016-10-11

Semiconducting lithium indium diselenide, {sup 6}LiInSe{sub 2} or LISe, has promising characteristics for neutron detection applications. The 95% isotopic enrichment of {sup 6}Li results in a highly efficient thermal neutron-sensitive material. In this study, we report on a proof-of-principle investigation of a semiconducting LISe pixel detector to demonstrate its potential as an efficient neutron imager. The LISe pixel detector had a 4×4 of pixels with a 550 µm pitch on a 5×5×0.56 mm{sup 3} LISe substrate. An experimentally verified spatial resolution of 300 µm was observed utilizing a super-sampling technique.

DUNBID, the Delft University neutron backscattering imaging detector

International Nuclear Information System (INIS)

Bom, V.R.; Eijk, C.W.E. van; Ali, M.A.

2005-01-01

In the search for low-metallic land mines, the neutron backscattering technique may be applied if the soil is sufficiently dry. An advantage of this method is the speed of detection: the scanning speed may be made comparable to that of a metal detector. A two-dimensional position sensitive detector is tested to obtain an image of the back scattered thermal neutron radiation. Results of experiments using a radionuclide neutron source are presented. The on-mine to no-mine signal ratio can be improved by the application of a window on the neutron time-of-flight. Results using a pulsed neutron generator are also presented

Fast neutron radiography using photoluminescent imaging plates

International Nuclear Information System (INIS)

Rant, J.; Kristof, E.; Balasko, M.; Stade, J.

1999-01-01

Fast neutron radiography (FNR) and resonance neutron radiography (RNR) are complementary to the conventional radiography with high energy gamma-rays or brems-strahlung radiation used for the inspection of thick metal objects. In both non-destructive methods, the contrast sensitivity and the penetration power can be improved by using higher energy neutrons. At present direct techniques based either n Solid State Nuclear Track detectors (SSNTDs) or scintillating screens and transfer techniques using activation threshold detectors and radiographic films are applied for the detection of fast neutron images. Rather low detection sensitivity of film and SSNTD based fast neutron imaging methods and also rather poor inherent image contrast of SSNTD pose a problem for FNR in the fast neutron energy region 1-15 MeV interesting for NDT. For more efficient detection of fast neutron images the use of novel highly sensitive photoluminescent imaging plates (IP) in combination with threshold at the KFKI research reactor. The conventional IP produced by FUJI Photo Film Co. for the detection of beta and X-ray radiation were used. The threshold activation detectors were the reactions 115 In(n, n') 115m In, 64 Zn(n,p) 64 Cu, 56 Fe(n, p) 56 Mn, 24 Mg(n, p) 24 Na and 27 Al(n, α) 24 Na. These threshold reactions cover the fast neutron energy region between 0,7 MeV and 12 MeV. Pure, commercially available metals 0,1 mm to 0,25 mm thick made of In, Zn, Fe, Mg and Al were used as converter screens. The very high sensitivity of IP, the linearity of their response over 5 decades of exposure dose and the high dynamic digitalisation latitude enabled fast neutron radiography of image quality comparable to the quality of thermal NR. In our experimental conditions (φ n ∼ 10 8 n/cm 2 s, R Cd ∼ 2) the neutron exposure and IP exposure periods were still practical and comparable to the half life of the corresponding reaction products (half an hour to several hours). Even with the 27 Al(n.α) 24

Detection of 10B distributions in histological samples by NCAR using thermal and cold neutrons and photoluminiscent imaging plates. New results

International Nuclear Information System (INIS)

Rant, J.; Skvarc, J.; Ilic, R.; Gabel, D.; Bayon, G.; Yanagie, H.; Kobayashi, H.; Lehmann, E.; Kuehne, G.

1999-01-01

The Neutron Capture Autoradiography (NCAR) using various Solid State Nuclear Track Detectors (SSNTDs) is a well established and accurate method to detect and measure the distributions of 10 B in the ppm range on macroscopic and microscopic level in biological samples, such as histological sections of tumours loaded with 10 B compounds used for BNCT (e.g. 1,2). recently a new technique of NCAR using sensitive photoluminescent Imaging Plates (IP) has been proposed to detect 10 B distributions in histological sections (3), exploiting excellent detection properties of IP systems such as very high detection sensitivity and quantum detection efficiency, broad linear response and dynamic range, very small image distortion, reusability of IP and possibilities of digital autoradiography. The advantage of IP-NCAR vs. NCAR with SSNTDs should be the much lower neutron fluence (10 7 10 9 vs. 10 10 10 13 n/cm 2 with SSNTDs), no intermediate chemical treatment (track etching) and direct and fast compuitational handling and evaluation of the digitized autoradiographic image. However, the spatial resolution of the present available IP detection systems is somewhat lower (∼ 0,04 mm) than with SSNTDs (∼ 0,01 mm). Another problem with IP NCAR is rather high sensitivity of IP to all types of ionizing radiations. Therefore the background of direct and induced gamma-rays as well as of epithermal and fast neutrons has to be filtered out of thermal neutron beam to be used for IP-NCAR. To improve the signal/background ratio and to increase the detectibility of 10 B we propose to use clean cold neutron beams for the IP-NCAR of 10 B distributions in histological samples in BNCT experiments (4,5). In the present work the recent results of experiments in IP-NCAR with cold neutrons from the neutron radiographic channel of the ORPHEE reactor in Saclay and with the rather clean thermal neutron beam of the NEUTRA neutron radiography facility of the PSI (Villingen) will be presented. For the

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Use and imaging performance of CMOS flat panel imager with LiF/ZnS(Ag) and Gadox scintillation screens for neutron radiography

Science.gov (United States)

Cha, B. K.; kim, J. Y.; Kim, T. J.; Sim, C.; Cho, G.; Lee, D. H.; Seo, C.-W.; Jeon, S.; Huh, Y.

2011-01-01

In digital neutron radiography system, a thermal neutron imaging detector based on neutron-sensitive scintillating screens with CMOS(complementary metal oxide semiconductor) flat panel imager is introduced for non-destructive testing (NDT) application. Recently, large area CMOS APS (active-pixel sensor) in conjunction with scintillation films has been widely used in many digital X-ray imaging applications. Instead of typical imaging detectors such as image plates, cooled-CCD cameras and amorphous silicon flat panel detectors in combination with scintillation screens, we tried to apply a scintillator-based CMOS APS to neutron imaging detection systems for high resolution neutron radiography. In this work, two major Gd2O2S:Tb and 6LiF/ZnS:Ag scintillation screens with various thickness were fabricated by a screen printing method. These neutron converter screens consist of a dispersion of Gd2O2S:Tb and 6LiF/ZnS:Ag scintillating particles in acrylic binder. These scintillating screens coupled-CMOS flat panel imager with 25x50mm2 active area and 48μm pixel pitch was used for neutron radiography. Thermal neutron flux with 6x106n/cm2/s was utilized at the NRF facility of HANARO in KAERI. The neutron imaging characterization of the used detector was investigated in terms of relative light output, linearity and spatial resolution in detail. The experimental results of scintillating screen-based CMOS flat panel detectors demonstrate possibility of high sensitive and high spatial resolution imaging in neutron radiography system.

Neutron Thermalization and Reactor Spectra. Vol. II. Proceedings of the Symposium on Neutron Thermalization and Reactor Spectra

International Nuclear Information System (INIS)

1968-01-01

Proceedings of a Symposium organized by the IAEA and held at Ann Arbor, Michigan, USA, 17 - 21 July 1967. The meeting was attended by 143 participants from 24 Member States and one international organization. Contents: (Vol.I) Theory of neutron thermalization (15 papers); Scattering law (20 papers); Angular, space, temperature and time dependence of neutron spectra (9 papers). (Vol.II) Measurement of thermal neutron spectra and spectral indices, and comparison with theory (17 papers); Time-dependent problems in neutron thermalization (12 papers). Each paper is in its original language (61 English, 1 French and 11 Russian) and is preceded by an abstract in English with one in the original language if this is not English. Discussions are in English.

New neutron imaging using pulsed sources. Characteristics of a pulsed neutron source and principle of pulsed neutron imaging

International Nuclear Information System (INIS)

Kiyanagi, Yoshiaki

2012-01-01

Neutron beam is one of important tools to obtain the transmission image of an object. Until now, steady state neutron sources such as reactors are mainly used for this imaging purpose. Recently, it has been demonstrated that pulsed neutron imaging based on accelerator neutron sources can provide a real-space distribution of physical information of materials such as crystallographic structure, element, temperature, hydrogen bound state, magnetic field and so on, by analyzing wavelength dependent transmission spectrum, which information cannot be observed or difficult to obtain with a traditional imaging method using steady state neutrons. Here, characteristics of the pulsed neutron source and principle of the pulsed neutron imaging are explained as a basic concept of the new method. (author)

Thermal Neutron Capture and Thermal Neutron Burn-up of K isomeric state of 177mLu: a way to the Neutron Super-Elastic Scattering cross section

International Nuclear Information System (INIS)

Roig, O.; Belier, G.; Meot, V.; Daugas, J.-M.; Romain, P.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Letourneau, A.; Marie, F.; Veyssiere, Ch.

2006-01-01

Thermal neutron radiative capture and burn-up measurements of the K isomeric state in 177Lu form part of an original method to indirectly obtain the neutron super-elastic scattering cross section at thermal energy. Neutron super-elastic scattering, also called neutron inelastic acceleration, occurs during the neutron collisions with an excited nuclear level. In this reaction, the nucleus could partly transfer its excitation energy to the scattered neutron

Computed tomography with thermal neutrons and gaseous position sensitive detector; Tomografia computadorizada com neutrons termicos e detetor a gas sensivel a posicao

Energy Technology Data Exchange (ETDEWEB)

Souza, Maria Ines Silvani

2001-12-01

A third generation tomographic system using a parallel thermal neutron beam and gaseous position sensitive detector has been developed along three discrete phases. At the first one, X-ray tomographic images of several objects, using a position sensitive detector designed and constructed for this purpose have been obtained. The second phase involved the conversion of that detector for thermal neutron detection, by using materials capable to convert neutrons into detectable charged particles, testing afterwards its performance in a tomographic system by evaluation the quality of the image arising from several test-objects containing materials applicable in the engineering field. High enriched {sup 3} He, replacing the argon-methane otherwise used as filling gas for the X-ray detection, as well as, a gadolinium foil, have been utilized as converters. Besides the pure enriched {sup 3} He, its mixture with argon-methane and later on with propane, have been also tested, in order to evaluate the detector efficiency and resolution. After each gas change, the overall performance of the tomographic system using the modified detector, has been analyzed through measurements of the related parameters. This was done by analyzing the images produced by test-objects containing several materials having well known attenuation coefficients for both thermal neutrons and X-rays. In order to compare the performance of the position sensitive detector as modified to detect thermal neutrons, with that of a conventional BF{sub 3} detector, additional tomographs have been conducted using the last one. The results have been compared in terms of advantages, handicaps and complementary aspects for different kinds of radiation and materials. (author)

Characterization of a neutron imaging setup at the INES facility

Energy Technology Data Exchange (ETDEWEB)

Durisi, E.A., E-mail: elisabettaalessandra.durisi@unito.it [Università di Torino, Dipartimento di Fisica, Via Pietro Giuria 1, 10125 Torino (Italy); Istituto Nazionale di Fisica Nucleare—Sezione di Torino, Via Pietro Giuria 1, 10125 Torino (Italy); Visca, L. [Università di Torino, Dipartimento di Fisica, Via Pietro Giuria 1, 10125 Torino (Italy); Istituto Nazionale di Fisica Nucleare—Sezione di Torino, Via Pietro Giuria 1, 10125 Torino (Italy); Albertin, F.; Brancaccio, R. [Istituto Nazionale di Fisica Nucleare—Sezione di Torino, Via Pietro Giuria 1, 10125 Torino (Italy); Corsi, J. [Università di Torino, Dipartimento di Fisica, Via Pietro Giuria 1, 10125 Torino (Italy); Istituto Nazionale di Fisica Nucleare—Sezione di Torino, Via Pietro Giuria 1, 10125 Torino (Italy); Dughera, G. [Istituto Nazionale di Fisica Nucleare—Sezione di Torino, Via Pietro Giuria 1, 10125 Torino (Italy); Ferrarese, W. [Università di Torino, Dipartimento di Fisica, Via Pietro Giuria 1, 10125 Torino (Italy); Istituto Nazionale di Fisica Nucleare—Sezione di Torino, Via Pietro Giuria 1, 10125 Torino (Italy); Giovagnoli, A.; Grassi, N. [Fondazione Centro per la Conservazione ed il Restauro dei Beni Culturali “La Venaria Reale”, Piazza della Repubblica, 10078 Venaria Reale, Torino (Italy); Grazzi, F. [Consiglio Nazionale delle Ricerche, Istituto dei Sistemi Complessi, Via Madonna del Piano 10, 50019 Sesto Fiorentino, Firenze (Italy); Lo Giudice, A.; Mila, G. [Università di Torino, Dipartimento di Fisica, Via Pietro Giuria 1, 10125 Torino (Italy); Istituto Nazionale di Fisica Nucleare—Sezione di Torino, Via Pietro Giuria 1, 10125 Torino (Italy); and others

2013-10-21

The Italian Neutron Experimental Station (INES) located at the ISIS pulsed neutron source (Didcot, United Kingdom) provides a thermal neutron beam mainly used for diffraction analysis. A neutron transmission imaging system was also developed for beam monitoring and for aligning the sample under investigation. Although the time-of-flight neutron diffraction is a consolidated technique, the neutron imaging setup is not yet completely characterized and optimized. In this paper the performance for neutron radiography and tomography at INES of two scintillator screens read out by two different commercial CCD cameras is compared in terms of linearity, signal-to-noise ratio, effective dynamic range and spatial resolution. In addition, the results of neutron radiographies and a tomography of metal alloy test structures are presented to better characterize the INES imaging capabilities of metal artifacts in the cultural heritage field. -- Highlights: A full characterization of the present INES imaging set-up was carried out. Two CCD cameras and two scintillators (ZnS/{sup 6}LiF) of different thicknesses were tested. Linearity, effective dynamic range and spatial resolution were determined. Radiographies of steep wedges were performed using the highest dynamic range setup. Tomography of a bronze cube was performed using the best spatial resolution setup.

Characterization of a neutron imaging setup at the INES facility

International Nuclear Information System (INIS)

Durisi, E.A.; Visca, L.; Albertin, F.; Brancaccio, R.; Corsi, J.; Dughera, G.; Ferrarese, W.; Giovagnoli, A.; Grassi, N.; Grazzi, F.; Lo Giudice, A.; Mila, G.

2013-01-01

The Italian Neutron Experimental Station (INES) located at the ISIS pulsed neutron source (Didcot, United Kingdom) provides a thermal neutron beam mainly used for diffraction analysis. A neutron transmission imaging system was also developed for beam monitoring and for aligning the sample under investigation. Although the time-of-flight neutron diffraction is a consolidated technique, the neutron imaging setup is not yet completely characterized and optimized. In this paper the performance for neutron radiography and tomography at INES of two scintillator screens read out by two different commercial CCD cameras is compared in terms of linearity, signal-to-noise ratio, effective dynamic range and spatial resolution. In addition, the results of neutron radiographies and a tomography of metal alloy test structures are presented to better characterize the INES imaging capabilities of metal artifacts in the cultural heritage field. -- Highlights: A full characterization of the present INES imaging set-up was carried out. Two CCD cameras and two scintillators (ZnS/ 6 LiF) of different thicknesses were tested. Linearity, effective dynamic range and spatial resolution were determined. Radiographies of steep wedges were performed using the highest dynamic range setup. Tomography of a bronze cube was performed using the best spatial resolution setup

Neutron Imaging at Compact Accelerator-Driven Neutron Sources in Japan

Directory of Open Access Journals (Sweden)

Yoshiaki Kiyanagi

2018-03-01

Full Text Available Neutron imaging has been recognized to be very useful to investigate inside of materials and products that cannot be seen by X-ray. New imaging methods using the pulsed structure of neutron sources based on accelerators has been developed also at compact accelerator-driven neutron sources and opened new application fields in neutron imaging. The world’s first dedicated imaging instrument at pulsed neutron sources was constructed at J-PARC in Japan owing to the development of such new methods. Then, usefulness of the compact accelerator-driven neutron sources in neutron science was recognized and such facilities were newly constructed in Japan. Now, existing and new sources have been used for neutron imaging. Traditional imaging and newly developed pulsed neutron imaging such as Bragg edge transmission have been applied to various fields by using compact and large neutron facilities. Here, compact accelerator-driven neutron sources used for imaging in Japan are introduced and some of their activities are presented.

Neutron beams. Tracks analysis, imaging and medicine

International Nuclear Information System (INIS)

Pepy, G.

2006-01-01

Thermal neutron beams can supply informations about the arrangement of atoms and molecules and about their movement inside the matter. This article treats of the preparation of thermal neutron beams and of the applications that use their penetration and matter activation properties: 1 - thermal neutrons production; 2 - basic properties of thermal neutrons: neutrons scattering, absorbing materials, activating materials, transparent materials, preparation of a neutron beam; 3 - tracks measurement by activation: activation method, measurement of marine pollution by heavy elements, historical evolution of glass composition; 4 - neutron radiography: neutronography, neutronoscopy: viscosity measurement; 5 - cancer treatment. (J.S.)

Attenuation of thermal neutron through graphite

International Nuclear Information System (INIS)

Adib, M.; Ismaail, H.; Fathaallah, M.; Abbas, Y.; Habib, N.; Wahba, M.

2004-01-01

Calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections as a function of graphite temperature and crystalline from for neutron energies from 1 me V< E<10 eV were carried out. Computer programs have been developed which allow calculation for the graphite hexagonal closed-pack structure in its polycrystalline form and pyrolytic one. I The calculated total cross-section for polycrystalline graphite were compared with the experimental values. An overall agreement is indicated between the calculated values and experimental ones. Agreement was also obtained for neutron cross-section measured for oriented pyrolytic graphite at room and liquid nitrogen temperatures. A feasibility study for use of graphite in powdered form as a cold neutron filter is details. The calculated attenuation of thermal neutrons through large mosaic pyrolytic graphite show that such crystals can be used effectively as second order filter of thermal neutron beams and that cooling improve their effectiveness

Thermal neutron diffusion parameters in homogeneous mixtures

Energy Technology Data Exchange (ETDEWEB)

Drozdowicz, K.; Krynicka, E. [Institute of Nuclear Physics, Cracow (Poland)

1995-12-31

A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs.

Thermal neutron diffusion parameters in homogeneous mixtures

International Nuclear Information System (INIS)

Drozdowicz, K.; Krynicka, E.

1995-01-01

A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs

Trial production of hyper-thermal neutron generator for Neutron Capture Therapy (NCT) and its radiation properties

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Toru

1999-01-01

In NCT, it was at first important to give a cancer portion to radiation dose required for its recovery. By finding out that whole cross-section of water comprising of a living body decreased monotonously with increase of neutron energy from about 100 barn against thermal neutron, became about 40 barn at about 0.5 eV and kept constant to 40 barn till at about 100 eV, application of thermal neutron shifted to higher temperature side, called Hyper thermal neutron, to NCT is proposed. The Hyper thermal neutron radiation can be expected to have similar controllability to that of the thermal neutron radiation. In 1977 fiscal year, a trial Hyper thermal neutron generator was produced on a base of up-to-date investigation results. As a part of property evaluation of the generator, evaluation of energy spectra in the Hyper thermal neutron generated at LINAC by TOF was conducted to confirm shift of the spectra to high temperature side. And, a Fantom experiment at KUR heavy water neutron radiation facility was also conducted to confirm effect of improvement in deep portion dose distribution. (G.K.)

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

Smith, H.D. Jr.; Smith, M.P.; Schultz, W.E.

1979-01-01

A simultaneous pulsed neutron porosity and thermal neutron capture cross section logging system is provided for radiological well logging of subsurface earth formations. A logging tool provided with a 14 MeV pulsed neutron source, an epithermal neutron detector, and a combination gamma ray and fast neutron detector is moved through a borehole. Repetitive bursts of neutrons irradiate the earth formations; and, during the bursts, the fast neutron and epithermal neutron populations are sampled. During the interval between bursts the thermal neutron capture gamma ray population is sampled in two or more time intervals. The fast and epithermal neutron population measurements are combined to provide a measurement of formation porosity phi. The capture gamma ray measurements are combined to provide a simultaneous determination of the thermal neutron capture cross section Σ

Neutron imaging plates

International Nuclear Information System (INIS)

Niimura, Nobuo

1995-01-01

Imaging plates have been used in the field of medical diagnosis since long ago, but their usefulness was verified as the two-dimensional detector for analyzing the X-ray crystalline structure of high bio molecules like protein, and they have contributed to the remarkable progress in this field. The great contribution is due to the excellent features, such as the detection efficiency of about 100%, the positional resolution smaller than 0.2 mm, the dynamic range of five digits, and the area of several hundreds mm square. The neutron imaging plates have not yet obtained the sufficient results. It was planned to construct the neutron diffractometer for biological matters, and to put imaging plate neutron detectors (IP-ND) to practical use as the detector. The research on the development of IP-NDs was carried out, and the IPp-NDs having the performance comparable with that for X-ray were able to be produced. Imaging plates are the integral type two-dimensional radiation detector using photostimulated luminescence matters, and their principle is explained. As to neutron imaging plates, the converter, neutron detection efficiency and the flight of secondary particles in photo-stimulated luminescence matters are described. As for the present state of development of neutron imaging plates, the IP-NDs made for trial, the dynamic range, the positional resolution, the detection efficiency and the kinds of converters, and the application of IP-NDs are reported. (K.I.)

Three frequency modulated combination thermal neutron lifetime log and porosity

International Nuclear Information System (INIS)

Paap, H.J.; Arnold, D.M.; Smith, M.P.

1976-01-01

Methods are disclosed for measuring simultaneously the thermal neutron lifetime of the borehole fluid and earth formations in the vicinity of a well borehole, together with the formation porosity. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different modulation frequencies. Intensity modulated clouds of thermal neutrons at each of the three modulation frequencies are detected by dual spaced detectors and the relative phase shift of the thermal neutrons with respect to the fast neutrons is determined at each of the three modulation frequencies at each detector. These measurements are then combined to determine simultaneously the thermal neutron decay time of the borehole fluid, the thermal neutron decay time of surrounding earth formation media and the porosity of the formation media

Problems and prospects of neutron imaging

International Nuclear Information System (INIS)

Kobayashi, Hisao

2008-01-01

Technical problems and future prospects of neutron imaging and neutron radiography are reviewed and discussed for further development. For technical problems, neutron sources together with cold neutron, ultra-cold neutron, epithermal and fast-neutron beams, energy converters, and the intensity of neutron beam, dynamic range associated with imaging procedure, etc, are reviewed. As standardization, such indicators as beam purity, sensitivity, image quality, and beam quality are discussed and limitation of neutron radiography is also presented. As neutron imaging has developed as a nondestructive testing technique in industrial applications, further problems and prospects of quality control and qualification to perform neutron radiography, standardization and international cooperation of neutron imaging are discussed. (S. Ohno)

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments.

Science.gov (United States)

Miller, Marcelo E; Sztejnberg, Manuel L; González, Sara J; Thorp, Silvia I; Longhino, Juan M; Estryk, Guillermo

2011-12-01

A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

Energy Technology Data Exchange (ETDEWEB)

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo [Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429, Argentina and CONICET, Av. Rivadavia 1917, Ciudad de Buenos Aires 1033 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina)

2011-12-15

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

International Nuclear Information System (INIS)

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo

2011-01-01

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and global

Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

Directory of Open Access Journals (Sweden)

Kaiser Krista

2016-01-01

Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

Development of a hybrid MSGC detector for thermal neutron imaging with a MHz data acquisition and histogramming system

CERN Document Server

Gebauer, B; Richter, G; Levchanovsky, F V; Nikiforov, A

2001-01-01

For thermal neutron imaging at the next generation of high-flux pulsed neutron sources a large area and fourfold segmented, hybrid, low-pressure, two-dimensional position sensitive, microstrip gas chamber detector, fabricated in a multilayer technology on glass substrates, is presently being developed, which utilizes a thin composite sup 1 sup 5 sup 7 Gd/CsI neutron converter. The present article focusses on the readout scheme and the data acquisition (DAQ) system. For position encoding, interpolating and fast multihit delay line based electronics is applied with up to eightfold sub-segmentation per geometrical detector segment. All signals, i.e. position, time-of-flight and pulse-height signals, are fed into deadtime-less 8-channel multihit TDC chips with 120 ps LSB via constant fraction and time-over-threshold discriminators, respectively. The multihit capability is utilized to raise the count rate limit in combination with a sum check algorithm for disentangling pulses from different events. The first vers...

Digital image processing in neutron radiography

International Nuclear Information System (INIS)

Koerner, S.

2000-11-01

computer. After extensive test measurements performed at Paul Scherrer Institute in Switzerland, the following detector components have been selected: a nitrogen cooled slow scan CCD-camera with 16 bit digitization (65536 gray levels). The chip array consists of (512x512) pixels with a pixel size of (24x24) m2. Depending on the required spatial resolution, it can be chosen between two different lenses, a 58 mm F1,2 and a 180 mm F2,8 lens. The detective area is then (22x22) cm2 or (5x5) cm2, respectively. After extensive test measurements with different neutron sensitive scintillator screens, a Levy Hill ZnS(Ag)-6Li scintillator has been selected. Reasonable exposure times at the thermal column of the TRIGA reactor at the Atominstitut (flux: 1,3x105 neutrons/(cm2s)) can get as low as 20 s per image. This means, that it is possible to perform a set of measurements necessary for neutron tomography (about 200 images) within several hours. Therefore, in a second working step, an entirely automated neutron tomography facility has been built at the Atominstitut with this detector. Digital Image Processing: Due to special detector properties of the CCD-camera NR detector, a standard image processing procedure has been developed that always has to be applied, when the CCD-detector is used. It consists of the following steps: white spot correction - dark frame correction and flat field correction. Radiation, which hits the CCD-chip causes an overflow of one or several pixels, which appears in the image as white spots. These disturbing spots have to be removed by means of digital image processing. Several filters have been tested, but the results were insufficient. Therefore, a new threshold-median-mean value filter was designed and a proper code was written in IDL (interactive data language). The new filter removes white spots very well by hardly blurring the images. A dark frame is an image made with closed camera shutter. It contains undesired detector signal caused by read out

Digital image processing in neutron radiography

International Nuclear Information System (INIS)

Koerner, S.

2000-11-01

computer. After extensive test measurements performed at Paul Scherrer Institute in Switzerland, the following detector components have been selected: a nitrogen cooled slow scan CCD-camera with 16 bit digitization (65536 gray levels). The chip array consists of (512x512) pixels with a pixel size of (24x24) m2. Depending on the required spatial resolution, it can be chosen between two different lenses, a 58 mm F1,2 and a 180 mm F2,8 lens. The detective area is then (22x22) cm2 or (5x5) cm2, respectively. After extensive test measurements with different neutron sensitive scintillator screens, a Levy Hill ZnS(Ag)-6Li scintillator has been selected. Reasonable exposure times at the thermal column of the TRIGA reactor at the Atominstitut (flux: 1,3x105 neutrons/(cm2s)) can get as low as 20s per image. This means, that it is possible to perform a set of measurements necessary for neutron tomography (about 200 images) within several hours. Therefore, in a second working step, an entirely automated neutron tomography facility has been built at the Atominstitut with this detector. Digital image processing: due to special detector properties of the CCD-camera NR detector, a standard image processing procedure has been developed that always has to be applied, when the CCD-detector is used. It consists of the following steps: white spot correction - dark frame correction and flat field correction. □-radiation, which hits the CCD-chip causes an overflow of one or several pixels, which appears in the image as white spots. These disturbing spots have to be removed by means of digital image processing. Several filters have been tested, but the results were insufficient. Therefore, a new threshold-median-mean value filter was designed and a proper code was written in IDL (interactive data language). The new filter removes white spots very well by hardly blurring the images. A dark frame is an image made with closed camera shutter. It contains undesired detector signal caused by read out

Detector for imaging and dosimetry of laser-driven epithermal neutrons by alpha conversion

Science.gov (United States)

Mirfayzi, S. R.; Alejo, A.; Ahmed, H.; Wilson, L. A.; Ansell, S.; Armstrong, C.; Butler, N. M. H.; Clarke, R. J.; Higginson, A.; Notley, M.; Raspino, D.; Rusby, D. R.; Borghesi, M.; Rhodes, N. J.; McKenna, P.; Neely, D.; Brenner, C. M.; Kar, S.

2016-10-01

An epithermal neutron imager based on detecting alpha particles created via boron neutron capture mechanism is discussed. The diagnostic mainly consists of a mm thick Boron Nitride (BN) sheet (as an alpha converter) in contact with a non-borated cellulose nitride film (LR115 type-II) detector. While the BN absorbs the neutrons in the thermal and epithermal ranges, the fast neutrons register insignificantly on the detector due to their low neutron capture and recoil cross-sections. The use of solid-state nuclear track detectors (SSNTD), unlike image plates, micro-channel plates and scintillators, provide safeguard from the x-rays, gamma-rays and electrons. The diagnostic was tested on a proof-of-principle basis, in front of a laser driven source of moderated neutrons, which suggests the potential of using this diagnostic (BN+SSNTD) for dosimetry and imaging applications.

Method for determining thermal neutron decay times of earth formations

International Nuclear Information System (INIS)

Arnold, D.M.

1976-01-01

A method is disclosed for measuring the thermal neutron decay time of earth formations in the vicinity of a well borehole. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different intensity modulation frequencies. The tangents of the relative phase angles of the fast neutrons and the resulting thermal neutrons at each of the three frequencies of modulation are measured. First and second approximations to the earth formation thermal neutron decay time are derived from the three tangent measurements. These approximations are then combined to derive a value for the true earth formation thermal neutron decay time

Activation measurements for thermal neutrons. Part J. Evaluation of thermal neutron transmission factors

International Nuclear Information System (INIS)

Egbert, Stephen D.

2005-01-01

In order to relate thermal neutron activation measurements in samples to the calculated free-in-air thermal neutron activation levels given in Chapter 3, use is made of sample transmission factors. Transmission factors account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. The procedures for calculation of TF's and example results are presented in this section. (author)

Thermal neutron albedo measurements for multilithic reflectors

International Nuclear Information System (INIS)

Mehboob, Khurram; Ahmed, Raheel; Ali, Majid; Tabassam, Uzma

2013-01-01

Highlights: • Measurement of thermal neuron albedo for multilithic reflectors. • Modeling of experiments in MATLAB. • Comparison of numerical calculated and experimental values. • Study of thermal neutron albedo in different multilayered shielding. - Abstract: An experimental measurement of the thermal neutron (0.025 eV) albedo (αth) has been carried out for multilithic shielding by using Am–Be neutron source and BF 3 detector. The measured saturation value for the thermal albedo of paraffin wax has been found to be 0.734 ± 0.020, which is in close agreement to the corresponding value 0.83 quoted in the literature. The thermal neutron albedo has been measured for the multilayered shielding in copper–wood, copper–aluminum, wood–paraffin and paraffin–iron combinations in horizontal geometric configurations. Modeling and numerical simulation have been carried out by developing a MATLAB code which solves the diffusion equation in order to calculate the experimental results. Good agreement has been found between the numerical calculated and experimental results. The uncertainties in the measurements have also been calculated based on error propagation of the underlying Poisson distribution

GEM-based thermal neutron beam monitors for spallation sources

International Nuclear Information System (INIS)

Croci, G.; Claps, G.; Caniello, R.; Cazzaniga, C.; Grosso, G.; Murtas, F.; Tardocchi, M.; Vassallo, E.; Gorini, G.; Horstmann, C.; Kampmann, R.; Nowak, G.; Stoermer, M.

2013-01-01

The development of new large area and high flux thermal neutron detectors for future neutron spallation sources, like the European Spallation Source (ESS) is motivated by the problem of 3 He shortage. In the framework of the development of ESS, GEM (Gas Electron Multiplier) is one of the detector technologies that are being explored as thermal neutron sensors. A first prototype of GEM-based thermal neutron beam monitor (bGEM) has been built during 2012. The bGEM is a triple GEM gaseous detector equipped with an aluminum cathode coated by 1μm thick B 4 C layer used to convert thermal neutrons to charged particles through the 10 B(n, 7 Li)α nuclear reaction. This paper describes the results obtained by testing a bGEM detector at the ISIS spallation source on the VESUVIO beamline. Beam profiles (FWHM x =31 mm and FWHM y =36 mm), bGEM thermal neutron counting efficiency (≈1%), detector stability (3.45%) and the time-of-flight spectrum of the beam were successfully measured. This prototype represents the first step towards the development of thermal neutrons detectors with efficiency larger than 50% as alternatives to 3 He-based gaseous detectors

Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques. 1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy. 1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-...

Attenuation of Thermal Neutrons by Crystalline Silicon

International Nuclear Information System (INIS)

Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

2002-01-01

A simple formula is given which allows to calculate the contribution of the total neutron cross - section including the Bragg scattering from different (hkt) planes to the neutron * transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy .A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500μ eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

Parameters measurement for the thermal neutron beam in the thermal column hole of Xi’an pulse reactor

Institute of Scientific and Technical Information of China (English)

无

2010-01-01

The distribution of the neutron spectra in the thermal column hole of Xi’an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.

Chemical warfare agents identification by thermal neutron detection

International Nuclear Information System (INIS)

Liu Boxue; Ai Xianyun; Tan Daoyuan; Zhang Dianqin

2000-01-01

The hydrogen concentration determination by thermal neutron detection is a non-destructive, fast and effective method to identify chemical warfare agents and TNT that contain different hydrogen fraction. When an isotropic neutron source is used to irradiate chemical ammunition, hydrogen atoms of the agent inside shell act as a moderator and slow down neutrons. The number of induced thermal neutrons depends mainly upon hydrogen content of the agent. Therefore measurement of thermal neutron influence can be used to determine hydrogen atom concentration, thereby to determine the chemical warfare agents. Under a certain geometry three calibration curves of count rate against hydrogen concentration were measured. According to the calibration curves, response of a chemical agent or TNT could be calculated. Differences of count rate among chemical agents and TNT for each kind of shells is greater than five times of standard deviations of count rate for any agent, so chemical agents or TNT could be identified correctly. Meanwhile, blast tube or liquid level of chemical warfare agent could affect the response of thermal neutron count rate, and thereby the result of identification. (author)

Study and development of new dosemeters for thermal neutrons

International Nuclear Information System (INIS)

Urena N, F.

1998-01-01

An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation (φ = 5 x 10 7 n/cm 2 -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence Φ = φ Δt ; where Δt = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Thermal neutron shield and method of manufacture

Science.gov (United States)

Brindza, Paul Daniel; Metzger, Bert Clayton

2013-05-28

A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

Experimental characterization of semiconductor-based thermal neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Bortot, D.; Pola, A.; Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN—Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Sacco, D. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); INAIL—DIT, Via di Fontana Candida 1, 00040 Monteporzio Catone (Italy); Esposito, A.; Gentile, A.; Buonomo, B. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Palomba, M.; Grossi, A. [ENEA Triga RC-1C.R. Casaccia, via Anguillarese 301, 00060 S. Maria di Galeria, Roma (Italy)

2015-04-21

In the framework of NESCOFI@BTF and NEURAPID projects, active thermal neutron detectors were manufactured by depositing appropriate thickness of {sup 6}LiF on commercially available windowless p–i–n diodes. Detectors with different radiator thickness, ranging from 5 to 62 μm, were manufactured by evaporation-based deposition technique and exposed to known values of thermal neutron fluence in two thermal neutron facilities exhibiting different irradiation geometries. The following properties of the detector response were investigated and presented in this work: thickness dependence, impact of parasitic effects (photons and epithermal neutrons), linearity, isotropy, and radiation damage following exposure to large fluence (in the order of 10{sup 12} cm{sup −2})

Thermal, epithermal and thermalized neutron attenuation properties of ilmenite-serpentine heat resistant concrete shield

International Nuclear Information System (INIS)

Kany, A.M.I.; El-Gohary, M.I.; Kamal, S.M.

1994-01-01

Experimental measurements were carried out to study the attenuation properties of low-energy neutrons transmitted through unheated and preheated barriers of heavy-weight, highly hydrated and heat-resistant concrete shields. The concrete shields under investigation have been prepared from naturally occurring ilmenite and serpentine Egyptian ores. A collimated beam obtained from an Am-Be source was used as a source of neutrons, while the measurements of total thermal, epithermal, and thermalized neutron fluxes were performed using a BF-3 detector, multichannel analyzer and Cd filter. Results show that the ilmenite-serpentine concrete proved to be a better thermal, epithermal and thermalized neutron attenuator than the ordinary concrete especially at a high temperature of concrete exposure. (Author)

Neutron image intensifier tubes

International Nuclear Information System (INIS)

Verat, M.; Rougeot, H.; Driard, B.

1983-01-01

The most frequently used techniques in neutron radiography employ a neutron converter consisting of either a scintillator or a thin metal sheet. The radiation created by the neutrons exposes a photographic film that is in contact with the converter: in the direct method, the film is exposed during the time that the object is irradiated with neutrons; in the transfer method, the film is exposed after the irradiation of the object with neutrons. In industrial non-destructive testing, when many identical objects have to be checked, these techniques have several disadvantages. Non-destructive testing systems without these disadvantages can be constructed around neutron-image intensifier tubes. A description and the operating characteristics of neutron-image intensifier tubes are given. (Auth.)

Development of temperature related thermal neutron scattering database for MCNP

International Nuclear Information System (INIS)

Mei Longwei; Cai Xiangzhou; Jiang Dazhen; Chen Jingen; Guo Wei

2013-01-01

Based on ENDF/B-Ⅶ neutron library, the thermal neutron scattering library S(α, β) for molten salt reactor moderators was developed. The temperatures of this library were chose as the characteristic temperature of the molten salt reactor. The cross section of the thermal neutron scattering of ACE format was investigated, and this library was also validated by the benchmarks of ICSBEP. The uncertainties shown in the validation were in reasonable range when compared with the thermal neutron scattering library tmccs which included in the MCNP data library. It was proved that the thermal neutron scattering library processed in this study could be used in the molten salt reactor design. (authors)

Neutron imaging integrated circuit and method for detecting neutrons

Science.gov (United States)

Nagarkar, Vivek V.; More, Mitali J.

2017-12-05

The present disclosure provides a neutron imaging detector and a method for detecting neutrons. In one example, a method includes providing a neutron imaging detector including plurality of memory cells and a conversion layer on the memory cells, setting one or more of the memory cells to a first charge state, positioning the neutron imaging detector in a neutron environment for a predetermined time period, and reading a state change at one of the memory cells, and measuring a charge state change at one of the plurality of memory cells from the first charge state to a second charge state less than the first charge state, where the charge state change indicates detection of neutrons at said one of the memory cells.

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1992-09-01

Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800 degree C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280 degree F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found

Thermal neutron standard fields with the KUR heavy water facility

International Nuclear Information System (INIS)

Kanda, K.; Kobayashi, K.; Shibata, T.

1978-01-01

A heavy water facility attached to the KUR (Kyoto University Reactor, swimming pool type, 5 MW) yields pure thermal neutrons in the Maxwellian distribution. The facility is faced to the core of KUR and it contains about 2 tons of heavy water. The thickness of the layer is about 140 cm. The neutron spectrum was measured with the time of flight technique using a fast chopper. The measured spectrum was in good agreement with the Maxwellian distribution in all energy region for thermal neutrons. The neutron temperature was slightly higher than the heavy water temperature. The contamination of epithermal and fast neutrons caused by photo-neutrons of the γ-n reaction of heavy water was very small. The maximum intensity of thermal neutrons is 3x10 11 n/cm 2 sec. When the bismuth scatterer is attached, the gamma rays contamination is eliminated by the ratio of 0.05 of gamma rays to neutrons in rem. This standard neutron field has been used for such experiments as thermal neutron cross section measurement, detector calibration, activation analysis, biomedical purposes etc. (author)

Thermalization of monoenergetic neutrons in a concrete room

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado, G.A. [UAZ, A.P. 336, 98000 Zacatecas (Mexico); Iniguez, M.P.; Martin M, A. [Universidad de Valladolid, (Spain)

2006-07-01

The thermalization of neutrons from monoenergetic neutron sources in a concrete room has been studied. During calibration of neutron detectors it is mandatory to make corrections due to neutron scattering produced by the room walls, therefore this factor must be known in advance. The scattered neutrons are thermalized and produce a neutron field that is directly proportional to source strength and inversely proportional to room total wall-surfaces, the proportional coefficient has been calculated for neutrons whose energy goes from 1 eV to 20 MeV. This coefficient was calculated using Monte Carlo methods for 150, 200 and 300 cm-radius spherical cavity, where monoenergetic neutrons were located at the center, along the spherical cavity radius neutron spectra were calculated at several source-to-detector distances inside the cavity. The obtained coefficient is almost three times larger than the factor normally utilized. (Author)

Study of a nTHGEM-based thermal neutron detector

Science.gov (United States)

Li, Ke; Zhou, Jian-Rong; Wang, Xiao-Dong; Xiong, Tao; Zhang, Ying; Xie, Yu-Guang; Zhou, Liang; Xu, Hong; Yang, Gui-An; Wang, Yan-Feng; Wang, Yan; Wu, Jin-Jie; Sun, Zhi-Jia; Hu, Bi-Tao

2016-07-01

With new generation neutron sources, traditional neutron detectors cannot satisfy the demands of the applications, especially under high flux. Furthermore, facing the global crisis in 3He gas supply, research on new types of neutron detector as an alternative to 3He is a research hotspot in the field of particle detection. GEM (Gaseous Electron Multiplier) neutron detectors have high counting rate, good spatial and time resolution, and could be one future direction of the development of neutron detectors. In this paper, the physical process of neutron detection is simulated with Geant4 code, studying the relations between thermal conversion efficiency, boron thickness and number of boron layers. Due to the special characteristics of neutron detection, we have developed a novel type of special ceramic nTHGEM (neutron THick GEM) for neutron detection. The performance of the nTHGEM working in different Ar/CO2 mixtures is presented, including measurements of the gain and the count rate plateau using a copper target X-ray source. A detector with a single nTHGEM has been tested for 2-D imaging using a 252Cf neutron source. The key parameters of the performance of the nTHGEM detector have been obtained, providing necessary experimental data as a reference for further research on this detector. Supported by National Natural Science Foundation of China (11127508, 11175199, 11205253, 11405191), Key Laboratory of Neutron Physics, CAEP (2013DB06, 2013BB04) and CAS (YZ201512)

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

Shultz, W.E.

1980-01-01

A method for simultaneously determining the porosity and thermal neutron capture cross-section of earth formations in the vicinity of a well borehole is claimed. It comprises the following steps: passing a well tool into a cased well borehole. The tool has a pulsed source of fast neutrons, a combination fast neutron and gamma ray detector and an epithermal neutron detector; repetitively irradiating the earth formations in the vicinity of the borehole with bursts of fast neutrons; detecting the fast neutron and epithermal neutron populations in the borehole (during the neutron bursts) and generating first and second measurement signals; detecting for second and third time intervals during the time between the neutron bursts, the gamma radiation present in the borehole due to the capture of thermalized neutrons by the nuclei of elements comprising the earth formations and generating third and fourth measurement signals; and combining the first and second measurement signals according to a predetermined relationship to derive an indication of the porosity of the earth formations and combining the third and fourth measurement signals to derive an indication of the thermal neutron capture cross-section of the earth formations

Thermal neutron detectors based on complex oxide crystals

CERN Document Server

Ryzhikov, V; Volkov, V; Chernikov, V; Zelenskaya, O

2002-01-01

The ways of improvement of spectrometric quality of CWO and GSO crystals have been investigated with the aim of their application in thermal neutron detectors based on radiation capture reactions. The efficiency of the neutron detection by these crystals was measured, and the obtained data were compared with the results for sup 6 LiI(Tl) crystals. It is shown that the use of complex oxide crystals and neutron-absorption filters for spectrometry of thermal and resonance neutrons could be a promising method in combination with computer data processing. Numerical calculations are reported for spectra of gamma-quanta due to radiation capture of the neutrons. To compensate for the gamma-background lines, we used a crystal pair of heavy complex oxides with different sensitivity to neutrons.

Image processing techniques for thermal, x-rays and nuclear radiations

International Nuclear Information System (INIS)

Chadda, V.K.

1998-01-01

The paper describes image acquisition techniques for the non-visible range of electromagnetic spectrum especially thermal, x-rays and nuclear radiations. Thermal imaging systems are valuable tools used for applications ranging from PCB inspection, hot spot studies, fire identification, satellite imaging to defense applications. Penetrating radiations like x-rays and gamma rays are used in NDT, baggage inspection, CAT scan, cardiology, radiography, nuclear medicine etc. Neutron radiography compliments conventional x-rays and gamma radiography. For these applications, image processing and computed tomography are employed for 2-D and 3-D image interpretation respectively. The paper also covers main features of image processing systems for quantitative evaluation of gray level and binary images. (author)

Using thermalizers in measuring 'Ukryttia' object's FCM neutron fluxes

CERN Document Server

Krasnyanskaya, O G; Odinokin, G I; Pavlovich, V N

2003-01-01

The results of research of a thermalizer (heater) width influence on neutron thermalization efficiency during FCM neutron flux measuring in the 'Ukryttia' are described. The calculations of neutron flux densities were performed by the Monte-Carlo method with the help of computer code MCNP-4C for FCM different models.Three possible installations of detectors were considered: on FCM surface,inside the FCM, and inside the concrete under the FCM layer. It was shown,that in order to increase the sensitivity of neutron detectors in intermediate and fast neutrons field,and consequently, to decrease the dependence of the readings of spectral distribution of neutron flux,it is necessary to position the detector inside the so-called thermalizer or heater. The most reasonable application of thick 'heaters' is the situation, when the detector is placed on FCM surface.

Fluence-compensated down-scattered neutron imaging using the neutron imaging system at the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Casey, D. T., E-mail: casey21@llnl.gov; Munro, D. H.; Grim, G. P.; Landen, O. L.; Spears, B. K.; Fittinghoff, D. N.; Field, J. E.; Smalyuk, V. A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Volegov, P. L.; Merrill, F. E. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

2016-11-15

The Neutron Imaging System at the National Ignition Facility is used to observe the primary ∼14 MeV neutrons from the hotspot and down-scattered neutrons (6-12 MeV) from the assembled shell. Due to the strong spatial dependence of the primary neutron fluence through the dense shell, the down-scattered image is convolved with the primary-neutron fluence much like a backlighter profile. Using a characteristic scattering angle assumption, we estimate the primary neutron fluence and compensate the down-scattered image, which reveals information about asymmetry that is otherwise difficult to extract without invoking complicated models.

Basic research of neutron radiography using cold neutron beam

International Nuclear Information System (INIS)

Oda, Masahiro; Tamaki, Masayoshi; Tasaka, Kanji

1995-01-01

As the result of demanding high quality images, now the nuclear reactors which can supply stably intense neutron beam have become the most general neutron source for radiography. For the purpose, mostly thermal neutrons have been used, but it is indispensable to use other neutrons than thermal neutrons for advancing neutron radiography technology and expanding the application fields. The radiography using cold neutrons is most behind in the development because the suitable neutron source was not available in Japan. The neutron sources for exclusively obtaining intense cold neutron beam were installed in the Kyoto University reactor in 1986 and in the JRR-3M of Japan Atomic Energy Research Institute in 1991. Basically as neutron energy lowers, the cross section of substances increases. In certain crystalline substances, the Bragg cutoff arises. The removal of scattered neutrons, the measurement of parallelism of beam and the relation of the thickness of objects with the transmissivity of cold neutrons are described. The imaging by TV method and the cold neutron CT in the CNRF and the simplified neutron CT by film method are reported. (K.I.)

A prototype detector using the neutron image intensifier and multi-anode type photomultiplier tube for pulsed neutron imaging

International Nuclear Information System (INIS)

Ishikawa, Hirotaku; Sato, Hirotaka; Hara, Kaoru Y.; Kamiyama, Takashi

2016-01-01

We developed a neutron two-dimensional (2-D) detector for pulsed neutron imaging as a prototype detector, which was composed of a neutron image intensifier and a multi-anode type photomultiplier tube. A neutron transmission spectrum of α-Fe plate was measured by the prototype detector, and compared with the one measured by a typical neutron 2-D detector. The spectrum was in reasonable agreement with the one measured by the typical detector in the neutron wavelength region above 0.15 nm. In addition, a neutron transmission image of a cadmium indicator was obtained by the prototype detector. The usefulness of the prototype detector for pulsed neutron imaging was demonstrated. (author)

LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

Characterization of two Japanese ancient swords through neutron imaging

International Nuclear Information System (INIS)

Grazzi, Francesco; Zoppi, Marco; Salvemini, Filomena; Kaestner, Anders; Lehmann, Eberhard; Civita, Francesco

2015-01-01

Japanese blades are culturally interesting objects both from the stylistic point of view and because of their fantastic performances. In this work, we present new results, using a non-invasive approach, concerning these peculiar artefacts. Two integer Japanese swords, pertaining to Kyoto (987-1596) and Shinto (1596-1781) periods have been analysed through neutron-imaging techniques. The experiments have been performed at the ICON beam line, operating at the spallation neutron source SINQ, Paul Scherrer Institut in Switzerland. The reconstruction of projection data into neutron tomographic slices or volumes, allowed us to identify some very peculiar characteristics, related to the forging methods and to the different thermal treatments applied to produce the cutting edge and its unique feature. (author)

Response of six neutron survey meters in mixed fields of fast and thermal neutrons.

Science.gov (United States)

Kim, S I; Kim, B H; Chang, I; Lee, J I; Kim, J L; Pradhan, A S

2013-10-01

Calibration neutron fields have been developed at KAERI (Korea Atomic Energy Research Institute) to study the responses of commonly used neutron survey meters in the presence of fast neutrons of energy around 10 MeV. The neutron fields were produced by using neutrons from the (241)Am-Be sources held in a graphite pile and a DT neutron generator. The spectral details and the ambient dose equivalent rates of the calibration fields were established, and the responses of six neutron survey meters were evaluated. Four single-moderator-based survey meters exhibited an under-responses ranging from ∼9 to 55 %. DINEUTRUN, commonly used in fields around nuclear reactors, exhibited an over-response by a factor of three in the thermal neutron field and an under-response of ∼85 % in the mixed fields. REM-500 (tissue-equivalent proportional counter) exhibited a response close to 1.0 in the fast neutron fields and an under-response of ∼50 % in the thermal neutron field.

Neutron Imaging Developments at LANSCE

Science.gov (United States)

Nelson, Ron; Hunter, James; Schirato, Richard; Vogel, Sven; Swift, Alicia; Ickes, Tim; Ward, Bill; Losko, Adrian; Tremsin, Anton

2015-10-01

Neutron imaging is complementary to x-ray imaging because of its sensitivity to light elements and greater penetration of high-Z materials. Energy-resolved neutron imaging can provide contrast enhancements for elements and isotopes due to the variations with energy in scattering cross sections due to nuclear resonances. These cross section differences exist due to compound nuclear resonances that are characteristic of each element and isotope, as well as broader resonances at higher energies. In addition, multi-probe imaging, such as combined photon and neutron imaging, is a powerful tool for discerning properties and features in materials that cannot be observed with a single probe. Recently, we have demonstrated neutron imaging, both radiography and computed tomography, using the moderated (Lujan Center) and high-energy (WNR facility) neutron sources at LANSCE. Flat panel x-ray detectors with suitable scintillator-converter screens provide good sensitivity for both low and high neutron energies. Micro-Channel-Plate detectors and iCCD scintillator camera systems that provide the fast time gating needed for energy-resolved imaging have been demonstrated as well. Examples of recent work will be shown including fluid flow in plants and imaging through dense thick objects. This work is funded by the US Department of Energy, National Nuclear Security Administration, and performed by Los Alamos National Security LLC under Contract DE-AC52-06NA25396.

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

Directory of Open Access Journals (Sweden)

Hu J.-P.

2016-01-01

Full Text Available Radiation dosimetry for Neutron Capture Therapy (NCT has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR. In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1 in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2 out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3 beam shutter upgrade to reduce strayed neutrons and gamma dose, (4 beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5 beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates to reduce prompt gamma and fast neutron doses, (6 sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7 holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4–7

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

Energy Technology Data Exchange (ETDEWEB)

Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N.

2014-05-23

Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D₂O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7

Computerized tomography using fast neutrons

International Nuclear Information System (INIS)

Maier-Schuler, P.

1992-03-01

The equipment is transportable and can be used at different neutron sources. CT-images are presented showing that it is possible to get good results by using CT with fast neutrons in non destructive testing. Small defects with high contrasts can be detected as well as larger defects with small differences in material density. Since the neutrons interact with the nuclei and not with the electron density the CT-images contain different information compared with X-ray or γ images. As neutron sources always emit γ-radiation too, this radiation can be detected simultaneously with the neutrons. Therefore one can get a γ CT-image along with the neutron image. For the examination of small samples or objects containing materials with great differences in the linear attenuation coefficients like Al and H 2 thermal neutrons have been used for CT-measurements too. A spatial resolution and a density resolution of 0.1 mm and about 5% respectively could be achieved in the CT-images with fast neutrons and 0.04 mm with thermal neutrons. (orig./HP) [de

A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2000-01-01

A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

Calculation of the thermal neutron flux depression in the loop VISA-1

International Nuclear Information System (INIS)

Martinc, R.

1961-01-01

Among other applications, the VISA-1 loop is to be used for thermal load testing of materials. For this type of testing one should know the maximum power generated in the loop. This power is determined from the maximum thermal neutron flux in the VK-5 channel and mean flux depression in the fissile component of the loop. Thermal neutron flux depression is caused by neutron absorption in the components of the loop, shape of the components and neutron leaking through gaps as well as properties of the surrounding medium of the core. All these parameters were taken into account for calculating the depression of thermal neutron flux in the VISA-1 loop. Two group diffusion theory was used. Fast neutron from the fission in the loop and slowed down were taken into account. Depression of the thermal neutron flux is expressed by depression factor which represents the ratio of the mean thermal neutron flux in the fissile loop component and the thermal neutron flux in the VK-5 without the loop. Calculation error was estimated and it was recommended to determine the depression factor experimentally as well [sr

Method and apparatus for measuring thermal neutron characteristics

International Nuclear Information System (INIS)

Johnstone, C.W.

1983-01-01

The thermal neutron decay characteristics of an earth formation are measured by detecting indications of the thermal neutron concentration in the formation during a selected set of two measurement intervals following irradiation of the formation with a burst of fast neutrons. These measurement intervals may comprise a sequence of time gates following a delay after the neutron burst. The duration of the neutron bursts, of the delay between the burst and the start of the sequence, and of the individual time gates, may all be adjusted by a common, selected one of a finite number of scale factor values. The set of two measurement intervals is selected from among a number of possible sets as a function of a previously measured value of the decay characteristic. Each measurement interval set is used over only a specific range of decay characteristic values for which it has been determined, in accordance with a previously established relationship between the decay characteristic value and a function of the thermal neutron concentration measurements for the set, to afford enhanced statistical accuracy in the measured value of the decay characteristic. (author)

A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

1999-01-01

In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

Status of thermal neutron scattering data for graphite

International Nuclear Information System (INIS)

Mattes, M.; Keinert, J.

2005-07-01

At thermal neutron energies, the binding of the scattering nucleus in a solid, liquid, or gas affects the cross sections and the angular and energy distributions of the scattered neutrons. These effects are described in the thermal sub-library of evaluated files in File 7 of the ENDF-6 format. A re-evaluation of thermal neutron scattering data for carbon bound in graphite has been performed to investigate the impact of models (e.g., generalised frequency distributions) based on different experimental and theoretical data for the generation of scattering law data files S(α,β,T) and coherent elastic scattering data. Two phonon frequency distributions of graphite published in 2002 and 2004 were considered and the results compared with those based on the phonon spectra from Koppel et al. (published in 1968), on which the evaluations of ENDF/B-VI and JEFF-3.1 are based. The new frequency distributions were partly derived from ab initio simulations. Detailed comparisons with measurements of differential and integral neutron cross sections and other relevant data are reported. In addition, thermal MCNP data sets for use in the continuous Monte Carlo codes MCNP and MCNPX were generated from these evaluations for different temperatures. Calculated neutron spectra were found to be in good agreement with the measurements. (author)

Applications of thermal neutron scattering

International Nuclear Information System (INIS)

Kostorz, G.

1978-01-01

Although in the past neutrons have been used quite frequently in the study of condensed matter, a more recent development has lead to applications of thermal neutron scattering in the investigation of more practical rather than purely academic problems. Physicists, chemists, materials scientists, biologists, and others have recognized and demonstrated that neutron scattering techniques can yield supplementary information which, in many cases, could not be obtained with other methods. The paper illustrates the use of neutron scattering in these areas of applied research. No attempt is made to present all the aspects of neutron scattering which can be found in textbooks. From the vast amount of experimental data, only a few examples are presented for the study of structure and atomic arrangement, ''extended'' structure, and dynamic phenomena in substances of current interest in applied research. (author)

Neutron radiography

International Nuclear Information System (INIS)

Hiraoka, Eiichi

1988-01-01

The thermal neutron absorption coefficient is essentially different from the X-ray absorption coefficient. Each substance has a characteristic absorption coefficient regardless of its density. Neutron deams have the following features: (1) neutrons are not transmitted efficiently by low molecular weight substances, (2) they are transmitted efficiently by heavy metals, and (3) the transmittance differs among isotopes. Thus, neutron beams are suitable for cheking for foreign matters in heavy metals and testing of composites consisting of both heavy and light materials. A neutron source generates fast neutrons, which should be converted into thermal neutrons by reducing their energy. Major neutron souces include nuclear reactors, radioisotopes and particle accelerators. Photographic films and television systems are mainly used to observe neutron transmission images. Computers are employed for image processing, computerized tomography and three-dimensional analysis. The major applications of neutron radiography include inspection of neclear fuel; evaluation of material for airplane; observation of fuel in the engine and oil in the hydraulic systems in airplanes; testing of composite materials; etc. (Nogami, K.)

Transmutation of Minor Actinide in well thermalized neutron field and application of advanced neutron source (ANS)

International Nuclear Information System (INIS)

Iwasaki, Tomohiko; Hirakawa, Naohiro

1995-01-01

Transmutation of Minor Actinide (MA) in a well thermalized neutron field was studied. Since MA nuclides have large effective cross sections in the well thermalized neutron field, the transmutation in the well thermalized neutron field has an advantage of high transmutation rate. However, the transmutation rate largely decreases by accumulation of 246 Cm when MA is transmuted only in the well thermalized neutron field for a long period. An acceleration method of burn-up of 246 Cm was studied. High transmutation rate can be obtained by providing a neutron field with high flux in the energy region between 1 and 100 eV. Two stage transmutation using the well thermalized neutron field and this field can transmute MA rapidly. The applicability of the Advanced Neutron Source (ANS) to the transmutation of MA was examined for a typical MA with the composition in the high-level waste generated in the conventional PWR. If the ANS is applied without changing the fuel inventory, the amount of MA which corresponds to that produced by a conventional 1,175 MWe PWR in one year can be transmuted by the ANS in one year. Furthermore, the amount of the residual can be reduced to about 1g (10 -5 of the initial MA weight) by continuing the transmutation for 5 years owing to the two stage transmutation. (author)

Nondestructive elemental analysis of coins using accelerator-based thermal neutrons

International Nuclear Information System (INIS)

Khairi, F.Z.; Aksoy, A.; Al-Haddad, M.N.

2007-01-01

The accelerator-based thermal-neutrons activation analysis setup at KFUPM has an adequate thermal -neutron flux that can be advantageously used for the elemental analysis of a variety of samples including archeological ones. The thermal neutrons are derived from the moderation of fast neutrons from the D (d, n) He reaction which produces fast 2.5 MeV neutrons. A maximum thermals flux of about 2.5x10 n/m-s was achieved. For the purpose of determining the suitability of the set up for the analysis of contemporary and ancient coins, we carried out a feasibility study by irradiating a selected number of Saudi Arabian coins dating from 1958 to 1987 in the thermal-neutron flux. The induced gamma-ray activities were then counted using a HP-GMX detector coupled to a PC-based data acquisition and analysis system. The elements that were determined in the coins were copper (75%), nickel (around 25%) and manganese (<0.5%). Calibration curves were also established for these elements. The determined concentrations are in agreement with the data published by the Standard Catalogue of World Coins. (author)

The neutron imaging diagnostic at NIF (invited).

Science.gov (United States)

Merrill, F E; Bower, D; Buckles, R; Clark, D D; Danly, C R; Drury, O B; Dzenitis, J M; Fatherley, V E; Fittinghoff, D N; Gallegos, R; Grim, G P; Guler, N; Loomis, E N; Lutz, S; Malone, R M; Martinson, D D; Mares, D; Morley, D J; Morgan, G L; Oertel, J A; Tregillis, I L; Volegov, P L; Weiss, P B; Wilde, C H; Wilson, D C

2012-10-01

A neutron imaging diagnostic has recently been commissioned at the National Ignition Facility (NIF). This new system is an important diagnostic tool for inertial fusion studies at the NIF for measuring the size and shape of the burning DT plasma during the ignition stage of Inertial Confinement Fusion (ICF) implosions. The imaging technique utilizes a pinhole neutron aperture, placed between the neutron source and a neutron detector. The detection system measures the two dimensional distribution of neutrons passing through the pinhole. This diagnostic has been designed to collect two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically the first image measures the distribution of the 14 MeV neutrons and the second image of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core.

The neutron imaging diagnostic at NIF (invited)

Energy Technology Data Exchange (ETDEWEB)

Merrill, F. E.; Clark, D. D.; Danly, C. R.; Drury, O. B.; Fatherley, V. E.; Gallegos, R.; Grim, G. P.; Guler, N.; Loomis, E. N.; Martinson, D. D.; Mares, D.; Morley, D. J.; Morgan, G. L.; Oertel, J. A.; Tregillis, I. L.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); Bower, D.; Dzenitis, J. M. [Livermore National Laboratory, Livermore, California 94550 (United States); and others

2012-10-15

A neutron imaging diagnostic has recently been commissioned at the National Ignition Facility (NIF). This new system is an important diagnostic tool for inertial fusion studies at the NIF for measuring the size and shape of the burning DT plasma during the ignition stage of Inertial Confinement Fusion (ICF) implosions. The imaging technique utilizes a pinhole neutron aperture, placed between the neutron source and a neutron detector. The detection system measures the two dimensional distribution of neutrons passing through the pinhole. This diagnostic has been designed to collect two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically the first image measures the distribution of the 14 MeV neutrons and the second image of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core.

The single-collision thermalization approximation for application to cold neutron moderation problems

International Nuclear Information System (INIS)

Ritenour, R.L.

1989-01-01

The single collision thermalization (SCT) approximation models the thermalization process by assuming that neutrons attain a thermalized distribution with only a single collision within the moderating material, independent of the neutron's incident energy. The physical intuition on which this approximation is based is that the salient properties of neutron thermalization are accounted for in the first collision, and the effects of subsequent collisions tend to average out statistically. The independence of the neutron incident and outscattering energy leads to variable separability in the scattering kernel and, thus, significant simplification of the neutron thermalization problem. The approximation also addresses detailed balance and neutron conservation concerns. All of the tests performed on the SCT approximation yielded excellent results. The significance of the SCT approximation is that it greatly simplifies thermalization calculations for CNS design. Preliminary investigations with cases involving strong absorbers also indicates that this approximation may have broader applicability, as in the upgrading of the thermalization codes

Intercomparison of personnel dosimetry for thermal neutron dose equivalent in neutron and gamma-ray mixed fields

International Nuclear Information System (INIS)

Ogawa, Yoshihiro

1985-01-01

In order to consider the problems concerned with personnel dosimetry using film badges and TLDs, an intercomparison of personnel dosimetry, especially dose equivalent responses of personnel dosimeters to thermal neutron, was carried out in five different neutron and gamma-ray mixed fields at KUR and UTR-KINKI from the practical point of view. For the estimation of thermal neutron dose equivalent, it may be concluded that each personnel dosimeter has good performances in the precision, that is, the standard deviations in the measured values by individual dosimeter were within 24 %, and the dose equivalent responses to thermal neutron were almost independent on cadmium ratio and gamma-ray contamination. However, the relative thermal neutron dose equivalent of individual dosimeter normalized to the ICRP recommended value varied considerably and a difference of about 4 times was observed among the dosimeters. From the results obtained, it is suggested that the standardization of calibration factors and procedures is required from the practical point of radiation protection and safety. (author)

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

International Nuclear Information System (INIS)

Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

2007-01-01

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

Applications of thermal neutron scattering in biology, biochemistry and biophysics

International Nuclear Information System (INIS)

Worcester, D.L.

1977-01-01

Biological applications of thermal neutron scattering have increased rapidly in recent years. The following categories of biological research with thermal neutron scattering are presently identified: crystallography of biological molecules; neutron small-angle scattering of biological molecules in solution (these studies have already included numerous measurements of proteins, lippoproteins, viruses, ribosomal subunits and chromatin subunit particles); neutron small-angle diffraction and scattering from biological membranes and membrane components; and neutron quasielastic and inelastic scattering studies of the dynamic properties of biological molecules and materials. (author)

Neutron radiography of thick hydrogenous materials with use of an imaging plate neutron detector

International Nuclear Information System (INIS)

Kato, K.; Matsumoto, G.; Karasawa, Y.; Niimura, N.; Matsubayashi, M.; Tsuruno, A.

1996-01-01

The value of the neutron mass attenuation coefficient of hydrogen being very high, it is extremely difficult to image normal size, living animals with neutron radiography. However, the authors suggest the possibility of applying neutron radiography for biomedical specimens. The organs in the breast, bones and cartilages in the extremities, and the tail of mice and rats were clearly imaged by neutron radiography with Gd foils as neutron converters and X-ray films. However, no contours of the organs in the mouse abdomen were visible with neutron radiography with an exposure time of 200 s. By adding Gd or Li compounds as neutron converters to imaging X-ray plates, imaging plates have been developed for neutron detectors. A trial using these imaging plates for neutron radiography of water-filled containers and the abdomen of mice was completed. The roundness of a 100 ml-beaker was imaged with a neutron exposure of 180 s. Obscure contours of the liver and kidneys of the mouse were imaged with a neutron exposure of 100 s. (orig.)

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

2015-03-15

Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

On the use of silicon as thermal neutron filter

International Nuclear Information System (INIS)

Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

2003-01-01

A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 μeV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

On the use of silicon as thermal neutron filter

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Habib, N.; Ashry, A.; Fathalla, M. E-mail: mohamedfathalla@hotmail.com

2003-12-01

A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 {mu}eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given.

A Detector for 2-D Neutron Imaging for the Spallation Neutron Source

International Nuclear Information System (INIS)

Britton, Charles L. Jr.; Bryan, W.L.; Wintenberg, Alan Lee; Clonts, Lloyd G.; Warmack, Robert J. Bruce; McKnight, Timothy E.; Frank, Steven Shane; Cooper, Ronald G.; Dudney, Nancy J.; Veith, Gabriel M.

2006-01-01

We have designed, built, and tested a 2-D pixellated thermal neutron detector. The detector is modeled after the MicroMegas-type structure previously published for collider-type experiments. The detector consists of a 4X4 square array of 1 cm 2 pixels each of which is connected to an individual preamplifier-shaper-data acquisition system. The neutron converter is a 10B film on an aluminum substrate. We describe the construction of the detector and the test results utilizing 252Cf sources in Lucite to thermalize the neutrons. Drift electrode (Aluminum) Converter (10B) 3 mm Conversion gap neutron (-900 V)

Boron neutron capture therapy (BNCT). Recent aspect, a change from thermal neutron to epithermal neutron beam and a new protocol

International Nuclear Information System (INIS)

Nakagawa, Yoshinobu

1999-01-01

Since 1968, One-hundred seventy three patients with glioblastoma (n=81), anaplastic astrocytoma (n=44), low grade astrocytoma (n=16) or other types of tumor (n=32) were treated by boron-neutron capture therapy (BNCT) using a combination of thermal neutron and BSH in 5 reactors (HTR n=13, JRR-3 n=1, MuITR n=98, KUR n=28, JRR-2 n=33). Out of 101 patients with glioma treated by BNCT under the recent protocol, 33 (10 glioblastoma, 14 anaplastic astrocytoma, 9 low grade astrocytoma) patients lived or have lived longer than 3 years. Nine of these 33 lived or have lived longer than 10 years. According to the retrospective analysis, the important factors related to the clinical results were tumor dose radiation dose and maximum radiation dose in thermal brain cortex. The result was not satisfied as it was expected. Then, we decided to introduce mixed beams which contain thermal neutron and epithermal neutron beams. KUR was reconstructed in 1996 and developed to be available to use mixed beams. Following the shutdown of the JRR-2, JRR-4 was renewed for medical use in 1998. Both reactors have capacity to yield thermal neutron beam, epithermal neutron beam and mixed beams. The development of the neutron source lead us to make a new protocol. (author)

Analysis of the Photoneutron Yield and Thermal Neutron Flux in an Unreflected Electron Accelerator-Driven Neutron Source

International Nuclear Information System (INIS)

Dale, Gregory E.; Gahl, John M.

2005-01-01

There are several potential uses for a high-flux thermal neutron source in both industrial and clinical applications. The viable commercial implementation of these applications requires a low-cost, high-flux thermal neutron generator suitable for installation in industrial and clinical environments. This paper describes the Monte Carlo for N-Particle modeling results of a high-flux thermal neutron source driven with an electron accelerator. An electron linear accelerator (linac), fitted with a standard X-ray converter, can produce high neutron yields in materials with low photonuclear threshold energies, such as D and 9 Be. Results indicate that a 10-MeV, 10-kW electron linac can produce on the order of 10 12 n/s in a heavy water photoneutron target. The thermal neutron flux in an unreflected heavy water target is calculated to be on the order of 10 10 n.cm -2 .s. The sensitivity of these answers to heavy water purity is also investigated, specifically the dilution of heavy water with light water. It is shown that the peak thermal neutron flux is not adversely effected by dilution up to a light water weight fraction of 35%

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1990-03-01

The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing, These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale of neutron shield from the cask. The test article was heated in an environmental prescribed by NRC regulations. Results of this second testing phase showed that all three materials are thermally acceptable

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.N.

1990-01-01

The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing. These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale section of neutron shield from the cask. The test article was heated in an environment prescribed by NRC regulations. Results of this second testing phase show that all three materials are thermally acceptable

Bibliography for thermal neutron scattering

International Nuclear Information System (INIS)

Sakamoto, M.; Chihara, J.; Nakahara, Y.; Kadotani, H.; Sekiya, T.

1976-12-01

It contains bibliographical references to measurements, calculations, reviews and basic studies on thermal neutron scatterings and dynamical properties of condensed matter. About 2,700 documents up to the end of 1975 are covered. (auth.)

Magneto–Thermal Evolution of Neutron Stars with Emphasis to ...

Indian Academy of Sciences (India)

The magnetic and thermal evolution of neutron stars is a very complex process with many non-linear interactions. For a decent understanding of neutron star physics, these evolutions cannot be considered isolated. A brief overview is presented, which describes the main magneto–thermal interactions that determine the fate ...

Study and development of new dosemeters for thermal neutrons; Estudio y desarrollo de nuevos dosimetros para neutrones termicos

Energy Technology Data Exchange (ETDEWEB)

Urena N, F

1998-12-31

An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation ({phi} = 5 x 10{sup 7} n/cm{sup 2} -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence {Phi} = {phi} {Delta}t ; where {Delta}t = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

Neutronics methods for thermal radiative transfer

International Nuclear Information System (INIS)

Larsen, E.W.

1988-01-01

The equations of thermal radiative transfer are time discretized in a semi-implicit manner, yielding a linear transport problem for each time step. The governing equation in this problem has the form of a neutron transport equation with fission but no scattering. Numerical methods are described, whose origins lie in neutron transport, and that have been successfully adapted to this new problem. Acceleration methods that have been developed specifically for the radiative transfer problem, but may have generalizations applicable in neutronics problems, are also discussed

Thickness optimization of various moderator materials for maximization of thermal neutron fluence

International Nuclear Information System (INIS)

Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

2015-01-01

Plasma focus device is widely being used as pulsed neutron source for variety of applications. Measurements of neutron yield by largely preferred Helium-3 proportional counter and Silver activation counter are mainly sensitive to thermal neutrons and are typically used with a neutron moderator. Thermalization of neutron is based on scattering reaction and hydrogenous materials are the best thermalizing medium. The efficiency of aforementioned neutron detectors is considerably affected by physical and geometrical properties of thermalizing medium i.e. moderator material, its thickness and shape. In view of the same, simulations have been performed to explore the effective utilization of Polyethylene, Perspex and Light water as moderating mediums for cylindrical and spherical geometry. In this study, estimated thermal fluence value up to 0.5 eV has been considered as the benchmark factor for comparing efficient thermalization by specific material, its thickness and shape. In either of the shapes being cylindrical or spherical, use of Polyethylene as moderating medium has resulted in minimum optimum thickness along with highest thermal fluence. (author)

The measurements of thermal neutron flux distribution in a paraffin

Indian Academy of Sciences (India)

The term `thermal flux' implies a Maxwellian distribution of velocity and energy corresponding to the most probable velocity of 2200 ms-1 at 293.4 K. In order to measure the thermal neutron flux density, the foil activation method was used. Thermal neutron flux determination in paraffin phantom by counting the emitted rays of ...

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

International Nuclear Information System (INIS)

Tzika, F.; Stamatelatos, I.E.

2004-01-01

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

Determination of the thermal and epithermal neutron sensitivities of an LBO chamber

Energy Technology Data Exchange (ETDEWEB)

Endo, Satoru; Kotani, Kei; Kajimoto, Tsuyoshi; Tanaka, Kenichi [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Sato, Hitoshi; Nakajima, Erika [Ibaraki Prefectural University of Health Science, Radiological Sciences, Ibaraki (Japan); Shimazaki, Takuto [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Delta Kogyo Co., Ltd., Hiroshima (Japan); Suda, Mitsuru; Hamano, Tsuyoshi [National Institute of Radiological Sciences, Chiba-Shi, Chiba (Japan); Hoshi, Masaharu [Hiroshima University, Institute for Peace Science, Hiroshima (Japan)

2017-08-15

An LBO (Li{sub 2}B{sub 4}O{sub 7}) walled ionization chamber was designed to monitor the epithermal neutron fluence in boron neutron capture therapy clinical irradiation. The thermal and epithermal neutron sensitivities of the device were evaluated using accelerator neutrons from the {sup 9}Be(d, n) reaction at a deuteron energy of 4 MeV (4 MeV d-Be neutrons). The response of the chamber in terms of the electric charge induced in the LBO chamber was compared with the thermal and epithermal neutron fluences measured using the gold-foil activation method. The thermal and epithermal neutron sensitivities obtained were expressed in units of pC cm{sup 2}, i.e., from the chamber response divided by neutron fluence (cm{sup -2}). The measured LBO chamber sensitivities were 2.23 x 10{sup -7} ± 0.34 x 10{sup -7} (pC cm{sup 2}) for thermal neutrons and 2.00 x 10{sup -5} ± 0.12 x 10{sup -5} (pC cm{sup 2}) for epithermal neutrons. This shows that the LBO chamber is sufficiently sensitive to epithermal neutrons to be useful for epithermal neutron monitoring in BNCT irradiation. (orig.)

Cross-section of single-crystal materials used as thermal neutron filters

International Nuclear Information System (INIS)

Adib, M.

2005-01-01

Transmission properties of several single crystal materials important for neutron scattering instrumentation are presented. A computer codes are developed which permit the calculation of thermal diffuse and Bragg-scattering cross-sections of silicon., and sapphire as a function of material's constants, temperature and neutron energy, E, in the range 0.1 MeV .A discussion of the use of their single-crystal as a thermal neutron filter in terms of the optimum crystal thickness, mosaic spread, temperature, cutting plane and tuning for efficient transmission of thermal-reactor neutrons is given

Flux distribution in phantom for biomedical use of beam-type thermal neutrons

International Nuclear Information System (INIS)

Aoki, Kazuhiko; Kobayashi, Tooru; Kanda, Keiji; Kimura, Itsuro

1985-01-01

For boron neutron capture therapy, the thermal neutron beam is worth using as therapeutic neutron irradiation without useless and unfavorable exposure of normal tissues around tumor and for microanalysis system to measure ppm-order 10 B concentrations in tissue and to search for the location of the metastasis of tumor. In the present study, the thermal neutron flux distribution in a phantom, when beam-type thermal neutrons were incident on it, was measured at the KUR Neutron Guide Tube. The measurements were carried out by two different methods using indium foil. The one is an ordinary foil activation technique by using the 115 In(n, γ) 116m 1 In reactions, while the other is to detect γ-rays from the 115 In(n, γ) 116m 2 In reactions during neutron irradiations with a handy-type Ge detector. The calculations with DOT 3.5 were performed to examine thermal neutron flux in the phantom for various beam size and phantom size. The experimental and calculated results are in good agreement and it is shown that the second type measurement has a potential for practical application as a new monitoring system of the thermal neutron flux in a living body for boron neutron capture therapy. (author)

Characteristics of thermal neutron calibration fields using a graphite pile

International Nuclear Information System (INIS)

Uchita, Yoshiaki; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio

2005-03-01

The Facility of Radiation Standards of Japan Atomic Energy Research Institute is equipped with thermal neutron fields for calibrating area and personal neutron dosemeters. The fields use moderated neutrons leaked from a graphite pile in which radionuclide sources are placed. In January 2003, we have renewed the pile with some modifications in its size. In accordance with the renewal, we measured and calculated thermal neutron fluence rates, neutron energy distributions and angular distributions of the fields. The thermal neutron fluence rates of the ''inside-pile fields'' and the outside-pile fields'' were determined by the gold foil activation method. The neutron energy distributions of the outside-pile fields were also measured with the Bonner multi-sphere spectrometer system. The contributions of epithermal and fast neutrons to the total dose-equivalents were 9% in the southern outside-pile field and 12% in the western outside-pile field. The personal dose-equivalents, H p,slab (10, α), in the outside-pile fields are evaluated by considering the calculated angular distributions of incoming neutrons. The H p,slab (10, α) was found to be about 40% higher than the value in assuming the unidirectional neutron between the pile and the test point. (author)

Measurement of thermal neutron capture cross section

International Nuclear Information System (INIS)

Huang Xiaolong; Han Xiaogang; Yu Weixiang; Lu Hanlin; Zhao Wenrong

2001-01-01

The thermal neutron capture cross sections of 71 Ga(n, γ) 72 Ga, 94 Zr(n, γ) 95 Zr and 191 Ir(n, γ) 192 Ir m1+g,m2 reactions were measured by using activation method and compared with other measured data. Meanwhile the half-life of 72 Ga was also measured. The samples were irradiated with the neutron in the thermal column of heavy water reactor of China Institute of Atomic Energy. The activities of the reaction products were measured by well-calibrated Ge(Li) detector

Thermal neutron detection by means of an organic solid-state track detector

International Nuclear Information System (INIS)

Doerschel, B.; Streubel, G.

1979-01-01

Thermal neutrons can be detected by means of organic solid-state track detectors if they are combined with radiators in which charged secondary particles are produced in neutron interaction processes. The secondary particles can produce etchable tracks in the detector material. For thermal neutron fluence determination from the track densities, the thermal neutron sensitivity was calculated for cellulose triacetate detectors with LiF radiators, taking into account energy and angular distribution of the alpha particles produced in the LiF radiator. This value is in good agreement with the sensitivity measured during irradiation in different neutron fields if corrections are considered the production of etchable or visuable tracks. Measuring range and measuring accuracy meet the requirements of thermal neutron detection in personnel dosimetry. Possibilities of extending the measuring range are discussed. (author)

Thermal neutron actinide data

International Nuclear Information System (INIS)

Tellier, H.

1992-01-01

During the 70's, the physicists involved in the cross section measurements for the low energy neutrons were almost exclusively interested in the resonance energy range. The thermal range was considered as sufficiently known. In the beginning of the 80's, reactor physicists had again to deal with the delicate problem of the power reactor temperature coefficient, essentially for the light water reactors. The measured value of the reactivity temperature coefficient does not agree with the computed one. The later is too negative. For obvious safety reasons, it is an important problem which must be solved. Several causes were suggested to explain this discrepancy. Among all these causes, the spectral shift in the thermal energy range seems to be very important. Sensibility calculations shown that this spectral shift is very sensitive to the shape of the neutron cross sections of the actinides for energies below one electron-volt. Consequently, reactor physicists require new and accurate measurements in the thermal and subthermal energy ranges. A part of these new measurement results were recently released and reviewed. The purpose of this study is to complete the preceding review with the new informations which are now available. In reactor physics the major actinides are the fertile nuclei, uranium 238, thorium 232 and plutonium 240 and the fissile nuclei, uranium 233, uranium 235 and plutonium 239. For the fertile nuclei the main datum is the capture cross section, for the fissile nuclei the data of interest are nu-bar, the fission and capture cross sections or a combination of these data such as η or α. In the following sections, we will review the neutron data of the major actinides for the energy below 1 eV

Cell death following thermal neutron exposure

Energy Technology Data Exchange (ETDEWEB)

Paterson, L.C. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Atanackovic, J. [Ontario Power Generation, Toronto, Ontario (Canada); Boyer, C. [Canadian Neutron Beam Centre, Chalk River, Ontario (Canada); El-Jaby, S.; Priest, N.D. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Seymour, C.B.; Boreham, D.R. [McMaster Univ., Hamilton, Ontario (Canada); Richardson, R.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2014-07-01

When individuals are exposed to unknown external ionizing radiation, it is desirable to have the means to assess both the absorbed dose received (Gy) and the radiation quality. Yet, conventional biodosimetry techniques, specifically the dicentric chromosome assay, cannot differentiate between the damage caused by high- and low-linear energy transfer (LET) exposures. Frequencies of apoptosis and necrosis, may provide an alternative method that assesses both the absorbed dose and radiation quality after unknown exposures. For this preliminary study, human lymphocytes were irradiated with {sup 60}Co gamma rays and thermal neutrons. Both apoptosis and necrosis increased with increasing gamma dose. In contrast, no dose-response was observed following thermal neutron exposure at doses up to 2.61 Gy. (author)

Thermal neutron capture cross sections of tellurium isotopes

International Nuclear Information System (INIS)

Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

2003-01-01

New values for thermal neutron capture cross sections of the tellurium isotopes 122 Te, 124 Te, 125 Te, 126 Te, 128 Te, and 130 Te are reported. These values are based on a combination of newly determined partial γ-ray cross sections obtained from experiments on targets contained natural Te and γ intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given

Thermal neutron capture cross sections of tellurium isotopes

International Nuclear Information System (INIS)

Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

2004-01-01

New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given

Thermal neutron capture cross sections of tellurium isotopes

Energy Technology Data Exchange (ETDEWEB)

Tomandl, I.; Honzatko, J.; von Egidy, T.; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

2004-03-01

New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given.

The National Ignition Facility Neutron Imaging System

International Nuclear Information System (INIS)

Wilke, Mark D.; Batha, Steven H.; Bradley, Paul A.; Day, Robert D.; Clark, David D.; Fatherley, Valerie E.; Finch, Joshua P.; Gallegos, Robert A.; Garcia, Felix P.; Grim, Gary P.; Jaramillo, Steven A.; Montoya, Andrew J.; Morgan, George L.; Oertel, John A.; Ortiz, Thomas A.; Payton, Jeremy R.; Pazuchanics, Peter; Schmidt, Derek W.; Valdez, Adelaida C.; Wilde, Carl H.

2008-01-01

The National Ignition Facility (NIF) is scheduled to begin deuterium-tritium (DT) shots possibly in the next several years. One of the important diagnostics in understanding capsule behavior and to guide changes in Hohlraum illumination, capsule design, and geometry will be neutron imaging of both the primary 14 MeV neutrons and the lower-energy downscattered neutrons in the 6-13 MeV range. The neutron imaging system (NIS) described here, which we are currently building for use on NIF, uses a precisely aligned set of apertures near the target to form the neutron images on a segmented scintillator. The images are recorded on a gated, intensified charge coupled device. Although the aperture set may be as close as 20 cm to the target, the imaging camera system will be located at a distance of 28 m from the target. At 28 m the camera system is outside the NIF building. Because of the distance and shielding, the imager will be able to obtain images with little background noise. The imager will be capable of imaging downscattered neutrons from failed capsules with yields Y n >10 14 neutrons. The shielding will also permit the NIS to function at neutron yields >10 18 , which is in contrast to most other diagnostics that may not work at high neutron yields. The following describes the current NIF NIS design and compares the predicted performance with the NIF specifications that must be satisfied to generate images that can be interpreted to understand results of a particular shot. The current design, including the aperture, scintillator, camera system, and reconstruction methods, is briefly described. System modeling of the existing Omega NIS and comparison with the Omega data that guided the NIF design based on our Omega results is described. We will show NIS model calculations of the expected NIF images based on component evaluations at Omega. We will also compare the calculated NIF input images with those unfolded from the NIS images generated from our NIS numerical

Neutron Imaging in Cultural Heritage Research at the FRM II Reactor of the Heinz Maier-Leibnitz Center

Directory of Open Access Journals (Sweden)

Burkhard Schillinger

2018-01-01

Full Text Available Neutron Imaging is ideally suited for applications in cultural heritage even at small reactors with moderate image resolution. However, recently, high resolution imaging is being increasingly used for advanced studies, especially in paleontology. The special contrast for hydrogen and between neighboring elements in the periodic system allows for new applications that are not accessible for X-rays, like organic material in enclosed containers made of ceramics or metals, fossilized bones in chalk rock or in ferrous “red” beds, and even for animal and hominid teeth. Fission neutrons permit the examination of large samples that otherwise show large attenuation for thermal neutrons.

Non-destructive assay employing 2D and 3D digital radiographic imaging acquired with thermal neutrons and reactor-produced radioisotopes

International Nuclear Information System (INIS)

Silvani, Maria Ines; Almeida, Gevaldo Lisboa de; Lopes, Ricardo T.

2011-01-01

The inner structure of some objects can only be visualized by using suitable techniques, when safety reasons or expensive costs preclude the application of invasive procedures. The kind of agent rendering an object partially transparent, unveiling thus its features, depends upon the object size and composition. As a rough rule of thumb, light materials are transparent to gamma and X-rays while the heavy ones are transparent to neutrons. When, after traversing an object, they hit a proper 2-D detector, a radiograph is produced representing a convoluted cross section, called projection, of that object. Taking a large number of such projections for different object attitudes, it is possible to obtain a 3-D tomography of the object as a map of attenuation coefficients. This procedure however, besides a time-consuming task, requires specially tailored equipment and software, not always available or affordable. Yet, in some circumstances it is feasible to replace the 3-D tomography by a stereoscopy, allowing one to visualize the spatial configuration of the object under analysis. In this work, 2-D and 3-D radiographic images have been acquired using thermal neutrons and reactor-produced radioisotopes and proper imaging plates as detectors. The stereographic vision has been achieved by taking two radiographs of the same object at different angles, from the detector point of view. After a treatment to render them red-white and green-white they were properly merged to yield a single image capable to be watched with red-green glasses. All the image treatment and rendering has been performed with the software ImageJ. (author)

Enhancement of thermal neutron attenuation of nano-B4C, -BN dispersed neutron shielding polymer nanocomposites

International Nuclear Information System (INIS)

Kim, Jaewoo; Lee, Byung-Chul; Uhm, Young Rang; Miller, William H.

2014-01-01

Highlights: • Preparation of B 4 C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B 4 C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B 4 C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1993-01-01

In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-A19897, R.H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280degF. Table 1 lists the neutron shield materials tested. The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found. The Bisco modified NS-4 and Reactor Experiments HMPP are both acceptable materials from a thermal accident standpoint for use in the shipping cask. Tests of the Kobe PP-R01 and Envirotech HDPE were stopped for safety reasons, due to inability to deal with the heavy smoke, before completion of the 30-minute heating phase. However these materials may prove satisfactory if they could undergo the complete heating. (J.P.N.)

Neutron thermalization in light water

International Nuclear Information System (INIS)

Abbate, M.J.; Lolich, J.V.

1975-05-01

Investigations related to neutron thermalization in light water have been made. Neutron spectra under quasi-infinite-medium conditions have been measured by the time-of-flight technique and calculations were performed with different codes. Through the use of improved experimental techniques and the best known calculational techniques available, the known discrepancies between experimentals and theoretical values were below from 40% to 16%. The present disagreement is believed to be due the scattering model used (ENDF-GASKET, based on the modified Haywood II frequency spectra), that shows to be very satisfactory for poisoned light water cases. Moreover, previous experiments were completed and differential, integral and pulse-source experimental techniques were improved. Also a second step of a neutron and reactor calculation system was completed. (author)

Introduction to the theory of thermal neutron scattering

CERN Document Server

Squires, G L

2012-01-01

Since the advent of the nuclear reactor, thermal neutron scattering has proved a valuable tool for studying many properties of solids and liquids, and research workers are active in the field at reactor centres and universities throughout the world. This classic text provides the basic quantum theory of thermal neutron scattering and applies the concepts to scattering by crystals, liquids and magnetic systems. Other topics discussed are the relation of the scattering to correlation functions in the scattering system, the dynamical theory of scattering and polarisation analysis. No previous knowledge of the theory of thermal neutron scattering is assumed, but basic knowledge of quantum mechanics and solid state physics is required. The book is intended for experimenters rather than theoreticians, and the discussion is kept as informal as possible. A number of examples, with worked solutions, are included as an aid to the understanding of the text.

System and plastic scintillator for discrimination of thermal neutron, fast neutron, and gamma radiation

Science.gov (United States)

Zaitseva, Natalia P.; Carman, M. Leslie; Faust, Michelle A.; Glenn, Andrew M.; Martinez, H. Paul; Pawelczak, Iwona A.; Payne, Stephen A.

2017-05-16

A scintillator material according to one embodiment includes a polymer matrix; a primary dye in the polymer matrix, the primary dye being a fluorescent dye, the primary dye being present in an amount of 3 wt % or more; and at least one component in the polymer matrix, the component being selected from a group consisting of B, Li, Gd, a B-containing compound, a Li-containing compound and a Gd-containing compound, wherein the scintillator material exhibits an optical response signature for thermal neutrons that is different than an optical response signature for fast neutrons and gamma rays. A system according to one embodiment includes a scintillator material as disclosed herein and a photodetector for detecting the response of the material to fast neutron, thermal neutron and gamma ray irradiation.

Determination of cadmium in zinc ores by thermal neutron absorption analysis

International Nuclear Information System (INIS)

De Norre, L.; Op de Beeck, J.; Hoste, J.

1983-01-01

A method has been developed for routine determination of cadmium in zinc ores by thermal neutron absorption analysis, based on the attenuation of a thermal neutron flux passing through a neutron absorbing material. The thermal neutron flux in related to the 52 V activity induced in a vanadium detector, surrounded by pellets pressed from a mixture of powdered material with graphite. Besides cadmium, also the major constituents zinc, iron and sulfur contribute significantly to the total attenuation of the thermal neutron flux. Calibration lines for these elements are worked out. All irradiations are carried out for 200 s in the partially thermalized neutron flux of a 5 Ci 227 Ac-Be isotope neutron source. After a decay of 30 s, the 52 V activity of the vanadium detector is measured for 400 s with a NaI(Tl) scintillation detector. The analysis sequence, including the computation of the results from the counting data, is automated by means of a LSI-11 Microprocessor with 12Kx16 bit memory. Zinc ores, containing 0.02 to 1.45% cadmium, have been analyzed with a precision ranging from 12.6% to 0.54%, resp. As a test for the reliability of the method, two NBS standard reference materials were analyzed in the same way as the zinc ore samples. (author)

Thermal neutron calibration channel at LNMRI/IRD

International Nuclear Information System (INIS)

Astuto, A.; Salgado, A.P.; Lopes, R.T.; Leite, S.P.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

2014-01-01

The Brazilian Metrology Laboratory of Ionizing Radiations (LNMRI) standard thermal neutron flux facility was designed to provide uniform neutron fluence for calibration of small neutron detectors and individual dosemeters. This fluence is obtained by neutron moderation from four 241 Am-Be sources, each with 596 GBq, in a facility built with blocks of graphite/paraffin compound and high-purity carbon graphite. This study was carried out in two steps. In the first step, simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The quality of the resulting neutron fluence in terms of spectrum, cadmium ratio and gamma-neutron ratio was evaluated. In the second step, the system was assembled based on the results obtained on the simulations, and new measurements are being made. These measurements will validate the system, and other intercomparisons will ensure traceability to the International System of Units. The pile construction form using blocks allows distinct arrangements for new studies and possibilities of other LNMRI reference fields. The results can be predicted by the simulation used in this work. Different number of each type of blocks and sources can be used. The main difference observed between the final measurement and simulation results might be due to the difference in composition of paraffin blocks used in assembling the pile. In order to confirm the thermal neutron field and fluence rate in the central chamber (inside the channel) that will be used to irradiate small neutron detectors, it is necessary to use another quantification method such as the gold foils activation with measurement traceability. It will be performed in a future stage. (authors)

Filtered thermal neutron captured cross sections measurements and decay heat calculations

International Nuclear Information System (INIS)

Pham Ngoc Son; Vuong Huu Tan

2015-01-01

Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R ed ) of 420 and neutron flux (Φ th ) of 1.6*10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross sections for nuclide of 51 V, by the activation method relative to the standard reaction 197 Au(n,γ) 198 Au. In addition to the activities of neutron capture cross sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U are introduced in this report. (author)

Scintillator Based Coded-Aperture Imaging for Neutron Detection

International Nuclear Information System (INIS)

Hayes, Sean-C.; Gamage, Kelum-A-A.

2013-06-01

In this paper we are going to assess the variations of neutron images using a series of Monte Carlo simulations. We are going to study neutron images of the same neutron source with different source locations, using a scintillator based coded-aperture system. The Monte Carlo simulations have been conducted making use of the EJ-426 neutron scintillator detector. This type of detector has a low sensitivity to gamma rays and is therefore of particular use in a system with a source that emits a mixed radiation field. From the use of different source locations, several neutron images have been produced, compared both qualitatively and quantitatively for each case. This allows conclusions to be drawn on how suited the scintillator based coded-aperture neutron imaging system is to detecting various neutron source locations. This type of neutron imaging system can be easily used to identify and locate nuclear materials precisely. (authors)

Neutron Dark-Field Imaging

Science.gov (United States)

Mullins, David

2017-09-01

Neutron imaging is typically used to image and reconstruct objects that are difficult to image using X-Ray imaging techniques. X-Ray absorption is primarily determined by the electron density of the material. This makes it difficult to image objects within materials that have high densities such as metal. However, the neutron scattering cross section primarily depends on the strong nuclear force, which varies somewhat randomly across the periodic table. In this project, an imaging technique known as dark field imaging using a far-field interferometer has been used to study a sample of granite. With this technique, interferometric phase images are generated. The dispersion of the microstructure of the sample dephases the beam, reducing the visibility. Collecting tomographic projections at different autocorrelation lengths (from 100 nanometers to 1.74 micrometers) essentially creates a 3D small angle scattering pattern, enabling mapping of how the microstructure is distributed throughout the sample.

Need for nuclear data for thermal neutron reactors

International Nuclear Information System (INIS)

Bouchard, J.; Golinelli, C.; Tellier, H.

1983-01-01

The need for nuclear data for thermal neutron reactors is conditioned by the persisting lack of agreement between the calculation and measurement of certain parameters, by the benefit that can be drawn from reduction of the marginal areas and by envisaged modifications. Three particular fields are delineated. Reduction of the deviation in temperature coefficients by modification of the shape of the effective capture cross sections of uranium-238 and -235 in the thermal range. The increase in precision of kinetic measurements by a better knowledge of data connected to slowed-down neutrons. Improvement in predicting the neutron activity of the fuels used in measuring the effective capture cross sections of plutonium-242 and americium-243. (Auth.)

Measurement of thermal neutron spectra using LINAC in Japan Atomic Energy Research Institute (JAERI)

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-01-01

The exact grasp of thermal neutron spectra in a core region is very important for obtaining accurate thermal neutron group constants in the calculation for the nuclear design of a reactor core. For the accurate grasp of thermal neutron spectra, the capability of thermal neutron spectra to describe the moderator cross-sections for thermal neutron scattering is a key factor. Accordingly, 0 deg angular thermal neutron spectra were measured by the time of flight (TOF) method using the JAERI LINAC as a pulsed neutron source, for light water system added with Cd and In, high temperature graphite system added with boron, and light water-natural uranium heterogeneous multiplication system among the reactor moderators of light water or graphite systems. First, the equations to give the time of flight and neutron flux by TOF method were analyzed, and several corrections were investigated, such as those for detector efficiency, background, the transmission coefficient of air and the Al window of a flight tube, mean emission time of neutrons, and the distortion effect of re-entrant hole on thermal neutron spectra. Then, the experimental system, results and calculation were reported for the experiments on the above three moderator systems. Finally, the measurement of fast neutron spectra in natural uranium system and that of the efficiency of a 6 Li glass scintillator detector are described. (Wakatsuki, Y.)

Comparison of Thermal Neutron Flux Measured by Uranium 235 Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

International Nuclear Information System (INIS)

Fourmentel, D.; Filliatre, P.; Barbot, L.; Villard, J.-F.; Lyoussi, A.; Geslot, B.; Malo, J.-Y.; Carcreff, H.; Reynard-Carette, C.

2013-06-01

Thermal neutron flux is one of the most important nuclear parameter to be measured on-line in Material Testing Reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with uranium 235 coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay - France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor. (authors)

VHTR core modeling: coupling between neutronic and thermal-hydraulics

International Nuclear Information System (INIS)

Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

2005-01-01

Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

Targets for bulk hydrogen analysis using thermal neutrons

CERN Document Server

Csikai, J; Buczko, C M

2002-01-01

The reflection property of substances can be characterized by the reflection cross-section of thermal neutrons, sigma subbeta. A combination of the targets with thin polyethylene foils allowed an estimation of the flux depression of thermal neutrons caused by a bulk sample containing highly absorbing elements or compounds. Some new and more accurate sigma subbeta values were determined by using the combined target arrangement. For the ratio, R of the reflection and the elastic scattering cross-sections of thermal neutrons, R=sigma subbeta/sigma sub E sub L a value of 0.60+-0.02 was found on the basis of the data obtained for a number of elements from H to Pb. Using this correlation factor, and the sigma sub E sub L values, the unknown sigma subbeta data can be deduced. The equivalent thicknesses, to polyethylene or hydrogen, of the different target materials were determined from the sigma subbeta values.

Results from the coded aperture neutron imaging system

International Nuclear Information System (INIS)

Brubaker, Erik; Steele, John T.; Brennan, James S.; Marleau, Peter

2010-01-01

Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging - a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.

The analysis and correction of neutron scattering effects in neutron imaging

International Nuclear Information System (INIS)

Raine, D.A.; Brenizer, J.S.

1997-01-01

A method of correcting for the scattering effects present in neutron radiographic and computed tomographic imaging has been developed. Prior work has shown that beam, object, and imaging system geometry factors, such as the L/D ratio and angular divergence, are the primary sources contributing to the degradation of neutron images. With objects smaller than 20--40 mm in width, a parallel beam approximation can be made where the effects from geometry are negligible. Factors which remain important in the image formation process are the pixel size of the imaging system, neutron scattering, the size of the object, the conversion material, and the beam energy spectrum. The Monte Carlo N-Particle transport code, version 4A (MCNP4A), was used to separate and evaluate the effect that each of these parameters has on neutron image data. The simulations were used to develop a correction algorithm which is easy to implement and requires no a priori knowledge of the object. The correction algorithm is based on the determination of the object scatter function (OSF) using available data outside the object to estimate the shape and magnitude of the OSF based on a Gaussian functional form. For objects smaller than 1 mm (0.04 in.) in width, the correction function can be well approximated by a constant function. Errors in the determination and correction of the MCNP simulated neutron scattering component were under 5% and larger errors were only noted in objects which were at the extreme high end of the range of object sizes simulated. The Monte Carlo data also indicated that scattering does not play a significant role in the blurring of neutron radiographic and tomographic images. The effect of neutron scattering on computed tomography is shown to be minimal at best, with the most serious effect resulting when the basic backprojection method is used

Characterization of the internal background for thermal and fast neutron detection with CLLB

Energy Technology Data Exchange (ETDEWEB)

Woolf, Richard S., E-mail: richard.woolf@nrl.navy.mil; Phlips, Bernard F.; Wulf, Eric A.

2016-12-01

We report on a set of experiments conducted to determine what effects, if any, the internal background in the CLLB scintillation detector has on the thermal neutron detection performance. We conducted source measurements using an unmoderated and moderated {sup 252}Cf neutron/γ-ray source and long (48-h), unshielded and shielded, background measurements to characterize the internal background with and without a source present. These measurements allowed us to determine the 2-d event selections needed to isolate the thermal neutron peak observed in pulse shape vs. energy space and apply those selections to our background measurements. Our results indicate that the thermal neutron detection capabilities of the CLLB are marginally affected by the presence of internal background. An unmoderated 113-µCi {sup 252}Cf source at 15 cm from the detector yields a thermal neutron rate of 8×10{sup −2}/s cm{sup 3}, while moderating the source with 5 cm of polyethylene yields a thermal neutron rate of 5.5×10{sup −1}/s cm{sup 3}. The measured background rate for events that fall within the selected thermal neutron region is 1.2×10{sup −3}/s cm{sup 3}. Lastly, the potential for CLLB for detecting fast neutrons was investigated.

Study of scattering in bi-dimensional neutron radiographic images

International Nuclear Information System (INIS)

Oliveira, K.A.M. de; Crispim, V.R.; Silva, F.C.

2009-01-01

The effect of neutron scattering frequently causes distortions in neutron radiographic images and, thus, reduces the quality. In this project, a type of filter, comprised of cadmium (a neutron absorber), was used in the form of a grid to correct this effect. This device generated image data in the discrete shadow bands of the absorber, components relative to neutron scattering on the test object and surroundings. Scattering image data processing, together with the original neutron radiographic image, resulted in a corrected image with improved edge delineation and, thus, greater definition in the neutron radiographic image of the test object. The objective of this study is to propose a theoretical/experimental methodology that is capable of eliminating the components relative to neutron scattering in neutron radiographic images, coming from the material that composes the test object and the materials that compose the surrounding area. (author)

Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

Energy Technology Data Exchange (ETDEWEB)

Fazal-ur-Rehman [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)]. E-mail: fazalr@kfupm.edu.sa; Al-Jarallah, M.I. [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Abu-Jarad, F. [Radiation Protection Unit, Environmental Protection Department, Saudi Aramco, P. O. Box 13027, Dhahran 31311 (Saudi Arabia); Qureshi, M.A. [Center for Applied Physical Sciences, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)

2005-11-15

A set of five {sup 241}Am-Be neutron sources are utilized in research and teaching at King Fahd University of Petroleum and Minerals (KFUPM). Three of these sources have an activity of 16Ci each and the other two are of 5Ci each. A well-shielded storage area was designed for these sources. The aim of the study is to check the effectiveness of shielding of the KFUPM neutron source storage area. Poly allyl diglycol carbonate (PADC) Nuclear track detectors (NTDs) based fast and thermal neutron area passive dosimeters have been utilized side by side for 33 days to assess accumulated low ambient dose equivalents of fast and thermal neutrons at 30 different locations around the source storage area and adjacent rooms. Fast neutron measurements have been carried out using bare NTDs, which register fast neutrons through recoils of protons, in the detector material. NTDs were mounted with lithium tetra borate (Li{sub 2}B{sub 4}O{sub 7}) converters on their surfaces for thermal neutron detection via B10(n,{alpha})Li6 and Li6(n,{alpha})H3 nuclear reactions. The calibration factors of NTD both for fast and thermal neutron area passive dosimeters were determined using thermoluminescent dosimeters (TLD) with and without a polyethylene moderator. The calibration factors for fast and thermal neutron area passive dosimeters were found to be 1.33 proton tracks cm{sup -2}{mu}Sv{sup -1} and 31.5 alpha tracks cm{sup -2}{mu}Sv{sup -1}, respectively. The results show variations of accumulated dose with the locations around the storage area. The fast neutron dose equivalents rates varied from as low as 182nSvh{sup -1} up to 10.4{mu}Svh{sup -1} whereas those for thermal neutron ranged from as low as 7nSvh{sup -1} up to 9.3{mu}Svh{sup -1}. The study indicates that the area passive neutron dosimeter was able to detect dose rates as low as 7 and 182nSvh{sup -1} from accumulated dose for thermal and fast neutrons, respectively, which were not possible to detect with the available active neutron

Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

International Nuclear Information System (INIS)

Vaiana, F.

2009-11-01

This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

Measurement and simulation of thermal neutron flux distribution in the RTP core

Science.gov (United States)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

Neutron imaging and applications a reference for the imaging community

CERN Document Server

McGreevy, Robert L; Bilheux, Hassina Z

2009-01-01

Offers an introduction to the basics of neutron beam production in addition to the wide scope of techniques that enhance imaging application capabilities. This title features a section that describes imaging single grains in polycrystalline materials, neutron imaging of geological materials and other materials science and engineering areas.

Fast and thermal neutron intensity measurements at the KFUPM PGNAA setup

CERN Document Server

Al-Jarallah, M I; Fazal-Ur-Rehman; Abu-Jarad, F A

2002-01-01

Fast and thermal neutron intensity distributions have been measured at an accelerator based prompt gamma ray neutron activation analysis (PGNAA) setup. The setup is built at the 350 keV accelerator laboratory of King Fahd University of Petroleum and Minerals (KFUPM). The setup is mainly designed to carry out PGNAA elemental analysis via thermal neutron capture. In this study relative intensity of fast and thermal neutrons was measured as a function of the PGNAA moderator assembly parameters using nuclear track detectors (NTDs). The relative intensity of the neutrons was measured inside the sample region as a function of front moderator thickness as well as sample length. Measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The pulsed deuteron beam with 5 ns pulse width and 30 kHz frequency was used to produce neutrons. Experimental results were compared with results of Monte Carlo design calculations of the PGNAA setup. A good agreement has bee...

Fast and thermal neutron intensity measurements at the KFUPM PGNAA setup

Energy Technology Data Exchange (ETDEWEB)

Al-Jarallah, M.I.; Naqvi, A.A. E-mail: aanaqvi@kfupm.edu.sa; Fazal-ur-Rehman; Abu-jarad, F

2002-10-01

Fast and thermal neutron intensity distributions have been measured at an accelerator based prompt gamma ray neutron activation analysis (PGNAA) setup. The setup is built at the 350 keV accelerator laboratory of King Fahd University of Petroleum and Minerals (KFUPM). The setup is mainly designed to carry out PGNAA elemental analysis via thermal neutron capture. In this study relative intensity of fast and thermal neutrons was measured as a function of the PGNAA moderator assembly parameters using nuclear track detectors (NTDs). The relative intensity of the neutrons was measured inside the sample region as a function of front moderator thickness as well as sample length. Measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The pulsed deuteron beam with 5 ns pulse width and 30 kHz frequency was used to produce neutrons. Experimental results were compared with results of Monte Carlo design calculations of the PGNAA setup. A good agreement has been found between the experimental results and the calculations.

Filtered thermal neutron captured cross-sections measurements and decay heat calculations

International Nuclear Information System (INIS)

Son, Pham Ngoc; Tan, Vuong Huu

2014-01-01

Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R cd ) of 420 and neutron flux (Φ th ) of 1.6x10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51 V, 55 Mn, 180 Hf and 186 W by the activation method relative to the standard reaction 197 Au(n,g) 198 Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U, 238 U, 239 Pu and 232 Th are introduced in this report. (author)

Enhancement of thermal neutron attenuation of nano-B{sub 4}C, -BN dispersed neutron shielding polymer nanocomposites

Energy Technology Data Exchange (ETDEWEB)

Kim, Jaewoo, E-mail: kimj@kaeri.re.kr [Nuclear Materials Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); WCI Quantum Beam based Radiation Research Center, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States); Lee, Byung-Chul [Nuclear Reactor Core Design Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Uhm, Young Rang [Radioisotopes Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Miller, William H. [Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States)

2014-10-15

Highlights: • Preparation of B{sub 4}C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B{sub 4}C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B{sub 4}C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption.

Fast-neutron, coded-aperture imager

International Nuclear Information System (INIS)

Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.

2015-01-01

This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led

Fast-neutron, coded-aperture imager

Energy Technology Data Exchange (ETDEWEB)

Woolf, Richard S., E-mail: richard.woolf@nrl.navy.mil; Phlips, Bernard F., E-mail: bernard.phlips@nrl.navy.mil; Hutcheson, Anthony L., E-mail: anthony.hutcheson@nrl.navy.mil; Wulf, Eric A., E-mail: eric.wulf@nrl.navy.mil

2015-06-01

This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led

Investigation of coupling scheme for neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

1988-01-01

Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

Neutron radiography with the cyclotron

International Nuclear Information System (INIS)

Tazawa, Shuichi; Asada, Yorihisa; Yano, Munehiko; Nakanii, Takehiko.

1985-01-01

Neutron radiography is well recognized as a powerful tool in nondestructive testing, but not widely used yet owing to lack of high intense thermal neutron source convenient for practical use. This article presents a new neutron radiograph facility, utilizing a sub-compact cyclotron as neutron source and is equipped with vertical and horizontal irradiation ports. The article describes a series of experiments, we conducted using beams of a variable energy cyclotron at Tohoku University to investigate the characteristics of thermal neutron obtained from 9 Be(p, n) reaction and thermalized by elastic scattering process. The article also describes a computer simulation of neutron moderator to analyze conditions getting maximal thermal neutron flux. Further, some of practical neutron radiograph examinations of aero-space components and museum art objects of classic bronze mirror and an attempt realizing real time imaging technique, are introduced in the article. (author)

Fundamentals and applications of neutron imaging. Application part 3. Application of neutron imaging in aircraft, space rocket, car and gunpowder industries

International Nuclear Information System (INIS)

Ikeda, Yasushi

2007-01-01

Neutron imaging is applied to nondestructive test. Four neutron imaging facilities are used in Japan. The application examples of industries are listed in the table: space rocket, aircraft, car, liquid metal, and works of art. Neutron imaging of transportation equipments are illustrated as an application to industry. X-ray radiography testing (XRT) image and neutron radiography testing (NRT) image of turbine blade of aircraft engine, honeycomb structure of aircraft, helicopter rotor blade, trigger tube, separation nut of space rocket, carburetor of car, BMW engine, fireworks and ammunitions are illustrated. (S.Y.)

Effect of Different Thermal Neutron Fluxes on Blood of Male Mice

International Nuclear Information System (INIS)

Abd El-Latif, A.A.; Saeid, Kh. S.; Abd El-Latif, A.A.; Emara, N.M.; Emara, N.M.

2010-01-01

This work deals with the exposing of male mice to different fluxes of thermal neutron .Investigation has been performed by calculating of thermal neutron fluxes(0.27x10 8 N/cm 2 . 1h , 0.54x10 8 N/cm 2 . 1h, 1.08x10 8 N/cm 2 . 1h, 2.16x10 8 N/cm 2 . 3h and 4.32x10 8 N/cm 2 . 6h) which emitted from neutron irradiation cell with source Ra - Be (α,n) have activity 3 m. Ci made by leybold(55930) . The number and differential leucocytes counts types of white blood cells in million per cubic millimeter (W. B. Cs. mm -3 ) ,the number of platelets mm -3 ,the number of red blood cells in million per cubic millimeter (R. B. Cs. mm -3 ), the hemoglobin in Blood (mg/dl), the lymphocytes ,and the eosiniphil leucocytes in blood decrease with increasing thermal neutron fluxes. But neutrophile and monocytes in blood increase with increasing the thermal neutron fluxes

Radiography using californium-252 neutron sources

International Nuclear Information System (INIS)

Ray, J.W.

1975-01-01

The current status in the technology of neutron radiography using californium-252 neutron sources is summarized. Major emphasis is on thermal neutron radiography since it has the widest potential applicability at the present time. Attention is given to four major factors which affect the quality and useability of thermal neutron radiography: source neutron thermalization, neutron beam extraction geometry, neutron collimator dimensions, and neutron imaging methods. Each of these factors has a major effect on the quality of the radiographs which are obtained from a californium source neutron radiography system and the exposure times required to obtain the radiographs; radiograph quality and exposure time in turn affect the practicality of neutron radiography for specific nondestructive inspection applications. A brief discussion of fast neutron radiography using californium-252 neutron sources is also included. (U.S.)

Fly's Eye: a counting camera for thermal neutrons, some applications, problems, and prospects

International Nuclear Information System (INIS)

Davidson, J.B.

1975-01-01

An area detector for thermal neutrons based on image intensification techniques is described and some capabilities and limitations of the detection system are discussed. Among the former are high spatial resolution high instantaneous counting rate, electronic zoom, time-gating, and integration. The detector is limited in that the maximum counting rate for a resolution element is 60 regularly spaced counts per second. Also, the nonuniformity of response over the detector limits the useful size and requires point-by-point calibration. In addition, a higher efficiency for neutron detection would be desirable. Some typical applications of the system are: crystal inspection, neutron magnetic diffraction topography, and searches for temperature induced changes in diffraction patterns. The future application of solid state television sensors and microchannel plate intensifiers to improve the system are briefly mentioned. (U.S.)

Fly's eye: a counting camera for thermal neutrons: some applications, problems, and prospects

International Nuclear Information System (INIS)

Davidson, J.B.

1976-01-01

An area detector for thermal neutrons based on image intensification techniques is described. Some capabilities and limitations of the detection system are discussed. Among the former are high spatial resolution, high instantaneous counting rate, electronic zoom, time-gating, and integration. The detector is limited in that the maximum counting rate for a resolution element is 60 regularly spaced counts per second. Also, the nonuniformity of response over the detector puts a limit on the useful size and necessitates point-by-point calibration. In addition, a higher efficiency for neutron detection would be desirable. Some typical applications of the system are crystal inspection, neutron magnetic diffraction topography, and searches for temperature-induced changes in diffraction patterns. The future application of solid-state television sensors and microchannel-plate intensifiers to improve the system is briefly mentioned

Results from neutron imaging of ICF experiments at NIF

Science.gov (United States)

Merrill, F. E.; Danly, C. R.; Fittinghoff, D. N.; Grim, G. P.; Guler, N.; Volegov, P. L.; Wilde, C. H.

2016-03-01

In 2011 a neutron imaging diagnostic was commissioned at the National Ignition Facility (NIF). This new system has been used to collect neutron images to measure the size and shape of the burning DT plasma and the surrounding fuel assembly. The imaging technique uses a pinhole neutron aperture placed between the neutron source and a neutron detector. The detection system measures the two-dimensional distribution of neutrons passing through the pinhole. This diagnostic collects two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically one image measures the distribution of the 14 MeV neutrons, and the other image measures the distribution of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core. Images have been collected for the majority of the experiments performed as part of the ignition campaign. Results from this data have been used to estimate a burn-averaged fuel assembly as well as providing performance metrics to gauge progress towards ignition. This data set and our interpretation are presented.

A time-of-flight detector for thermal neutrons from radiotherapy Linacs

Energy Technology Data Exchange (ETDEWEB)

Conti, V. [Universita degli Studi di Milano and INFN di Milano (Italy)], E-mail: conti.Valentina@gmail.com; Bartesaghi, G. [Universita degli Studi di Milano and INFN di Milano (Italy); Bolognini, D.; Mascagna, V.; Perboni, C.; Prest, M.; Scazzi, S. [Universita dell' Insubria, Como and INFN di Milano (Italy); Mozzanica, A. [Universita degli Studi di Brescia and INFN sezione di Pavia (Italy); Cappelletti, P.; Frigerio, M.; Gelosa, S.; Monti, A.; Ostinelli, A. [Fisica Sanitaria, Ospedale S. Anna di Como (Italy); Giannini, G.; Vallazza, E. [INFN, sezione di Trieste and Universita degli Studi di Trieste (Italy)

2007-10-21

Boron Neutron Capture Therapy (BNCT) is a therapeutic technique exploiting the release of dose inside the tumour cell after a fission of a {sup 10}B nucleus following the capture of a thermal neutron. BNCT could be the treatment for extended tumors (liver, stomach, lung), radio-resistant ones (melanoma) or tumours surrounded by vital organs (brain). The application of BNCT requires a high thermal neutron flux (>5x10{sup 8}ncm{sup -2}s{sup -1}) with the correct energy spectrum (neutron energy <10keV), two requirements that for the moment are fulfilled only by nuclear reactors. The INFN PhoNeS (Photo Neutron Source) project is trying to produce such a neutron beam with standard radiotherapy Linacs, maximizing with a dedicated photo-neutron converter the neutrons produced by Giant Dipole Resonance by a high energy (>8MeV) photon beam. In this framework, we have developed a real-time detector to measure the thermal neutron time-of -flight to compute the flux and the energy spectrum. Given the pulsed nature of Linac beams, the detector is a single neutron counting system made of a scintillator detecting the photon emitted after the neutron capture by the hydrogen nuclei. The scintillator signal is sampled by a dedicated FPGA clock thus obtaining the exact arrival time of the neutron itself. The paper will present the detector and its electronics, the feasibility measurements with a Varian Clinac 1800/2100CD and comparison with a Monte Carlo simulation.

The CLYC-6 and CLYC-7 response to γ-rays, fast and thermal neutrons

International Nuclear Information System (INIS)

Giaz, A.; Pellegri, L.; Camera, F.; Blasi, N.; Brambilla, S.; Ceruti, S.; Million, B.; Riboldi, S.; Cazzaniga, C.; Gorini, G.; Nocente, M.; Pietropaolo, A.; Pillon, M.; Rebai, M.; Tardocchi, M.

2016-01-01

The crystal Cs 2 LiYCl 6 :Ce (CLYC) is a very interesting scintillator material because of its good energy resolution and its capability to identify γ-rays and fast/thermal neutrons. The crystal Cs 2 LiYCl 6 :Ce contains 6 Li and 35 Cl isotopes, therefore, it is possible to detect thermal neutrons through the reaction 6 Li(n, α)t while 35 Cl ions allow to measure fast neutrons through the reactions 35 Cl(n, p) 35 S and 35 Cl(n, α) 32 P. In this work two CLYC 1″×1″ crystals were used: the first crystal, enriched with 6 Li at 95% (CLYC-6) is ideal for thermal neutron measurements while the second one, enriched with 7 Li at >99% (CLYC-7) is suitable for fast neutron measurements. The response of CLYC scintillators was measured with different PMT models: timing or spectroscopic, with borosilicate glass or quartz window. The energy resolution, the neutron-γ discrimination and the internal activity are discussed. The capability of CLYC scintillators to discriminate γ rays from neutrons was tested with both thermal and fast neutrons. The thermal neutrons were measured with both detectors, using an AmBe source. The measurements of fast neutrons were performed at the Frascati Neutron Generator facility (Italy) where a deuterium beam was accelerated on a deuterium or on a tritium target, providing neutrons of 2.5 MeV or 14.1 MeV, respectively. The different sensitivity to thermal and fast neutrons of a CLYC-6 and of a CLYC-7 was additionally studied.

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

Schultz, W.E.; Smith, H.D.; Smith, M.P.

1980-01-01

An improved method and apparatus are described for simultaneously measuring the porosity and thermal neutron capture cross section of earth formations in situ in the vicinity of a well borehole using pulsed neutron well logging techniques. The logging tool which is moved through the borehole consists of a 14 MeV pulsed neutron source, an epithermal neutron detector and a combination gamma ray and fast neutron detector. The associated gating systems, counters and combined digital computer are sited above ground. (U.K.)

CaSO{sub 4}:Dy microphosphor for thermal neutron dosimetry

Energy Technology Data Exchange (ETDEWEB)

Bhadane, Mahesh S. [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Mandlik, Nandkumar [Department of Physics, Fergusson College, Savitribai Phule Pune University, Pune 411007 (India); Patil, B.J. [Department of Physics, Abasaheb Garware College, Pune 411004 (India); Dahiwale, S.S.; Sature, K.R.; Bhoraskar, V.N. [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India)

2016-02-15

Dysprosium-doped calcium sulphate (CaSO{sub 4}:Dy) microphosphor was synthesized by acid re-crystallization method and its thermoluminescence (TL) properties irradiated with thermal neutrons was studied. Structural and morphological characteristics have been studied using X-ray diffraction and SEM which mainly exhibits a orthorhombic structure with particle size of 200 to 250 µm. Moreover, thermal neutron dosimetric characteristics of the microphosphor such as thermoluminescence glow curve, TL dose–response have been studied. This microphosphor powder represents a TL glow peak (T{sub max}) centered at around 240 °C. The TL response of CaSO{sub 4}:Dy microphosphor as a function of thermal neutron fluence is observed to be very linear upto the fluence of 52×10{sup 11} n/cm{sup 2} and further saturates. In addition, TL glow curves were deconvoluted by computerized glow curve deconvolution (CGCD) method and corresponding trapping parameters have been determined. It has been found that for every deconvoluted peak there is change in the order of kinetics. Overall, the experimental results show that the CaSO{sub 4}:Dy microphosphor can have potential to be an effective thermal neutron dosimetry. - Highlights: • Acid-recrystallization method is used to prepare CaSO{sub 4}:Dy microphosphor • CaSO{sub 4}:Dy phosphor irradiated thermal neutrons for dosimetric application. • TL response curve showed to be a perfect linear. • Trapping parameters has been calculated using CGCD curve fitting.

Calculation of neutron flux distribution of thermal neutrons from microtron converter in a graphite moderator with water reflector

International Nuclear Information System (INIS)

Andrejsek, K.

1977-01-01

The calculation is made of the thermal neutron flux in the moderator and reflector by solving the neutron diffusion equation using the four-group theory. The correction for neutron absorption in the moderator was carried out using the perturbation theory. The calculation was carried out for four groups with the following energy ranges: the first group 2 MeV to 3 keV, the second group 3 keV to 5 eV, the third group 5 eV to 0.025 eV and the fourth group 0.025 eV. The values of the macroscopic cross section of capture and scattering, of the diffusion coefficient, the macroscopic cross section of the moderator, of the neutron age and the extrapolation length for the water-graphite moderator used in the calculations are given. The spatial distribution of the thermal neutron flux is graphically represented for graphite of a 30, 40, and 50 cm radius and for graphite of a 30 and 40 cm radius with a 10 cm water reflector; a graphic comparison is made of the distribution of the thermal neutron flux in water and in graphite, both 40 cm in radius. The system of graphite with reflector proved to be the best and most efficient system for raising the flux density of thermal neutrons. (J.P.)

Physics of epi-thermal boron neutron capture therapy (epi-thermal BNCT).

Science.gov (United States)

Seki, Ryoichi; Wakisaka, Yushi; Morimoto, Nami; Takashina, Masaaki; Koizumi, Masahiko; Toki, Hiroshi; Fukuda, Mitsuhiro

2017-12-01

The physics of epi-thermal neutrons in the human body is discussed in the effort to clarify the nature of the unique radiologic properties of boron neutron capture therapy (BNCT). This discussion leads to the computational method of Monte Carlo simulation in BNCT. The method is discussed through two examples based on model phantoms. The physics is kept at an introductory level in the discussion in this tutorial review.

Effective source size as related to 252Cf neutron radiography

International Nuclear Information System (INIS)

Wada, Nobuo; Enomoto, Shigemasa; Tachikawa, Noboru; Nojiri, Toshiaki.

1977-01-01

The effective source size in 252 Cf thermal neutron radiography, relating to its geometrical unsharpness in image formation, is experimentally studied. A neutron radiographic system consists of a 160 μg 252 Cf neutron source, water moderator and divergent cadmium lined collimator. Thermal neutron image detection is performed with using a LiF scintillator and a high speed X-ray film to employ direct exposure method. The modulation transfer function, used for describing image quality, is derived from radiographic image corresponding to a cadmium plate with sharp edge. The modulation transfer function for the system is expressed by the product of the function for both geometrical and inherent unsharpness, and allows isolation of geometrical unsharpness as related to the effective size of the thermal neutron source. It is found to be 80 -- 90% of the collimator inlet diameter. (auth.)

Scattering of thermal neutron by the water molecule

International Nuclear Information System (INIS)

Rosa, L.P.

The calculation of the differenctial cross section for scattering of thermal neutrons by water, taking into account the translational, rotational and vibrational motions of the water molecule, is presented according to Nelkin' model. Some modifications are presented which have been introduced in the original method to improve the results and an application has been made to reactor physics, by calculating the thermal neutron flux in a homogenous medium containing water and absorver. Thirty thermal energy groups have been used to compute the spectra. Within the limits of error, better agreement has been obtained between theory and experiments by using a modified Nelkin kernel consisting of substituting the asymptotic formulae for the rotational and vibrational motions by more exact expressions, similar to the Buttler model for heavy water

About possibilities of obtaining focused beams of thermal neutrons of radionuclide source

International Nuclear Information System (INIS)

Aripov, G.A.; Kurbanov, B.I.; Sulaymanov, N.T.; Ergashev, A.

2004-01-01

Full text: In the last years significant progress is achieved in development of neutron focusing methods (concentrating neutrons in a given direction and a small area). In this, main attention is given to focusing of neutron beams of reactor, particularly cold neutrons and their applications. [1,2]. However, isotope sources also let obtain intensive neutron beams and solve quite important (tasks) problems (e.g. neutron capture therapy for malignant tumors) [3], and an actual problems is focusing of neutrons. We developed a device on the basis of californium source of neutrons, allowing to obtain focused (preliminarily) beam of thermal neutrons with the aid of respective choice of moderators, reflectors and geometry of their disposition. Here, fast neutrons and gamma rays in the beam are minimized. With the aid of the model we developed on the basis of Monte-Carlo method, it is possible to modify aforementioned device and dynamics of output neutrons in wide energy range and analyze ways of optimization of neutron beams of isotope sources with different neutron outputs. Device of preliminary focusing of thermal neutrons can serve as a basis for further focus of neutrons using micro- and nano-capillar systems. It is known that, capillary systems performed with certain technology can form beam of thermal neutrons increasing its density by more than two orders of magnitude and effectively divert beams up to 20 o with length of system 15 cm

About possibilities of obtaining focused beams of thermal neutrons of radionuclide source

International Nuclear Information System (INIS)

Aripov, G.A.; Kurbanov, B.I.; Sulaymanov, N.T.; Ergashev, A.

2004-01-01

In the last years significant progress is achieved in development of neutron focusing methods (concentrating neutrons in a given direction and a small area). In this, main attention is given to focusing of neutron beams of reactor, particularly cold neutrons and their applications. [1,2]. However, isotope sources also let obtain intensive neutron beams and solve quite important (tasks) problems (e.g. neutron capture therapy for malignant tumors) [3], and an actual problems is focusing of neutrons. We developed a device on the basis of californium source of neutrons, allowing to obtain focused (preliminarily) beam of thermal neutrons with the aid of respective choice of moderators, reflectors and geometry of their disposition. Here, fast neutrons and gamma rays in the beam are minimized. With the aid of the model we developed on the basis of Monte-Carlo method, it is possible to modify aforementioned device and dynamics of output neutrons in wide energy range and analyze ways of optimization of neutron beams of isotope sources with different neutron outputs. Device of preliminary focusing of thermal neutrons can serve as a basis for further focus of neutrons using micro- and nano-capillary systems. It is known that, capillary systems performed with certain technology can form beam of thermal neutrons increasing its density by more than two orders of magnitude and effectively divert beams up to 20 o with length of system 15 cm. (author)

Study of the Li2CO3 as thermal neutrons detector

International Nuclear Information System (INIS)

Herrera A, E.; Urena N, F.; Delfin L, A.

2003-01-01

The use every day but it frequents of the thermal neutrons in the treatment of tumours, using the neutron capture therapy technique in boron, there is generated the necessity to develop a dosimetric system that allows to evaluate in a reliable way the fluence and consequently the dose of neutrons that it is given in the tumours of the patients. One of the techniques but employees to determine the neutron fluence sub cadmic and epi cadmic in an indirect way, it is the activation of thin sheets of gold undress and covered with cadmium respectively that when being exposed to a neutron beam to the nuclear reaction 197 Au (n, γ ) 198 Au, emitting gamma radiation with an energy of 0.4118 MeV, being this, a disadvantage to be used as dosemeter. On the other hand, when exposing the lithium carbonate to a thermal neutron beam, free radicals of CO 3 that are quantified by the electron paramagnetic resonance technique are generated. This work analyzes those basic parameters that determine if those made up of Li 2 CO 3 complete with the requirements to be used as detectors and/or dosemeters of thermal neutrons. (Author)

Thermal neutrons streaming in straight duct

International Nuclear Information System (INIS)

Jehouani, A.; Boulkheir, M.; Ichaoui, R.

2000-01-01

The neutron streaming in duct is due to two phenomena: a) direct propagation and b) reflection on duct wall. We have used the Monte Carlo method to evaluate the ratio of the reflected neutrons flux by the duct wall to the total flux at the exit of the duct for iron and aluminium. Ten neutrons energy groups are considered between 10 -5 eV and 10 eV. A Fortran program is developed to evaluate the neutron double differential albedo. It is shown that the two following approximations are largely justified: i) Three collisions in the duct wall are sufficient to attain the asymptotic limit of the multiscattered neutron double differential albedo ii) The points of entry and exit of the neutron in the duct wall may be considered the same for the multiscattered neutrons. For a punctual source at the mouth of the duct, we have determined the direct and the reflected part of the total thermal neutron flux at the exit of the duct for different lengths and different radius of the duct. For a punctual source, we have found that the major contribution to the total flux of neutrons at the exit is due to the neutron reflection by walls and the reflection contribution decreases when the neutron energy decreases. For a constant length of the duct, the reflected part decreases when the duct radius increases while for the disk shaped source we have found the opposite phenomena. The transmitted neutron flux distribution at the exit of the duct are determined for disk shaped source for different neutron energy and for different distance from the exit center. (author)

Characterization of thermal neutron fields for calibration of neutron monitors in accordance with great equivalent dose environment H⁎(10)

International Nuclear Information System (INIS)

Silva, Larissa P. S. da; Silva, Felipe S.; Fonseca, Evaldo S.; Patrao, Karla C.S.; Pereira, Walsan W.

2017-01-01

The Laboratório Brasileiro de Nêutrons do Instituto de Radioproteção e Dosimetria (IRD/CNEN) has developed and built a thermal neutron flux facility to provide neutron fluence for dosimeters (Astuto, 2014). This fluency is obtained by four 16 Ci sources 241 AmBe (α, n) positioned around the channel positioned in the center of the Thermal Flow Unit (UFT). The UFT was built with blocks of paraffin with graphite addition and graphite blocks of high purity to obtain a central field with a homogeneous thermal neutron fluence for calibration purposes with the following measurements: 1.2 x 1.2 x 1.2 m 3 . The objective of this work is to characterize several points, in the thermal energy range, in terms of the equivalent ambient dose quantity H⁎(10) for calibration and irradiation of monitors neutrons

Thermal-hydraulic studies of the Advanced Neutron Source cold source

International Nuclear Information System (INIS)

Williams, P.T.; Lucas, A.T.

1995-08-01

The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite

International Nuclear Information System (INIS)

Zhou, T.; Hawari, A. I.; Wehring, B. W.

2007-01-01

Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured

Set of thermal neutron-scattering experiments for the Weapons Neutron Research Facility

International Nuclear Information System (INIS)

Brugger, R.M.

1975-12-01

Six classes of experiments form the base of a program of thermal neutron scattering at the Weapons Neutron Research (WNR) Facility. Three classes are to determine the average microscopic positions of atoms in materials and three are to determine the microscopic vibrations of these atoms. The first three classes concern (a) powder sample neutron diffraction, (b) small angle scattering, and (c) single crystal Laue diffraction. The second three concern (d) small kappa inelastic scattering, (e) scattering surface phonon measurements, and (f) line widths. An instrument to couple with the WNR pulsed source is briefly outlined for each experiment

Scintillating-fiber imaging detector for 14-MeV neutrons

International Nuclear Information System (INIS)

Ress, D.; Lerche, R.A.; Ellis, R.J.; Heaton, G.W.; Nelson, M.B.; Mant, G.; Lehr, D.E.

1994-01-01

The authors have created a detector to image the neutrons emitted by imploded inertial-confinement fusion targets. The 14-MeV neutrons, which are produced by deuterium-tritium fusion events in the target, pass through an aperture to create an image on the detector. The neutron radiation is converted to blue light (430 nm) with a 20-cm-square array of plastic scintillating fibers. Each fiber is 10-cm long with a 1-mm-square cross section; approximately 35-thousand fibers make up the array. The resulting blue-light image is reduced and amplified by a sequence of fiber-optic tapers and image intensifiers, then acquired by a CCD camera. The fiber-optic readout system was tested optically for overall throughput the resolution. The authors plan to characterize the scintillator array reusing an ion-beam neutron source as well as DT-fusion neutrons emitted by inertial confinement targets. Characterization experiments will measure the light-production efficiency, spatial resolution, and neutron scattering within the detector. Several neutron images of laser-fusion targets have been obtained with the detector. Several neutron images of laser-fusion targets have been obtained with the detector. They describe the detector and their characterization methods, present characterization results, and give examples of the neutron images

A novel dual mode neutron-gamma imager

International Nuclear Information System (INIS)

Cooper, Robert Lee; Gerling, Mark; Brennan, James S.; Mascarenhas, Nicholas; Mrowka, Stanley; Marleau, Peter

2010-01-01

The Neutron Scatter Camera (NSC) can image fission sources and determine their energy spectra at distances of tens of meters and through significant thicknesses of intervening materials in relatively short times (1). We recently completed a 32 element scatter camera and will present recent advances made with this instrument. A novel capability for the scatter camera is dual mode imaging. In normal neutron imaging mode we identify and image neutron events using pulse shape discrimination (PSD) and time of flight in liquid scintillator. Similarly gamma rays are identified from Compton scatter in the front and rear planes for our segmented detector. Rather than reject these events, we show it is possible to construct a gamma-ray image by running the analysis in a 'Compton mode'. Instead of calculating the scattering angle by the kinematics of elastic scatters as is appropriate for neutron events, it can be found by the kinematics of Compton scatters. Our scatter camera has not been optimized as a Compton gamma-ray imager but is found to work reasonably. We studied imaging performance using a Cs137 source. We find that we are able to image the gamma source with reasonable fidelity. We are able to determine gamma energy after some reasonable assumptions. We will detail the various algorithms we have developed for gamma image reconstruction. We will outline areas for improvement, include additional results and compare neutron and gamma mode imaging.

Modeling the National Ignition Facility neutron imaging system.

Science.gov (United States)

Wilson, D C; Grim, G P; Tregillis, I L; Wilke, M D; Patel, M V; Sepke, S M; Morgan, G L; Hatarik, R; Loomis, E N; Wilde, C H; Oertel, J A; Fatherley, V E; Clark, D D; Fittinghoff, D N; Bower, D E; Schmitt, M J; Marinak, M M; Munro, D H; Merrill, F E; Moran, M J; Wang, T-S F; Danly, C R; Hilko, R A; Batha, S H; Frank, M; Buckles, R

2010-10-01

Numerical modeling of the neutron imaging system for the National Ignition Facility (NIF), forward from calculated target neutron emission to a camera image, will guide both the reduction of data and the future development of the system. Located 28 m from target chamber center, the system can produce two images at different neutron energies by gating on neutron arrival time. The brighter image, using neutrons near 14 MeV, reflects the size and symmetry of the implosion "hot spot." A second image in scattered neutrons, 10-12 MeV, reflects the size and symmetry of colder, denser fuel, but with only ∼1%-7% of the neutrons. A misalignment of the pinhole assembly up to ±175 μm is covered by a set of 37 subapertures with different pointings. The model includes the variability of the pinhole point spread function across the field of view. Omega experiments provided absolute calibration, scintillator spatial broadening, and the level of residual light in the down-scattered image from the primary neutrons. Application of the model to light decay measurements of EJ399, BC422, BCF99-55, Xylene, DPAC-30, and Liquid A suggests that DPAC-30 and Liquid A would be preferred over the BCF99-55 scintillator chosen for the first NIF system, if they could be fabricated into detectors with sufficient resolution.

Design of thermal neutron beam based on an electron linear accelerator for BNCT.

Science.gov (United States)

Zolfaghari, Mona; Sedaghatizadeh, Mahmood

2016-12-01

An electron linear accelerator (Linac) can be used for boron neutron capture therapy (BNCT) by producing thermal neutron flux. In this study, we used a Varian 2300 C/D Linac and MCNPX.2.6.0 code to simulate an electron-photoneutron source for use in BNCT. In order to decelerate the produced fast neutrons from the photoneutron source, which optimize the thermal neutron flux, a beam-shaping assembly (BSA) was simulated. After simulations, a thermal neutron flux with sharp peak at the beam exit was obtained in the order of 3.09×10 8 n/cm 2 s and 6.19×10 8 n/cm 2 s for uranium and enriched uranium (10%) as electron-photoneutron sources respectively. Also, in-phantom dose analysis indicates that the simulated thermal neutron beam can be used for treatment of shallow skin melanoma in time of about 85.4 and 43.6min for uranium and enriched uranium (10%) respectively. Copyright © 2016. Published by Elsevier Ltd.

X-ray micro-CT and neutron CT as complementary imaging tools for non-destructive 3D imaging of rare silicified fossil plants

Science.gov (United States)

Karch, J.; Dudák, J.; Žemlička, J.; Vavřík, D.; Kumpová, I.; Kvaček, J.; Heřmanová, Z.; Šoltés, J.; Viererbl, L.; Morgano, M.; Kaestner, A.; Trtík, P.

2017-12-01

Computed tomography provides 3D information of inner structures of investigated objects. The obtained information is, however, strongly dependent on the used radiation type. It is known that as X-rays interact with electron cloud and neutrons with atomic nucleus, the obtained data often provide different contrast of sample structures. In this work we present a set of comparative radiographic and CT measurements of rare fossil plant samples using X-rays and thermal neutrons. The X-ray measurements were performed using large area photon counting detectors Timepix at IEAP CTU in Prague and Perkin Elmer flat-panel detector at Center of Excellence Telč. The neutron CT measurement was carried out at Paul Scherrer Institute using BOA beam-line. Furthermore, neutron radiography of fossil samples, provided by National Museum, were performed using a large-area Timepix detector with a neutron-sensitive converting 6LiF layer at Research Centre Rez, Czech Republic. The obtained results show different capabilities of both imaging approaches. While X-ray micro-CT provides very high resolution and enables visualization of fine cracks or small cavities in the samples neutron imaging provides high contrast of morphological structures of fossil plant samples, where X-ray imaging provides insufficient contrast.

Thermal neutron equivalent doses assessment around KFUPM neutron source storage area using NTDs

Energy Technology Data Exchange (ETDEWEB)

Abu-Jarad, F.; Fazal-ur-Rehman; Al-Haddad, M.N.; Al-Jarrallah, M.I.; Nassar, R

2002-07-01

Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li{sub 2}B{sub 4}O{sub 7}) layer for thermal neutron detection via {sup 10}B(N,{alpha}){sup 7}Li and {sup 6}Li(n,{alpha}){sup 3}H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks.cm{sup -2}.{mu}Sv{sup -1} using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSv.h{sup -1} up to 11 {mu}Sv.h{sup -1}. The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv.h{sup -1}, which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems. (author)

Bibliography for thermal neutron scattering

International Nuclear Information System (INIS)

Sakamoto, Masanobu; Chihara, Junzo; Gotoh, Yorio; Kadotani, Hiroyuki; Sekiya, Tamotsu.

1979-09-01

Bibliographic references are given for measurements, calculations, reviews and basic studies of thermal neutron scattering and dynamical properties of condensed matter. This is the sixth edition covering 3,326 articles collected up to 1978. The edition being the final issue of the present bibliography series, a forthcoming edition will be published in a new form of bibliography. (author)

Results from the Coded Aperture Neutron Imaging System (CANIS)

International Nuclear Information System (INIS)

Brubaker, Erik; Steele, John T.; Brennan, James S.; Hilton, Nathan R.; Marleau, Peter

2010-01-01

Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging- a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

International Nuclear Information System (INIS)

Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

2015-01-01

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

Energy Technology Data Exchange (ETDEWEB)

Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

2015-02-15

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

Development and investigation of a neutron radiography imaging system with a low-pressure multistep chamber

International Nuclear Information System (INIS)

Anisimov, Yu.S.; Chernenko, S.P.; Ivanov, A.B.; Netusil, T.; Peshekhonov, V.D.; Smykov, L.P.; Zanevsky, Yu.V.; Cisar, M.; Horacek, J.; Knourek, J.; Moucka, L.; Nezmar, L.; Pellar, L.; Pochman, J.; Schneider, Z.; Sidak, Z.; Vrba, I.; Bizek, V.; Zavadil, Z.; Beran, P.; Cerny, K.

1988-01-01

An imaging system of thermal neutrons for an investigation of digital neutron radiography has been developed and tested. Some characteristics obtained on a neutron radiography beam of an experimental reactor are reported. The coordinates of each event are determined in this system. After processing in a LSI 11/23 computer, a radiograph, accumulated in a histogramming memory of 64 K 16-bit words, is presented on a colour display. A 230x180 mm 2 low-pressure multistep chamber is used as a detector. Neutron conversion takes place in a 6 μm boron layer enriched to 86% in 10 B. The detection efficiency of thermal neutrons is no less than 3%. The count rate of the system reaches up to 2x10 5 events per second. A radiograph can be obtained within 10 minutes. The sensitivity of this system to gamma-background is low. One event/s is detected for a background of 1 R/h. The spatial resolution is found to be 0.7 mm (FWHM) using a cadmium knife edge. The integral nonlinearity is less than 0.4%. The possibility of using a hydrogeneous converter in this system for neutron radiography is discussed. (orig.)

Experimental characterization of HOTNES: A new thermal neutron facility with large homogeneity area

Energy Technology Data Exchange (ETDEWEB)

Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Sperduti, A. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pietropaolo, A.; Pillon, M. [ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN–Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

2017-01-21

A new thermal neutron irradiation facility, called HOTNES (HOmogeneous Thermal NEutron Source), was established in the framework of a collaboration between INFN-LNF and ENEA-Frascati. HOTNES is a polyethylene assembly, with about 70 cmx70 cm square section and 100 cm height, including a large, cylindrical cavity with diameter 30 cm and height 70 cm. The facility is supplied by a {sup 241}Am-B source located at the bottom of this cavity. The facility was designed in such a way that the iso-thermal-fluence surfaces, characterizing the irradiation volume, coincide with planes parallel to the cavity bottom. The thermal fluence rate across a given isofluence plane is as uniform as 1% on a disk with 30 cm diameter. Thermal fluence rate values from about 700 cm{sup −2} s{sup −1} to 1000 cm{sup −2} s{sup −1} can be achieved. The facility design, previously optimized by Monte Carlo simulation, was experimentally verified. The following techniques were used: gold activation foils to assess the thermal fluence rate, semiconductor-based active detector for mapping the irradiation volume, and Bonner Sphere Spectrometer to determine the complete neutron spectrum. HOTNES is expected to be attractive for the scientific community involved in neutron metrology, neutron dosimetry and neutron detector testing.

THERMAL: A routine designed to calculate neutron thermal scattering

International Nuclear Information System (INIS)

Cullen, D.E.

1995-01-01

THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy

Neutron radiography with the cyclotron, 3

International Nuclear Information System (INIS)

Hiraoka, Eiichi; Fujishiro, Masatoshi; Tsujii, Yukio

1985-01-01

Neutron radiography is well recognized as a powerful tool in nondestructive testing, but not widely used yet owing to lack of high intense thermal neutron source convenient for practical use. A new neutron radiograph facility, utilizing a sub-compact cyclotron as neutron source and equipped with vertical and horizontal irradiation ports, is presented in this article. A series of experiment, prior to its construction, was conducted using beams of a variable energy cyclotron at Tohoku University to investigate the characteristics of thermal neutron obtained, from 9 Be (p, n) reaction and thermalized by elastic scattering process. This article describes a computer simulation of neutron moderator to analyze conditions getting maximal thermal neutron flux. Some of practical neutron radiograph examination of aero-space components and museum art objects of classic bronze mirror are also presented together with an attempt realizing real time imaging technique. (author)

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

Science.gov (United States)

Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

2018-01-01

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

Histological and Physiological Alterations Induced by Thermal Neutron Fluxes in Male Swiss Albino Mice

International Nuclear Information System (INIS)

Alzergy, A.A.; Emara, N.M.; Abd El-Latif, A.A.; El-Saady, S.M.M.; Emara, N.M.; Abd El-Latif, A.A.

2010-01-01

This work was performed to investigate the biological effects of different thermal neutron fluxes (0.27x10 8 , 0.52X10 8 , 1.089X10 8 , 2.16X10 8 and 4.32X10 8 ) on liver and kidney of male mice using neutron irradiation cell with Ra-Be(α,n) 3 mCi neutron source Leybold (55930). Exposed to various fluxes of thermal neutron induced a dramatic alterations in hepatic and renal functions as indicated by biochemical estimation of several parameters (bilirubin, SGT, and alkaline phosphate .Urea , total protein, and albumin) and confirmed by histological examinations Thermal neutron exposure induces marked increase in the serum activities of total bilirubin, alanine amino transaminase (ALT or GPT), and alkaline phosphate, whereas, urea, total protein and albumin showed marked decline as compared to control group. The physiological changes induced in thermal neutron fluxes dependent manner. Histopathological results revealed mild to severe type of necrosis, and degenerative changes in liver and kidney of male mice exposed to thermal neutron fluxes. Also it was found that the histopathological alterations induced in thermal neutron fluxes dependent manner. It was found that exposed to thermal neutron fluxes irradiation plays prominent role in the development of the physiological alterations in male Swiss albino mice. The Former up normalities as a result of the sequence events followed interaction of radiation with the former biological mater (liver and kidney) of male Swiss albino mice, which are, physical, physicochemical, chemical, and biological stages.

Manufacturing of thermal neutron sensor using pMOS

International Nuclear Information System (INIS)

Lee, Nam Ho; Kim, Seung Ho

2005-05-01

A pMOSFET sensor having a Gadolinium converter has been invented successfully as a slow neutron sensor that is sensitive to neutron energy down to 0.025 eV. The Gd layer converts low energy neutrons to ionizing radiation of which the amount is proportional to neutron dose. Ionising radiation from neutron reactions changes the charge state of the gate oxide of the pMOSFET. The Gd-pMOSFETs were tested at a neutron beam port of HANARO research reactor and a 60 CO irradiation facility to investigate slow neutron response and gamma response, respectively. The voltage change was proportional to the accumulated slow neutron dose. The results from Gd coupled MOSFET neutron dosemeters shows an excellent sensitivity (15 - 16mV/cGy) and linearity to thermal neutrons with negligible background contamination. The results demonstrate the outstanding performance of the Gd coupled MOSFET neutron dosemeters clearly. The Gd-pMOSFET can also be used in a mixed radiation field by subtracting the voltage change of a pMOSFET without Gd from that of the Gd-pMOSFET

Future prospects of imaging at spallation neutron sources

International Nuclear Information System (INIS)

Strobl, M.

2009-01-01

The advent of state-of-the-art spallation neutron sources is a major step forward in efficient neutron production for most neutron scattering techniques. Although they provide lower time-averaged neutron flux than high flux reactor sources, advantage for different instrumental techniques can be derived from the pulsed time structure of the available flux, which can be translated into energy, respectively, wavelength resolution. Conventional neutron imaging on the other hand relies on an intense continuous beam flux and hence falls short in profiting from the new development. Nevertheless, some recently developed novel imaging techniques require and some can benefit from energy resolution. The impact of the emerging spallation sources on different imaging techniques has been investigated, ways to benefit will be identified (where possible) and prospects of future imaging instruments and possible options and layouts at a spallation neutron source will be discussed and outlined.

Neutron beam applications - A development of real-time imaging processing for neutron radiography

Energy Technology Data Exchange (ETDEWEB)

Kim, Whoi Yul; Lee, Sang Yup; Choi, Min Seok; Hwang, Sun Kyu; Han, Il Ho; Jang, Jae Young [Hanyang University, Seoul (Korea)

1999-08-01

This research is sponsored and supported by KAERI as a part of {sup A}pplication of Neutron Radiography Beam.{sup M}ain theme of the research is to develop a non-destructive inspection system for the task of studying the real-time behaviour of dynamic motion using neutron beam with the aid of a special purpose real-time image processing system that allows to capture an image of internal structure of a specimen. Currently, most off-the-shelf image processing programs designed for visible light or X-ray are not adequate for the applications that require neutron beam generated by the experimental nuclear reactor. In addition, study of dynamic motion of a specimen is severely constrained by such image processing systems. In this research, a special image processing system suited for such application is developed which not only supplements the commercial image processing system but allows to use neutron beam directly in the system for the study. 18 refs., 21 figs., 1 tab. (Author)

Development of Neutron Imaging System for Neutron Tomography at Thai Research Reactor TRR-1/M1

Science.gov (United States)

Wonglee, S.; Khaweerat, S.; Channuie, J.; Picha, R.; Liamsuwan, T.; Ratanatongchai, W.

2017-09-01

The neutron imaging is a powerful non-destructive technique to investigate the internal structure and provides the information which is different from the conventional X-ray/Gamma radiography. By reconstruction of the obtained 2-dimentional (2D) images from the taken different angle around the specimen, the tomographic image can be obtained and it can provide the information in more detail. The neutron imaging system at Thai Research Reactor TRR-1/M1 of Thailand Institute of Nuclear Technology (Public Organization) has been developed to conduct the neutron tomography since 2014. The primary goal of this work is to serve the investigation of archeological samples, however, this technique can also be applied to various fields, such as investigation of industrial specimen and others. This research paper presents the performance study of a compact neutron camera manufactured by Neutron Optics such as speed and sensitivity. Furthermore, the 3-dimentional (3D) neutron image was successfully reconstructed at the developed neutron imaging system of TRR-1/M1.

Development of the variety for resistance against bacterial leaf-blight in rice with thermal neutrons

International Nuclear Information System (INIS)

Nakai, Hirokazu

1990-01-01

In search for the development of genes for resistance against bacterial leaf-blight in rice, thermal neutrons generated from the Research Reactor at the Kyoto University have been applied to the breeding. In this paper, the developmental outcome is described, and a potential application of thermal neutrons for breeding the variety of resistance against bacterial leaf-blight in rice is reviewed. When thermal neutrons were delivered to the rice, the ratio of absorbed doses by B-10, which is contained in a small quantity in the plant, was found to be larger than expected. This implies characteristic effects of thermal neutrons on the plant. When boric acid was incorporated into the plant before irradiation, the effect of thermal neutrons per irradiation time was considered to become great. The frequency of mutations for resistance was significantly higher by thermal neutron, as compared with that induced by other mutagens, such as gamma radiation, ethylene-imine, ethyl-methane-sulfonate, and nitroso-methyl-urea. Genetic analysis of mutants for resistance revealed recessive genes and polygenes. Finally, the application of thermal neutrons and other radiations would contribute greatly to a resolution of serious pollution problems in global food and environment. (N.K.)

Status of the Neutron Imaging and Diffraction Instrument IMAT

Science.gov (United States)

Kockelmann, Winfried; Burca, Genoveva; Kelleher, Joe F.; Kabra, Saurabh; Zhang, Shu-Yan; Rhodes, Nigel J.; Schooneveld, Erik M.; Sykora, Jeff; Pooley, Daniel E.; Nightingale, Jim B.; Aliotta, Francesco; Ponterio, Rosa C.; Salvato, Gabriele; Tresoldi, Dario; Vasi, Cirino; McPhate, Jason B.; Tremsin, Anton S.

A cold neutron imaging and diffraction instrument, IMAT, is currently being constructed at the ISIS second target station. IMAT will capitalize on time-of-flight transmission and diffraction techniques available at a pulsed neutron source. Analytical techniques will include neutron radiography, neutron tomography, energy-selective neutron imaging, and spatially resolved diffraction scans for residual strain and texture determination. Commissioning of the instrument will start in 2015, with time-resolving imaging detectors and two diffraction detector prototype modules. IMAT will be operated as a user facility for material science applications and will be open for developments of time-of-flight imaging methods.

Neutron irradiation effect of thermally-sensitized stainless steels

Energy Technology Data Exchange (ETDEWEB)

Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

1998-03-01

Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

Using MCNP-4C code for design of the thermal neutron beam for neutron radiography at the MNSR

International Nuclear Information System (INIS)

Shaaban, I.

2009-11-01

Studies were carried out for determination of the parameters of a thermal neutron beam at the MNSR reactor (MNSR-30 kW) for neutron radiography in the vertical beam port by using the MCNP-4C (Monte Carlo Neutron - Photon transport). Thermal, epithermal and fast neutron energy ranges were selected as 10 keV respectively. To produce a good neutron beam in terms of intensity and quality, several materials Lead (Pb), Bismuth (Bi), Borated polyethelyene and Alumina Oxide (Al 2 O 3 ) were used as neutron and photon filters. Based on the current design, the L/D of the facility ranges between 125, 110 and 90. The thermal neutron flux at the beam exit is 1.436x10 5 n/cm2 .s ,1.843x10 5 n/cm2 .s and 2.845x10 5 n/cm2 .s respectively, middots with a Cd-ratio of ∼ 2.829, 2.766, 3.191 for the L/D = 125, 110, 90 respectively. The estimated values for gamma doses are 6.705x10 -2 Rem/h and 1.275x10 -1 Rem/h and 2.678x10 -1 Rem/ h with bismuth. The divergent angle of the collimator is 1.348 degree - 2.021 degree. Such neutron beams, if built into the Syrian MNSR reactor, could support the application of NRG in Syria. (author)

Comparing neutron and X-ray images from NIF implosions

Directory of Open Access Journals (Sweden)

Wilson D.C.

2013-11-01

Full Text Available Directly laser driven and X-radiation driven DT filled capsules differ in the relationship between neutron and X-ray images. Shot N110217, a directly driven DT-filled glass micro-balloon provided the first neutron images at the National Ignition Facility. As seen in implosions on the Omega laser, the neutron image can be enclosed inside time integrated X-ray images. HYDRA simulations show the X-ray image is dominated by emission from the hot glass shell while the neutron image arises from the DT fuel it encloses. In the absence of mix or jetting, X-ray images of a cryogenically layered THD fuel capsule should be dominated by emission from the hydrogen rather than the cooler plastic shell that is separated from the hot core by cold DT fuel. This cool, dense DT, invisible in X-ray emission, shows itself by scattering hot core neutrons. Germanium X-ray emission spectra and Ross pair filtered X-ray energy resolved images suggest that germanium doped plastic emits in the torus shaped hot spot, probably reducing the neutron yield.

Optimization of thermal neutron shield concrete mixture using artificial neural network

Energy Technology Data Exchange (ETDEWEB)

Yadollahi, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Nazemi, E., E-mail: nazemi.ehsan@yahoo.com [Young Researchers and Elite Club, Kermanshah Branch, Islamic Azad University, Kermanshah (Iran, Islamic Republic of); Zolfaghari, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Ajorloo, A.M. [Water and Environmental Engineering Department, Shahid Beheshti University, P.O. Box: 167651719, Tehran (Iran, Islamic Republic of)

2016-08-15

Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m{sup 3}, a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

Optimization of thermal neutron shield concrete mixture using artificial neural network

International Nuclear Information System (INIS)

Yadollahi, A.; Nazemi, E.; Zolfaghari, A.; Ajorloo, A.M.

2016-01-01

Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m 3 , a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

Thermal neutron converter for irradiations with fission neutrons

International Nuclear Information System (INIS)

Wagner, F.M.; Kampfer, S.; Kastenmuller, A.; Waschkowski, W.; Bucherl, Th.; Kampfer, S.

2007-01-01

The new research reactor FRM II at Garching started operation in March 2004. The compact core is cooled by light water, and moderated by heavy water. Two fuel plates mounted in the heavy water tank convert thermal to fast neutrons. The fast neutron flux in the connected beam tube is up to 7 centre dot 10 8 s -1 cm -2 (depending on filters and collimation); the mean neutron energy is about 1.6 MeV. There are two irradiation rooms along the beam. The first is mainly used for medical therapy (MEDAPP facility), the second for materials characterization (NECTAR facility). At the former therapy facility RENT at the old research reactor FRM, the same beam quality was available until July 2000. Therefore, only a small program is run for the determination of the biological effectiveness of the new beam. The neutron and gamma dose rates in the medical beam are 0.54 and 0.20 Gy/min, respectively. The therapy facility MEDAPP is still under examination according to European regulations for medical devices. Full medical operation will start in 2007. The radiography and tomography facility NECTAR is in operation and aims at non-destructive inspection of objects up to 400 kg mass and 80 centre dot 80 centre dot 80 cm 3 in size. As for fission neutrons the macroscopic cross section of hydrogen is much higher than for other materials (e. g. Fe and Pb), one special application is the detection of hydrogen-containing materials (e. g. oil) in dense materials

Moderator/collimator for a proton/deuteron linac to produce a high-intensity, high-quality thermal neutron beam for neutron radiography

International Nuclear Information System (INIS)

Singleterry, R.C. Jr.; Imel, G.R.; McMichael, G.E.

1995-01-01

Reactor based high resolution neutron radiography facilities are able to deliver a well-collimated (L/D ≥100) thermal flux of 10 6 n/cm 2 ·sec to an image plane. This is well in excess of that achievable with the present accelerator based systems such as sealed tube D-T sources, Van der Graaff's, small cyclotrons, or low duty factor linacs. However, continuous wave linacs can accelerate tens of milliamperes of protons to 2.5 to 4 MeV. The MCNP code has been used to analyze target/moderator configurations that could be used with Argonne's Continuous Wave Linac (ACWL). These analyses have shown that ACWL could be modified to generate a neutron beam that has a high intensity and is of high quality

Determination of Thermal Neutron Capture Cross Sections Using Cold Neutron Beams at the Budapest PGAA-NIPS Facilities

International Nuclear Information System (INIS)

Belgya, T.

2006-01-01

A complete elemental gamma-ray library was measured with our guided thermal beam at the Budapest PGAA facility in the period of 1995-2000. Using this data library in an IAEA CRP on PGAA it was managed to re-normalize the ENSDF intensity data with the Budapest intensities. Based on this renormalization thermal neutron cross sections were deduced for several isotopes. Most of these calculations were done by Richard B. Firestone. The Budapest PGAA-NIPS facilities have been used for routine prompt gamma activation analysis with cold neutrons since the year of 2000. The advantage of the cold neutron beam is that the neutron guide has much higher neutron transmission. This resulted in a gain factor about 20 relative to our thermal guide. For the analytical works a precise comparator technique was developed that is routinely used to determine partial gamma-ray production cross sections. An additional development of our methodology was necessary to be worked out to determine thermal neutron capture cross sections based on the partial gamma-ray production cross sections. In this talk our methodology of radiative capture cross section determination will be presented, including our latest results on 129 I, 204,206,207 Pb and 209 Bi. Most of these works were done in cooperation with people from EU-JRC-IRMM, Geel, Belgium and CEA Cadarache, France. Many partial cross sections of short lived nuclei have been re-measured with our new chopper technique. The uncertainty calculations of the radiative capture cross section determination procedures will be also shown. (authors)

Zone plate imaging of 14-MeV neutrons

International Nuclear Information System (INIS)

Lerche, R.A.; Lane, S.M.; Hawryluk, A.M.; Ceglio, N.M.

1986-01-01

At Livermore we are interested in imaging the thermonuclear burn region of fusion targets irradiated at our Nova laser facility. We expect compressed core diameters to be 10's of microns, and would like images with better than 10-μm resolution. Alpha particle images provided the first direct information about the thermonuclear burn geometry in thin walled exploding pusher targets. In future high density target experiments, only highly penetrating radiations like the 14-MeV neutrons will escape the target core to provide information about the burn region. To make the measurement with a neutron ''pinhole'' camera requires a 10μm pinhole through about 10 cm of material and 10 14 to 10 15 source neutrons. Penumbral imaging offers some improvement over a pinhole. Zone plate coded imaging (ZPCI) techniques are particularly well suited for imaging small objects like the compressed core of a laser fusion target. We have been using ZPCI techniques to image nonpenetrating radiations like x rays and alpha particles for about 10 years. The techniques are well developed. Imaging penetrating radiations like 14-MeV neutrons using ZPCI techniques has several possible advantages. The large solid angle subtended by the Zone plate might substantially reduce the required target neutron yield needed to produce a useful image, and a neutron zone plate system with 10-μm resolution might be easier to fabricate and characterize than a pinhole system. This paper explores the use of ZPCI techniques with penetrating radiation

Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

2016-01-01

In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)

A new position-sensitive detector for thermal and epithermal neutrons

International Nuclear Information System (INIS)

Jeavons, A.P.; Ford, N.L.; Lindberg, B.; Sachot, R.

1977-01-01

A new two-dimensional position-sensitive neutron detector is described. It is based on (n,γ) neutron resonance capture in a foil with subsequent detection of internal conversion electrons with a high-density proportional chamber. Large-area detectors with a 1 mm spatial resolution are feasible. A detection efficiency of 50% is possible for thermal neutrons using gadolinium-157 foil and for epithermal neutrons using hafnium-177. (Auth.)

One-dimensional neutron imager for the Sandia Z facility.

Science.gov (United States)

Fittinghoff, David N; Bower, Dan E; Hollaway, James R; Jacoby, Barry A; Weiss, Paul B; Buckles, Robert A; Sammons, Timothy J; McPherson, Leroy A; Ruiz, Carlos L; Chandler, Gordon A; Torres, José A; Leeper, Ramon J; Cooper, Gary W; Nelson, Alan J

2008-10-01

A multiinstitution collaboration is developing a neutron imaging system for the Sandia Z facility. The initial system design is for slit aperture imaging system capable of obtaining a one-dimensional image of a 2.45 MeV source producing 5x10(12) neutrons with a resolution of 320 microm along the axial dimension of the plasma, but the design being developed can be modified for two-dimensional imaging and imaging of DT neutrons with other resolutions. This system will allow us to understand the spatial production of neutrons in the plasmas produced at the Z facility.

The determination of thermal neutron cross section of 81Br

International Nuclear Information System (INIS)

Kovacs, Luciana; Zamboni, Cibele B.; Dalaqua Junior, Leonardo

2009-01-01

In this investigation several standard materials were used to determine the thermal neutron cross section of 81 Br. This nuclear parameter is an important data to perform several quantitative investigations, mainly in medical area. In other to confirm and to reduce the uncertainty, a new measurement was preformed using thermal neutron at IEA-R1 nuclear reactor of IPEN/CNEN-SP. The result obtained is compatible with the tabulated value and present small uncertainly. (author)

Lethal Effect of Thermal Neutrons on Hypoxic Elirlich Ascites Tumour Cells in vitro

OpenAIRE

MITSUHIKO, AKABOSHI; KENICHI, KAWAI; HIROTOSHI, MAKI; Research Reactor Institute, Kyoto University; Research Reactor Institute, Kyoto University; Research Reactor Institute, Kyoto University

1985-01-01

Ehrlich ascites tumour cells were irradiated in vitro with thermal neutrons under aerobic and hypoxic conditions, and the survival of their reproductive capacity was assayed in vivo. Only a slight hypoxic protection was observed for thermal neutron irradiation with an oxygen enhancement ratio (OER) of 1.2, as compared with OER of 3.3 for ^Co-γ-rays. Absorbed dose of thermal neutrons was calculated by assuming that the energies of recoiled nuclei were completely absorbed within a cell nucleus....

Thermal hydraulic and neutronic interaction in the rotating bed reactor

International Nuclear Information System (INIS)

Lee, C.C.

1986-01-01

Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors

Thermal neutron diffusion parameters dependent on the flux energy distribution in finite hydrogenous media

International Nuclear Information System (INIS)

Drozdowicz, K.

1999-01-01

Macroscopic parameters for a description of the thermal neutron transport in finite volumes are considered. A very good correspondence between the theoretical and experimental parameters of hydrogenous media is attained. Thermal neutrons in the medium possess an energy distribution, which is dependent on the size (characterized by the geometric buckling) and on the neutron transport properties of the medium. In a hydrogenous material the thermal neutron transport is dominated by the scattering cross section which is strongly dependent on energy. A monoenergetic treatment of the thermal neutron group (admissible for other materials) leads in this case to a discrepancy between theoretical and experimental results. In the present paper the theoretical definitions of the pulsed thermal neutron parameters (the absorption rate, the diffusion coefficient, and the diffusion cooling coefficient) are based on Nelkin's analysis of the decay of a neutron pulse. Problems of the experimental determination of these parameters for a hydrogenous medium are discussed. A theoretical calculation of the pulsed parameters requires knowledge of the scattering kernel. For thermal neutrons it is individual for each hydrogenous material because neutron scattering on hydrogen nuclei bound in a molecule is affected by the molecular dynamics (characterized with internal energy modes which are comparable to the incident neutron energy). Granada's synthetic model for slow-neutron scattering is used. The complete up-dated formalism of calculation of the energy transfer scattering kernel after this model is presented in the paper. An influence of some minor variants within the model on the calculated differential and integral neutron parameters is shown. The theoretical energy-dependent scattering cross section (of Plexiglas) is compared to experimental results. A particular attention is paid to the calculation of the diffusion cooling coefficient. A solution of an equation, which determines the

Impact of neutron irradiation on thermal helium desorption from iron

Energy Technology Data Exchange (ETDEWEB)

Hu, Xunxiang, E-mail: hux1@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Field, Kevin G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Taller, Stephen [University of Michigan, Ann Arbor, MI 48109 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wirth, Brian D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States)

2017-06-15

The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 10{sup 18} He/m{sup 2}. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

International Nuclear Information System (INIS)

Matausek, M.V.; Kunc, M.

1974-01-01

When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

Neutron radiography using a transportable superconducting cyclotron

Energy Technology Data Exchange (ETDEWEB)

Allen, D.A. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Hawkesworth, M.R. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Beynon, T.D. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Green, S. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Rogers, J.D. (Rolls-Royce, Derby (United Kingdom)); Allen, M.J. (Rolls-Royce, Derby (United Kingdom)); Plummer, H.C. (Rolls-Royce, MatEval, Derby (United Kingdom)); Boulding, N.J. (Oxford Instruments (United Kingdom)); Cox, M. (Oxford Instruments (United Kingdom)); McDougall, I. (Oxford Instruments (United Kingdom))

1994-12-30

A thermal neutron radiography system based on a compact 12 MeV superconducting proton cyclotron is described. Neutrons are generated using a thick beryllium target and moderated in high density polyethylene. Monte Carlo computer simulations have been used to model the neutron and photon transport in order to optimise the performance of the system. With proton beam currents in excess of 100 [mu]A, it can provide high thermal neutron fluxes with L/D ratios of between 50 and 300 for various applications. Both film and electronic imaging are used to produce radiographs. The electronic imaging system consists of a [sup 6]Li-loaded ZnS intensifier screen, and a low light CCD or SIT camera. High resolution images can be recorded and computer-controlled data processing, analysis and display are possible. ((orig.))

Thermal neutron scattering kernels for sapphire and silicon single crystals

International Nuclear Information System (INIS)

Cantargi, F.; Granada, J.R.; Mayer, R.E.

2015-01-01

Highlights: • Thermal cross section libraries for sapphire and silicon single crystals were generated. • Debye model was used to represent the vibrational frequency spectra to feed the NJOY code. • Sapphire total cross section was measured at Centro Atómico Bariloche. • Cross section libraries were validated with experimental data available. - Abstract: Sapphire and silicon are materials usually employed as filters in facilities with thermal neutron beams. Due to the lack of the corresponding thermal cross section libraries for those materials, necessary in calculations performed in order to optimize beams for specific applications, here we present the generation of new thermal neutron scattering kernels for those materials. The Debye model was used in both cases to represent the vibrational frequency spectra required to feed the NJOY nuclear data processing system in order to produce the corresponding libraries in ENDF and ACE format. These libraries were validated with available experimental data, some from the literature and others obtained at the pulsed neutron source at Centro Atómico Bariloche

Neutron imaging options at the BOA beamline at Paul Scherrer Institut

International Nuclear Information System (INIS)

Morgano, M.; Peetermans, S.; Lehmann, E.H.; Panzner, T.; Filges, U.

2014-01-01

The BOA beamline at the Swiss spallation neutron source SINQ at Paul Scherrer Institut is a flexible instrument used mainly for testing novel techniques and devices for neutron scattering and optics, but, due to the large and relatively homogeneous field of view, it can be successfully used for experiments in the field of neutron imaging. The beamline allows also for the exploitation of advanced imaging concepts such as polarized neutron imaging and diffractive neutron imaging. In this paper we present the characterization of the BOA beamline in the light of its neutron imaging capabilities. We show also the different techniques that can be employed there as user-friendly plugins for non-standard neutron imaging experiments

Fast-neutron, coded-aperture imager

Science.gov (United States)

Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.

2015-06-01

This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led

Thermal and magnetic properties of neutron matter

International Nuclear Information System (INIS)

Abd-Alla, M.; Ragab, H.S.; Hassan, M.Y.M.

1990-01-01

The Thomas-Fermi model is used to calculate the equation of state of thermal polarized neutron matter applying Seyler-Blanchard interaction. The resulting equation of state is stiff and has a small dependence on both the temperature and the spin excess parameter. We expand the Fermi integrals in powers of temperature up to second order to examine the T 2 approximation for neutron matter. It is found to be reliable up to T = 10 MeV. We also studied the ferromagnetic transition in neutron matter. We found a ferromagnetic transition at density ρ ≅ 2ρ0. This ferromagnetic transition is found to have a small dependence on both the temperature and the spin excess parameter. We also studied the dependence of the effective mass and the sound velocity for polarized neutron matter on temperature. (author). 36 refs, 17 figs

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

Transparent lithiated polymer films for thermal neutron detection

Energy Technology Data Exchange (ETDEWEB)

Mabe, Andrew N., E-mail: andrew.n.mabe@gmail.com [Department of Chemistry, University of Tennessee, Knoxville, TN 37996 (United States); Auxier, John D. [Department of Chemistry, University of Tennessee, Knoxville, TN 37996 (United States); Urffer, Matthew J. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Penumadu, Dayakar [Department of Civil and Environmental Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Schweitzer, George K. [Department of Chemistry, University of Tennessee, Knoxville, TN 37996 (United States); Miller, Laurence F. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States)

2013-09-11

Novel water-soluble {sup 6}Li loaded copolymer scintillation films have been designed and fabricated to detect thermal neutrons. Styrene and maleic anhydride were copolymerized to form an alternating copolymer, then the anhydride functionality was hydrolyzed using {sup 6}Li hydroxide. The resulting poly(styrene-co-lithium maleate) was mixed with salicylic acid as a fluor and cast as a thin film from water. The maximum {sup 6}Li loading obtained that resulted in a transparent film was 4.36% by mass ({sup 6}Li to polymer). The optimum fluorescence output was obtained for 11.7% salicylic acid by mass, presumably in the form of lithium salicylate, resulting in an optimum film containing 3.85% by mass of {sup 6}Li. A facile and robust synthesis method, film fabrication protocol, photoluminescence results, and scintillation responses are reported herein. -- Highlights: • A transparent polymer scintillator containing 3.85 wt% {sup 6}Li has been synthesized. • This class of polymeric thermal neutron scintillation detector is water-soluble. • Salicylic acid, presumably in the form of lithium salicylate, is used as a fluor. • The material emits 373 photons/α ({sup 241}Am) and an average of 139 photons/β ({sup 36}Cl). • The material emits 360 photons per thermal neutron capture event.

Non-destructive material investigation with thermal neutrons at the TRIGA Mark II reactor in Vienna

International Nuclear Information System (INIS)

Bastuerk, M.; Boeck, H.; Zamani, B.; Zawisky, M.; Rauch, H.

2004-01-01

Neutron tomography providing 3D information about interior of an object is a very efficient tool to visualize inner defects of the materials, non-destructively. In this study, some applications of neutron tomography in different fields such as geology, aerospace, civil engineering and archaeology were presented. Distribution of minerals in pumice and rock samples, visualization of inner defects within a new developed titan aluminum turbine blade, and distribution of silica gel as an important impregnating agent in construction and restoration of buildings were investigated. The measurements of tomography projections taken in the 0 to 180 o angle were performed with a thermal neutron flux of 10 5 at the TRIGA Mark II research reactor in Vienna, and the common filtered back projection method was used for the 3D image reconstruction. (author)

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2009-01-01

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

Gadolinium oxide coated fully depleted silicon-on-insulator transistors for thermal neutron dosimetry

Energy Technology Data Exchange (ETDEWEB)

Vitale, Steven A., E-mail: steven.vitale@ll.mit.edu; Gouker, Pascale M.

2013-09-01

Fully depleted silicon-on-insulator transistors coated with gadolinium oxide are shown to be effective thermal neutron dosimeters. The theoretical neutron detection efficiency is calculated to be higher for Gd{sub 2}O{sub 3} than for other practical converter materials. Proof-of-concept dosimeter devices were fabricated and tested during thermal neutron irradiation. The transistor current changes linearly with neutron dose, consistent with increasing positive charge in the SOI buried oxide layer generated by ionization from high energy {sup 157}Gd(n,γ){sup 158}Gd conversion electrons. The measured neutron sensitivity is approximately 1/6 the maximum theoretical value, possibly due to electron–hole recombination or conversion electron loss in interconnect wiring above the transistors. -- Highlights: • A novel Gd{sub 2}O{sub 3} coated FDSOI MOSFET thermal neutron dosimeter is presented. • Dosimeter can detect charges generated from {sup 157}Gd(n,γ){sup 158}Gd conversion electrons. • Measured neutron sensitivity is comparable to that calculated theoretically. • Dosimeter requires zero power during operation, enabling new application areas.

Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

Science.gov (United States)

Basiri, H.; Tavakoli-Anbaran, H.

2018-01-01

Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

Dosimetry boron neutron capture therapy in liver cancer (hepatocellular carcinoma) by means of MCNP-code with neutron source from thermal column

International Nuclear Information System (INIS)

Irhas; Andang Widi Harto; Yohannes Sardjono

2014-01-01

Boron Neutron Capture Therapy (BNCT) using physics principle when B 10 (Boron-10) irradiated by low energy neutron (thermal neutron). Boron and thermal neutron reaction produced B 11m (Boron-11m) (t 1/2 =10 -2 s). B 11m decay emitted alpha, Li 7 (Lithium-7) particle and gamma ray. Irradiated time needed to ensure cancer dose enough. Liver cancer was primary malignant who located in liver (Hepatocellular carcinoma). Malignant in liver were different to metastatic from Breast, Colon Cancer, and the other. This condition was Metastatic Liver Cancer. Monte Carlo method used by Monte Carlo N-Particle (MCNP) Software. Probabilistic approach used for probability of interaction occurred and record refers to characteristic of particle and material. In this case, thermal neutron produced by model of Collimated Thermal Column Kartini Research Nuclear Reactor, Yogyakarta. Modelling organ and source used liver organ that contain of cancer tissue and research reactor. Variation of boron concentration was 20, 25, 30, 35, 40, 45, and 47 µg/g cancers. Output of MCNP calculation were neutron scattering dose, gamma ray dose and neutron flux from reactor. Neutron flux used to calculate alpha, proton and gamma ray dose from interaction of tissue material and thermal neutron. Variation of boron concentration result dose rate to every variation were 0,059; 0,072; 0,084; 0,098; 0.108; 0,12; 0,125 Gy/sec. Irradiation time who need to every concentration were 841,5 see (14 min 1 sec); 696,07 sec(11 min 36 sec); 593.11 sec (9 min 53 sec); 461,35 sec (8 min 30 sec); 461,238 sec (7 min 41 sec); 414,23 sec (6 min 54 sec); 398,38 sec (6 min 38 sec). Irradiating time could shortly when boron concentration more high. (author)

Study of a transportable neutron radiography system

International Nuclear Information System (INIS)

Souza, S.N.A. de.

1991-05-01

This work presents a study a transportable neutron radiography system for a 185 GBq 241 Am-Be (α, η) source with a neutron yield roughly 1,25 x 10 7 n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio of 14, for neutron fluxes up to 4,09 x 10 2 n.cm -2 .s -1 . Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10 2 n.cm -2 .s -1 and 2,65 x 10 2 n.cm -2 .s -1 , respectively. (author). 30 refs, 39 figs, 9 tabs

Heat generation and temperature-rise in ordinary concrete due to capture of thermal neutrons

International Nuclear Information System (INIS)

Abdo, E.A.; Amin, E.

1997-01-01

The aim of this work is the evaluation of the heat generation and temperature-rise in local ordinary concrete as a biological shield due to capture of total thermal and reactor thermal neutrons. The total thermal neutron fluxes were measured and calculated. The channel number 2 of the ETRR-1 reactor was used in the measurements as a neutron source. Computer code ANISN (VAX version) and neutron multigroup cross-section library EURLiB-4 was used in the calculations. The heat generation and temperature-rise in local ordinary concrete were evaluated and calculated. The results were displayed in curves to show the distribution of thermal neutron fluxes and heat generation as well as temperature-rise with the shield thickness. The results showed that, the heat generation as well as the temperature-rise have their maximum values in the first layers of the shield thickness. 4 figs., 12 refs

Neutron spectral modulation as a new thermal neutron scattering technique. Pt. 1

International Nuclear Information System (INIS)

Ito, Y.; Nishi, M.; Motoya, K.

1982-01-01

A thermal neutron scattering technique is presented based on a new idea of labelling each neutron in its spectral position as well as in time through the scattering process. The method makes possible the simultaneous determination of both the accurate dispersion relation and its broadening by utilizing the resolution cancellation property of zero-crossing points in the cross-correlated time spectrum together with the Fourier transform scheme of the neutron spin echo without resorting to the echoing. The channel Fourier transform applied to the present method also makes possible the determination of the accurate direct energy scan profile of the scattering function with a rather broad incident neutron wavelength distribution. Therefore the intensity sacrifice for attaining high accurarcy is minimized. The technique is used with either a polarized or unpolarized beam at the sample position with no precautions against beam depolarization at the sample for the latter case. Relative time accurarcy of the order of 10 -3 to 10 -4 may be obtained for the general dispersion relation and for the quasi-elastic energy transfers using correspondingly the relative incident neutron wavelength spread of 10 to 1% around an incident neutron energy of a few meV. (orig.)

Electronic imaging applied to neutron radiography

International Nuclear Information System (INIS)

Garrett, D.A.; Bracher, D.A.

1976-01-01

A commercially - available image intensifier was used with a scan conversion memory and a mobile 252 Cf based neutron radiography system to obtain neutron radiographs on a television monitor in 0.5 minutes to 10.0 minutes

New neutron imaging techniques to close the gap to scattering applications

International Nuclear Information System (INIS)

Lehmann, Eberhard H.; Peetermans, S.; Trtik, P.; Betz, B.; Grünzweig, C.

2017-01-01

Neutron scattering and neutron imaging are activities at the strong neutron sources which have been developed rather independently. However, there are similarities and overlaps in the research topics to which both methods can contribute and thus useful synergies can be found. In particular, the spatial resolution of neutron imaging has improved recently, which - together with the enhancement of the efficiency in data acquisition- can be exploited to narrow the energy band and to implement more sophisticated methods like neutron grating interferometry. This paper provides a report about the current options in neutron imaging and describes how the gap to neutron scattering data can be closed in the future, e.g. by diffractive imaging, the use of polarized neutrons and the dark-field imagining of relevant materials. This overview is focused onto the interaction between neutron imaging and neutron scattering with the aim of synergy. It reflects mainly the authors’ experiences at their PSI facilities without ignoring the activities at the different other labs world-wide. (paper)

New neutron imaging techniques to close the gap to scattering applications

Science.gov (United States)

Lehmann, Eberhard H.; Peetermans, S.; Trtik, P.; Betz, B.; Grünzweig, C.

2017-01-01

Neutron scattering and neutron imaging are activities at the strong neutron sources which have been developed rather independently. However, there are similarities and overlaps in the research topics to which both methods can contribute and thus useful synergies can be found. In particular, the spatial resolution of neutron imaging has improved recently, which - together with the enhancement of the efficiency in data acquisition- can be exploited to narrow the energy band and to implement more sophisticated methods like neutron grating interferometry. This paper provides a report about the current options in neutron imaging and describes how the gap to neutron scattering data can be closed in the future, e.g. by diffractive imaging, the use of polarized neutrons and the dark-field imagining of relevant materials. This overview is focused onto the interaction between neutron imaging and neutron scattering with the aim of synergy. It reflects mainly the authors’ experiences at their PSI facilities without ignoring the activities at the different other labs world-wide.

Thermal imaging in medicine

Directory of Open Access Journals (Sweden)

Jaka Ogorevc

2015-12-01

Full Text Available AbstractIntroduction: Body temperature monitoring is one of the oldest and still one of the most basic diagnostic methods in medicine. In recent years thermal imaging has been increasingly used in measurements of body temperature for diagnostic purposes. Thermal imaging is non-invasive, non-contact method for measuring surface body temperature. Method is quick, painless and patient is not exposed to ionizing radiation or any other body burden.Application of thermal imaging in medicine: Pathological conditions can be indicated as hyper- or hypothermic patterns in many cases. Thermal imaging is presented as a diagnostic method, which can detect such thermal anomalies. This article provides an overview of the thermal imaging applications in various fields of medicine. Thermal imaging has proven to be a suitable method for human febrile temperature screening, for the detection of sites of fractures and infections, a reliable diagnostic tool in the detection of breast cancer and determining the type of skin cancer tumour. It is useful in monitoring the course of a therapy after spinal cord injury, in the detection of food allergies and detecting complications at hemodialysis and is also very effective at the course of treatment of breast reconstruction after mastectomy. With thermal imaging is possible to determine the degrees of burns and early detection of osteomyelitis in diabetic foot phenomenon. The most common and the oldest application of thermal imaging in medicine is the field of rheumatology.Recommendations for use and standards: Essential performance of a thermal imaging camera, measurement method, preparation of a patient and environmental conditions are very important for proper interpretation of measurement results in medical applications of thermal imaging. Standard for screening thermographs was formed for the human febrile temperature screening application.Conclusion: Based on presented examples it is shown that thermal imaging can

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

Science.gov (United States)

Khattab, K; Sulieman, I

2009-04-01

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Block-Based Compressed Sensing for Neutron Radiation Image Using WDFB

Directory of Open Access Journals (Sweden)

Wei Jin

2015-01-01

Full Text Available An ideal compression method for neutron radiation image should have high compression ratio while keeping more details of the original image. Compressed sensing (CS, which can break through the restrictions of sampling theorem, is likely to offer an efficient compression scheme for the neutron radiation image. Combining wavelet transform with directional filter banks, a novel nonredundant multiscale geometry analysis transform named Wavelet Directional Filter Banks (WDFB is constructed and applied to represent neutron radiation image sparsely. Then, the block-based CS technique is introduced and a high performance CS scheme for neutron radiation image is proposed. By performing two-step iterative shrinkage algorithm the problem of L1 norm minimization is solved to reconstruct neutron radiation image from random measurements. The experiment results demonstrate that the scheme not only improves the quality of reconstructed image obviously but also retains more details of original image.

A two-medium thermal neutron spectrum programme

International Nuclear Information System (INIS)

Bindon, D.C.

1960-07-01

A computer programme is described for computing the thermal neutron spectra and effective cross-sections in a reactor system of two media by the method of H. Takahashi. The programme has been prepared and tested for use with the Ferranti Mercury computer. (author)

Neutron radiography

International Nuclear Information System (INIS)

Alaa eldin, M.T.

2011-01-01

The digital processing of the neutron radiography images gives the possibility for data quantification. In this case an exact relation between the measured neutron attenuation and the real macroscopic attenuation coefficient for every point of the sample is required. The assumption that the attenuation of the neutron beam through the sample is exponential is valid only in an ideal case where a monochromatic beam, non scattering sample and non background contribution are assumed. In the real case these conditions are not fulfilled and in dependence on the sample material we have more or less deviation from the exponential attenuation law. Because of the high scattering cross-sections of hydrogen (σs=80.26 barn) for thermal neutrons, the problem with the scattered neutrons at quantitative radiography investigations of hydrogenous materials (as PE, Oil, H 2 O, etc) is not trivial. For these strong scattering materials the neutron beam attenuation is no longer exponential and a dependence of the macroscopic attenuation coefficient on the material thickness and on the distance between the sample and the detector appears. When quantitative radiography (2 D) or tomography investigations (3 D) are performed, some image correction procedures for a description of the scattering effect are required. This thesis presents a method that can be used to enhance the neutron radiography image for objects with high scattering materials like hydrogen, carbon and other light materials. This method uses the Monte Carlo code, MCNP5, to simulate the neutron radiography process and get the flux distribution for each pixel of the image and determine the scattered neutrons distribution that causes the image blur and then subtract it from the initial image to improve its quality.

Microstructured boron foil scintillating G-GEM detector for neutron imaging

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Takeshi, E-mail: fujiwara-t@aist.go.jp [Research Institute for Measurement and Analytical Instrumentation, Advanced Industrial Science and Technology (AIST), Tsukuba, Ibaraki (Japan); Center for Advanced Photonics, Neutron Beam Technology Team, RIKEN, Saitama (Japan); Bautista, Unico [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Philippine Nuclear Research Institute-Department of Science and Technology (PNRI-DOST), Commonwealth Avenue, Diliman, Quezon City (Philippines); Mitsuya, Yuki [Nuclear Professional School, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki (Japan); Takahashi, Hiroyuki [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Yamada, Norifumi L. [Neutron Science Laboratory, Institute of Material Structure Science, High Energy Accelerator Research Organization (KEK) (Japan); Otake, Yoshie; Taketani, Atsushi [Center for Advanced Photonics, Neutron Beam Technology Team, RIKEN, Saitama (Japan); Uesaka, Mitsuru [Nuclear Professional School, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki (Japan); Toyokawa, Hiroyuki [Research Institute for Measurement and Analytical Instrumentation, Advanced Industrial Science and Technology (AIST), Tsukuba, Ibaraki (Japan)

2016-12-01

In this study, a new simple neutron imaging gaseous detector was successfully developed by combining a micro-structured {sup 10}B foil, a glass gas electron multiplier (G-GEM), and a mirror–lens–charge-coupled device (CCD)–camera system. The neutron imaging system consists of a chamber filled with Ar/CF{sub 4} scintillating gas mixture. Inside this system, the G-GEM is mounted for gas multiplication. The neutron detection in this system is based on the reaction between {sup 10}B and neutrons. A micro-structured {sup 10}B is developed to overcome the issue of low detection efficiency. Secondary electrons excite Ar/CF{sub 4} gas molecules, and high-yield visible photons are emitted from those excited gas molecules during the gas electron multiplication process in the G-GEM holes. These photons are easily detected by a mirror–lens–CCD–camera system. A neutron radiograph is then simply formed. We obtain the neutron images of different materials with a compact accelerator-driven neutron source. We confirm that the new scintillating G-GEM-based neutron imager works properly with low gamma ray sensitivity and exhibits a good performance as a new simple digital neutron imaging device.

Single photon image from PET with insertable collimator for boron neutron capture therapy

International Nuclear Information System (INIS)

Jung, Jooyoung; Suh, Tae Suk; Hong, Key Jo

2014-01-01

Boron neutron capture therapy (BNCT) is a radiation therapy technique for treating deep-seated brain tumors by irradiation with a thermal neutron in which boron-labelled low molecular weight compounds. Once completed, a single photon emission computed tomography (SPECT) scan is conducted to investigate for the region of therapy using an isotope exclusive to SPECT. In the case of an existing PET/SPECT combination system, at least two types of isotopes should be used for each scan with their purposes. Recently, researchers examined the effects of PET/SPECT dual modality on animal imaging systems. They reported that the PET/SPECT combination system was effective for simultaneous achievement of a single event and coincidence. The aim of our proposed system is to confirm the feasibility of extraction of two types of images from one PET module with an insertable collimator for brain tumor treatment during the BNCT. We attempted to acquire the PET and SPECT images simultaneously using only PET without an additional isotope. Single photon images were acquired using an insertable collimator on a PET detector

Determination of thermal neutrons diffusion length in graphite

International Nuclear Information System (INIS)

Garcia Fite, J.

1959-01-01

The diffusion length of thermal neutrons in graphite using the less possible quantity of material has been determined. The proceeding used was the measurement in a graphite pile which has a punctual source of rapid neutrons inside surrounded by a reflector medium (paraffin or water). The measurement was done in the following conditions: a) introducing an aluminium plate between both materials. b) Introducing a cadmium plate between both materials. (Author) 91 refs

Effects of high thermal neutron fluences on Type 6061 aluminum

International Nuclear Information System (INIS)

Weeks, J.R.; Czajkowski, C.J.; Farrell, K.

1992-01-01

The control rod drive follower tubes of the High Flux Beam Reactor are contructed from precipitation-hardened 6061-T6 aluminum alloy and they operate in the high thermal neutron flux regions of the core. It is shown that large thermal neutron fluences up to ∼4 x 10 23 n/cm 2 at 333K cause large increases in tensile strength and relatively modest decreases in tensile elongation while significantly reducing the notch impact toughness at room temperature. These changes are attributed to the development of a fine distribution of precipitates of amorphous silicon of which about 8% is produced radiogenically. A proposed role of thermal-to-fast flux ratio is discussed

A SrBPO5: Eu2+ phosphor for neutron imaging

International Nuclear Information System (INIS)

Sakasai, K.; Katagiri, M.; Toh, K.; Nakamura, T.

2001-01-01

A SrBPO 5 : Eu 2+ phosphor material has been investigated for neutron imaging. This phosphor showed photostimulated luminescence (PSL) by illumination of 635 nm laser light after X-ray irradiation. The spectral characteristics of the phosphor were similar to those of BaFBr: Eu 2+ , which is a commonly used phosphor of imaging plates. In addition, we found that this phosphor also showed PSL for neutron irradiation. It comes from the fact that it contains atomic boron in base matrix. Therefore, this phosphor can be used for neutron imaging without adding neutron sensitive materials such as Gd in commercially available neutron imaging plates. The PSL intensity and the neutron detection will be increased by using enriched boron instead of natural boron. (author)

Preliminary experiment of fast neutron imaging with direct-film method

International Nuclear Information System (INIS)

Pei Yuyang; Tang Guoyou; Guo Zhiyu; Zhang Guohui

2005-01-01

A preliminary experiment is conducted with direct-film method under the condition that fast neutron is generated by the reaction of 9 Be(d, n) on the Beijing University 4.5 MV Van de Graaff, whose energy is lower than 7 MeV. Basic characteristics of direct-film neutron radiography system are investigated with the help of samples in different materials, different thickness and holes of different diameter. The fast neutron converter, which is vital for fast neutron imaging, is produced with the materials made in China. The result indicates that fast neutron converter can meet the requirement of fast neutron imaging; further research of fast neutron imaging can be conducted on the accelerator and neutron-generator in China. (authors)

Light-water-reactor coupled neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Diamond, D.J.

1982-01-01

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

Study on thermal neutron spectra in reactor moderators by time-of-flight method

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-12-01

Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

Neutron and thermal dynamics of a gaseous core fission reactor

International Nuclear Information System (INIS)

van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

1989-01-01

In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

Image enhancement using MCNP5 code and MATLAB in neutron radiography

International Nuclear Information System (INIS)

Tharwat, Montaser; Mohamed, Nader; Mongy, T.

2014-01-01

This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. - Highlights: • This work is applicable for static based film neutron radiography and digital neutron imaging. • MATLAB is a useful tool for imaging enhancement in radiographic film. • Advanced imaging processing is available in the ETRR-2 for imaging processing and data extraction. • The digital imaging system is suitable for complex shapes and sizes, while MATLAB technique is suitable for simple shapes and sizes. • Quantitative measurements are available

Factors affecting neutron measurements and calculations. Part C. Trace element concentrations in granite and their impact on thermal neutron activation

International Nuclear Information System (INIS)

Ruehm, Werner; Huber, Thomas; Nolte, Eckehart; Kato, Kazuo; Imanaka, Tetsuji; Egbert, Stephen D.

2005-01-01

Trace elements such as Li, B, Sm, and Gd can, despite their low elemental concentration in mineral materials, influence thermal neutron activation in Hiroshima and Nagasaki samples, due to their high thermal neutron absorption cross sections. This was demonstrated for a granite core, where the addition of those trace elements to the elemental composition of granite reduces the production of 152 Eu by some 25% at a depth of 25 cm from the surface. If typical concentrations of those trace elements are added to DS02 reference soil, however, the production of 152 Eu one meter above ground is not changed significantly, because of the high water content of the soil. This indicates that DS02 soil represents a reasonable reference material for the air-over-ground transport calculations. It must be kept in mind, however, that the local environment of any sample investigated for thermal neutron activation might be characterized by other elemental compositions. In particular, trace element and hydrogen concentrations could be considerably different from those used for DS02 reference soil. As an example it was demonstrated that in a granite gravestone thermal neutron activation of 36 Cl close to the surface might be, in the worst case, reduced by some 30%, due to increased local granite concentration in this type of environment. Beside other parameters such as, for example, individual sample geometry, the variability of trace elements in soil might be one reason for the variability that is observed in the individual thermal neutron activation measurements (Gold 1995). It is necessary, therefore, to carefully model the exposure geometry of the exposed material, its chemical composition, and the surrounding interface materials in order to obtain the best possible agreement in comparisons between calculated and measured data for thermal neutrons. (author)

TEMPEST-2, Thermalization Program for Neutron Spectra and Multigroup Cross-Sections

International Nuclear Information System (INIS)

Gowins, G.

1984-01-01

Description of problem or function: TEMPEST2 is a neutron thermalization program based upon the Wigner-Wilkins approximation for light moderators and the Wilkins approximation for heavy moderators. A Maxwellian distribution may also be used. The model used may be selected as a function of energy. The second-order differential equations are integrated directly rather than transformed to the Riccati equation. The program provides microscopic and macroscopic cross-section averages over the thermal neutron spectrum

New development of neutron radiography with a small cyclotron

International Nuclear Information System (INIS)

Ikeda, Yasushi; Ohkubo, Kohei; Kato, Toshihiko; Nakamura, Tomihisa; Fuji, Takayoshi.

1990-01-01

A series of neutron radiography testing has been performed for several years by using a small accelerator called 'Baby Cyclotron' manufactured by Japan Steel Works, Ltd. The Baby Cyclotron produces fast neutrons at the rate of 4x10 12 n/cm 2 s, and enables to perform neutron radiography imaging by various techniques. The most important application of this Baby Cyclotron radiography system is the non-destructive testing (NDT) of various explosive devices prepared for space launch vehicles. It is assured that thermal neutron radiography testing is a very useful means for the NDT. Also fast neutron radiography testing is in progress. The fast neutron radiography with a CR39 track-etch image recorder was developed, and it was shown to be the very useful NDT means when the thicker objects used for new H-2 launch vehicles had to be examined. Because thermal neutron radiography has the high detectability of hydrogenous materials, organic elastomers such as O-rings and explosive powder are clearly observable through the opaque steel walls of containers. The Baby Cyclotron and the neutron radiography facility, thermal neutron and fast neutron radiography testings and so on are reported. (K.I.)

MINER - A Mobile Imager of Neutrons for Emergency Responders

Energy Technology Data Exchange (ETDEWEB)

Goldsmith, John E. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, James S. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark D [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kiff, Scott D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mascarenhas, Nicholas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Van De Vreugde, James L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-10-01

We have developed a mobile fast neutron imaging platform to enhance the capabilities of emergency responders in the localization and characterization of special nuclear material. This mobile imager of neutrons for emergency responders (MINER) is based on the Neutron Scatter Camera, a large segmented imaging system that was optimized for large-area search applications. Due to the reduced size and power requirements of a man-portable system, MINER has been engineered to fit a much smaller form factor, and to be operated from either a battery or AC power. We chose a design that enabled omnidirectional (4π) imaging, with only a ~twofold decrease in sensitivity compared to the much larger neutron scatter cameras. The system was designed to optimize its performance for neutron imaging and spectroscopy, but it does also function as a Compton camera for gamma imaging. This document outlines the project activities, broadly characterized as system development, laboratory measurements, and deployments, and presents sample results in these areas. Additional information can be found in the documents that reside in WebPMIS.

MCFT: a program for calculating fast and thermal neutron multigroup constants

International Nuclear Information System (INIS)

Yang Shunhai; Sang Xinzeng

1993-01-01

MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825

Thermal conductivity of beryllium under low temperature high dose neutron irradiation

International Nuclear Information System (INIS)

Chakin, V.P.; Latypov, R.N.; Suslov, D.N.; Kupriyanov, I.B.

2004-01-01

Thermal conductivity of compact beryllium of several Russian grades such as TE-400, TE-56, TE-30, TIP and DIP differing in the production technology, grain size and impurity content has been investigated. The thermal diffusivity of beryllium was measured on the disks in the initial and irradiated conditions using the pulse method in the range from room temperature to 200degC. The thermal conductivity was calculated using the table values for the beryllium thermal capacity. The specimens and beryllium neutron source fragments were irradiation in the SM reactor at 70degC and 200degC to a neutron fluence of (0.5-11.4)·10 22 cm -2 (E>0.1 MeV) and in the BOR-60 reactor at 400degC to 16·10 22 cm -2 (E>0.1MeV), respectively. The low-temperature irradiation leads to the drop decrease of the beryllium thermal conductivity and the effect depends on the irradiation parameters. The paper analyses the effect of irradiation parameters (temperature, neutron fluence), measurement temperature and structural factors on beryllium conductivity. The experiments have revealed that the short time post-irradiation annealing at high temperature results in partial reduction of the thermal conductivity of irradiated beryllium. (author)

Studies on thermal neutron perturbation factor needed for bulk sample activation analysis

CERN Document Server

Csikai, J; Sanami, T; Michikawa, T

2002-01-01

The spatial distribution of thermal neutrons produced by an Am-Be source in a graphite pile was measured via the activation foil method. The results obtained agree well with calculated data using the MCNP-4B code. A previous method used for the determination of the average neutron flux within thin absorbing samples has been improved and extended for a graphite moderator. A procedure developed for the determination of the flux perturbation factor renders the thermal neutron activation analysis of bulky samples of unknown composition possible both in hydrogenous and graphite moderators.

RBE of thermal neutrons for induction of chromosome aberrations in human lymphocytes.

Science.gov (United States)

Schmid, E; Wagner, F M; Canella, L; Romm, H; Schmid, T E

2013-03-01

The induction of chromosome aberrations in human lymphocytes irradiated in vitro with slow neutrons was examined to assess the maximum low-dose RBE (RBE(M)) relative to (60)Co γ-rays. For the blood irradiations, cold neutron beam available at the prompt gamma activation analysis facility at the Munich research reactor FRM II was used. The given flux of cold neutrons can be converted into a thermally equivalent one. Since blood was taken from the same donor whose blood had been used for previous irradiation experiments using widely varying neutron energies, the greatest possible accuracy was available for such an estimation of the RBE(M) avoiding the inter-individual variations or differences in methodology usually associated with inter-laboratory comparisons. The magnitude of the coefficient α of the linear dose-response relationship (α = 0.400 ± 0.018 Gy(-1)) and the derived RBE(M) of 36.4 ± 13.3 obtained for the production of dicentrics by thermal neutrons confirm our earlier observations of a strong decrease in α and RBE(M) with decreasing neutron energy lower than 0.385 MeV (RBE(M) = 94.4 ± 38.9). The magnitude of the presently estimated RBE(M) of thermal neutrons is-with some restrictions-not significantly different to previously reported RBE(M) values of two laboratories.

Design of 6 Mev linear accelerator based pulsed thermal neutron source: FLUKA simulation and experiment

Energy Technology Data Exchange (ETDEWEB)

Patil, B.J., E-mail: bjp@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India)

2012-01-15

The 6 MeV LINAC based pulsed thermal neutron source has been designed for bulk materials analysis. The design was optimized by varying different parameters of the target and materials for each region using FLUKA code. The optimized design of thermal neutron source gives flux of 3 Multiplication-Sign 10{sup 6}ncm{sup -2}s{sup -1} with more than 80% of thermal neutrons and neutron to gamma ratio was 1 Multiplication-Sign 10{sup 4}ncm{sup -2}mR{sup -1}. The results of prototype experiment and simulation are found to be in good agreement with each other. - Highlights: Black-Right-Pointing-Pointer The optimized 6 eV linear accelerator based thermal neutron source using FLUKA simulation. Black-Right-Pointing-Pointer Beryllium as a photonuclear target and reflector, polyethylene as a filter and shield, graphite as a moderator. Black-Right-Pointing-Pointer Optimized pulsed thermal neutron source gives neutron flux of 3 Multiplication-Sign 10{sup 6} n cm{sup -2} s{sup -1}. Black-Right-Pointing-Pointer Results of the prototype experiment were compared with simulations and are found to be in good agreement. Black-Right-Pointing-Pointer This source can effectively be used for the study of bulk material analysis and activation products.

Neutron moderation theory with thermal motion of the moderator nuclei

Energy Technology Data Exchange (ETDEWEB)

Rusov, V.D.; Tarasov, V.A.; Chernezhenko, S.A.; Kakaev, A.A.; Smolyar, V.P. [Odessa National Polytechnic University, Department of Theoretical and Experimental Nuclear Physics, Odessa (Ukraine)

2017-09-15

In this paper we present the analytical expression for the neutron scattering law for an isotropic source of neutrons, obtained within the framework of the gas model with the temperature of the moderating medium as a parameter. The obtained scattering law is based on the solution of the general kinematic problem of elastic scattering of neutrons on nuclei in the L-system. Both the neutron and the nucleus possess arbitrary velocities in the L-system. For the new scattering law we obtain the flux densities and neutron moderation spectra as functions of temperature for the reactor fissile medium. The expressions for the moderating neutrons spectra allow reinterpreting the physical nature of the underlying processes in the thermal region. (orig.)

Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions

International Nuclear Information System (INIS)

Santiago, Daniela Maiolino Norberto

2011-01-01

In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

International Nuclear Information System (INIS)

Hawari, Ayman

2014-01-01

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

Thermal neutron spectrum distribution in TRIGA fuels

International Nuclear Information System (INIS)

Gui Ah Auu; Harasawa, Susumu; An, Shigehiro

1989-01-01

The dependence of thermal neutron spectrum in TRIGA fuel cell on fuel temperature and TRIGA fuel types were studied using LIBP and THERMOS codes. Some characteristics of the TRIGA fuel including its prompt negative temperature coefficient of reactivity were explained using the results of the study. (author)

Instrumentation to handle thermal polarized neutron beams

NARCIS (Netherlands)

Kraan, W.H.

2004-01-01

In this thesis we investigate devices needed to handle the polarization of thermal neutron beams: Ï/2-flippers (to start/stop Larmor precession) and Ï-flippers (to reverse polarization/precession direction) and illustrate how these devices are used to investigate the properties of matter and of the

Thermal neutron scattering cross sections of beryllium and magnesium oxides

International Nuclear Information System (INIS)

Al-Qasir, Iyad; Jisrawi, Najeh; Gillette, Victor; Qteish, Abdallah

2016-01-01

Highlights: • Neutron thermalization in BeO and MgO was studied using Ab initio lattice dynamics. • The BeO phonon density of states used to generate the current ENDF library has issues. • The BeO cross sections can provide a more accurate ENDF library than the current one. • For MgO an ENDF library is lacking: a new accurate one can be built from our results. • BeO is a better filter than MgO, especially when cooled down to 77 K. - Abstract: Alkaline-earth beryllium and magnesium oxides are fundamental materials in nuclear industry and thermal neutron scattering applications. The calculation of the thermal neutron scattering cross sections requires a detailed knowledge of the lattice dynamics of the scattering medium. The vibrational properties of BeO and MgO are studied using first-principles calculations within the frame work of the density functional perturbation theory. Excellent agreement between the calculated phonon dispersion relations and the experimental data have been obtained. The phonon densities of states are utilized to calculate the scattering laws using the incoherent approximation. For BeO, there are concerns about the accuracy of the phonon density of states used to generate the current ENDF/B-VII.1 libraries. These concerns are identified, and their influences on the scattering law and inelastic scattering cross section are analyzed. For MgO, no up to date thermal neutron scattering cross section ENDF library is available, and our results represent a potential one for use in different applications. Moreover, the BeO and MgO efficiencies as neutron filters at different temperatures are investigated. BeO is found to be a better filter than MgO, especially when cooled down, and cooling MgO below 77 K does not significantly improve the filter’s efficiency.

Computed image analysis of neutron radiographs

International Nuclear Information System (INIS)

Dinca, M.; Anghel, E.; Preda, M.; Pavelescu, M.

2008-01-01

Similar with X-radiography, using neutron like penetrating particle, there is in practice a nondestructive technique named neutron radiology. When the registration of information is done on a film with the help of a conversion foil (with high cross section for neutrons) that emits secondary radiation (β,γ) that creates a latent image, the technique is named neutron radiography. A radiographic industrial film that contains the image of the internal structure of an object, obtained by neutron radiography, must be subsequently analyzed to obtain qualitative and quantitative information about the structural integrity of that object. There is possible to do a computed analysis of a film using a facility with next main components: an illuminator for film, a CCD video camera and a computer (PC) with suitable software. The qualitative analysis intends to put in evidence possibly anomalies of the structure due to manufacturing processes or induced by working processes (for example, the irradiation activity in the case of the nuclear fuel). The quantitative determination is based on measurements of some image parameters: dimensions, optical densities. The illuminator has been built specially to perform this application but can be used for simple visual observation. The illuminated area is 9x40 cm. The frame of the system is a comparer of Abbe Carl Zeiss Jena type, which has been adapted to achieve this application. The video camera assures the capture of image that is stored and processed by computer. A special program SIMAG-NG has been developed at INR Pitesti that beside of the program SMTV II of the special acquisition module SM 5010 can analyze the images of a film. The major application of the system was the quantitative analysis of a film that contains the images of some nuclear fuel pins beside a dimensional standard. The system was used to measure the length of the pellets of the TRIGA nuclear fuel. (authors)

Measurement of two-dimensional thermal neutron flux in a water phantom and evaluation of dose distribution characteristics

International Nuclear Information System (INIS)

Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Torii, Yoshiya; Horiguchi, Yoji

2001-03-01

To evaluate nitrogen dose, boron dose and gamma-ray dose occurred by neutron capture reaction of the hydrogen at the medical irradiation, two-dimensional distribution of the thermal neutron flux is very important because these doses are proportional to the thermal neutron distribution. This report describes the measurement of the two-dimensional thermal neutron distribution in a head water phantom by neutron beams of the JRR-4 and evaluation of the dose distribution characteristic. Thermal neutron flux in the phantom was measured by gold wire placed in the spokewise of every 30 degrees in order to avoid the interaction. Distribution of the thermal neutron flux was also calculated using two-dimensional Lagrange's interpolation program (radius, angle direction) developed this time. As a result of the analysis, it was confirmed to become distorted distribution which has annular peak at outside of the void, though improved dose profile of the deep direction was confirmed in the case which the radiation field in the phantom contains void. (author)

Thermal neutron dose calculation in synovium membrane for BNCS

International Nuclear Information System (INIS)

Abdalla, Khalid; Naqvi, A.A.; Maalej, N.; El-Shahat, B.

2006-01-01

A D(d,n) reaction based setup has been optimized for Boron Neutron Capture Synovectomy (BNCS). The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield. The neutron dose was calculated at various depths in a knee phantom loaded with boron to determine therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values. (author)

Influence of orientation averaging on the anisotropy of thermal neutrons scattering on water molecules

International Nuclear Information System (INIS)

Markovic, M. I.; Radunovic, J. B.

1976-01-01

Determination of spatial distribution of neutron flux in water, most frequently used moderator in thermal reactors, demands microscopic scattering kernels dependence on cosine of thermal neutrons scattering angle when solving the Boltzmann equation. Since spatial orientation of water molecules influences this dependence it is necessary to perform orientation averaging or rotation-vibrational intermediate scattering function for water molecules. The calculations described in this paper and the obtained results showed that methods of orientation averaging do not influence the anisotropy of thermal neutrons scattering on water molecules, but do influence the inelastic scattering

Numerical simulation for neutron pinhole imaging in ICF

International Nuclear Information System (INIS)

Chen Faxin; Yang Jianlun; Wen Shuhuai

2005-01-01

Pinhole imaging of the neutron production in laser-driven inertial confinement fusion experiments can provide important information about performance of various capsule designs. In order to get good results in experiments, it is needed to judge performance of various pinhole designs qualitatively or quantitatively before experiment. Calculation of imaging can be simply separated into pinhole imaging and image spectral analysis. In this paper, pinhole imaging is discussed, codes for neutron pinhole imaging and image showing is programed. The codes can be used to provide theoretical foundation for pinhole designing and simulating data for image analysing. (authors)

Miscellaneous neutron techniques

International Nuclear Information System (INIS)

Iddings, F.A.

1976-01-01

Attention is brought to the less often uses of neutrons in the areas of neutron radiography, well logging, and neutron gaging. Emphasis on neutron radiography points toward the isotopic sensitivity of the method versus the classical bulk applications. Also recognized is the ability of neutron radiography to produce image changes that correspond to thickness and density changes obtained in photon radiography. Similarly, neutron gaging applications center on the measurement of radiography. Similarly, neutron gaging applications center on the measurement of water, oil, or plastics in industrial samples. Well logging extends the neutron gaging to encompass many neutron properties and reactions besides thermalization and capture. Neutron gaging also gives information on organic structure and concentrations of a variety of elements or specific compounds in selected matrices

Accounting for the thermal neutron flux depression in voluminous samples for instrumental neutron activation analysis

International Nuclear Information System (INIS)

Overwater, R.M.W.; Hoogenboom, J.E.

1994-01-01

At the Delft University of Technology Interfaculty Reactor Institute, a facility has been installed to irradiate cylindrical samples with diameters up to 15 cm and weights up to 50 kg for instrumental neutron activation analysis (INAA) purposes. To be able to do quantitative INAA on voluminous samples, it is necessary to correct for gamma-ray absorption, gamma-ray scattering, neutron absorption, and neutron scattering in the sample. The neutron absorption and the neutron scattering are discussed. An analytical solution is obtained for the diffusion equation in the geometry of the irradiation facility. For samples with known composition, the neutron flux--as a function of position in the sample--can be calculated directly. Those of unknown composition require additional flux measurements on which least-squares fitting must be done to obtain both the thermal neutron diffusion coefficient D s and the diffusion length L s of the sample. Experiments are performed to test the theory

Experimental investigation of thermal neutron analysis based landmine detection technology

International Nuclear Information System (INIS)

Zeng Jun; Chu Chengsheng; Ding Ge; Xiang Qingpei; Hao Fanhua; Luo Xiaobing

2013-01-01

Background: Recently, the prompt gamma-rays neutron activation analysis method is wildly used in coal analysis and explosive detection, however there were less application about landmine detection using neutron method especially in the domestic research. Purpose: In order to verify the feasibility of Thermal Neutron Analysis (TNA) method used in landmine detection, and explore the characteristic of this technology. Methods: An experimental system of TNA landmine detection was built based on LaBr 3 (Ce) fast scintillator detector and 252 Cf isotope neutron source. The system is comprised of the thermal neutron transition system, the shield system, and the detector system. Results: On the basis of the TNA, the wide energy area calibration method especially to the high energy area was investigated, and the least detection time for a typical mine was defined. In this study, the 72-type anti-tank mine, the 500 g TNT sample and several interferential objects are tested in loess, red soil, magnetic soil and sand respectively. Conclusions: The experimental results indicate that TNA is a reliable demining method, and it can be used to confirm the existence of Anti-Tank Mines (ATM) and large Anti-Personnel Mines (APM) in complicated condition. (authors)

Neutron diffractometer for bio-crystallography (BIX) with an imaging plate neutron detector

Energy Technology Data Exchange (ETDEWEB)

Niimura, Nobuo [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan)

1994-12-31

We have constructed a dedicated diffractometer for neutron crystallography in biology (BIX) on the JRR-3M reactor at JAERI (Japan Atomic Energy Research Institute). The diffraction intensity from a protein crystal is weaker than that from most inorganic materials. In order to overcome the intensity problem, an elastically bent silicon monochromator and a large area detector system were specially designed. A preliminary result of diffraction experiment using BIX has been reported. An imaging plate neutron detector has been developed and a feasibility experiment was carried out on BIX. Results are reported. An imaging plate neutron detector has been developed and a feasibility test was carried out using BIX.

Characterization of the new neutron imaging and materials science facility IMAT

Science.gov (United States)

Minniti, Triestino; Watanabe, Kenichi; Burca, Genoveva; Pooley, Daniel E.; Kockelmann, Winfried

2018-04-01

IMAT is a new cold neutron imaging and diffraction instrument located at the second target station of the pulsed neutron spallation source ISIS, UK. A broad range of materials science and materials testing areas will be covered by IMAT. We present the characterization of the imaging part, including the energy-selective and energy-dispersive imaging options, and provide the basic parameters of the radiography and tomography instrument. In particular, detailed studies on mono and bi-dimensional neutron beam flux profiles, neutron flux as a function of the neutron wavelength, spatial and energy dependent neutron beam uniformities, guide artifacts, divergence and spatial resolution, and neutron pulse widths are provided. An accurate characterization of the neutron beam at the sample position, located 56 m from the source, is required to optimize collection of radiographic and tomographic data sets and for performing energy-dispersive neutron imaging via time-of-flight methods in particular.

New feature of the neutron color image intensifier

Science.gov (United States)

Nittoh, Koichi; Konagai, Chikara; Noji, Takashi; Miyabe, Keisuke

2009-06-01

We developed prototype neutron color image intensifiers with high-sensitivity, wide dynamic range and long-life characteristics. In the prototype intensifier (Gd-Type 1), a terbium-activated Gd 2O 2S is used as the input-screen phosphor. In the upgraded model (Gd-Type 2), Gd 2O 3 and CsI:Na are vacuum deposited to form the phosphor layer, which improved the sensitivity and the spatial uniformity. A europium-activated Y 2O 2S multi-color scintillator, emitting red, green and blue photons with different intensities, is utilized as the output screen of the intensifier. By combining this image intensifier with a suitably tuned high-sensitive color CCD camera, higher sensitivity and wider dynamic range could be simultaneously attained than that of the conventional P20-phosphor-type image intensifier. The results of experiments at the JRR-3M neutron radiography irradiation port (flux: 1.5×10 8 n/cm 2/s) showed that these neutron color image intensifiers can clearly image dynamic phenomena with a 30 frame/s video picture. It is expected that the color image intensifier will be used as a new two-dimensional neutron sensor in new application fields.

New feature of the neutron color image intensifier

International Nuclear Information System (INIS)

Nittoh, Koichi; Konagai, Chikara; Noji, Takashi; Miyabe, Keisuke

2009-01-01

We developed prototype neutron color image intensifiers with high-sensitivity, wide dynamic range and long-life characteristics. In the prototype intensifier (Gd-Type 1), a terbium-activated Gd 2 O 2 S is used as the input-screen phosphor. In the upgraded model (Gd-Type 2), Gd 2 O 3 and CsI:Na are vacuum deposited to form the phosphor layer, which improved the sensitivity and the spatial uniformity. A europium-activated Y 2 O 2 S multi-color scintillator, emitting red, green and blue photons with different intensities, is utilized as the output screen of the intensifier. By combining this image intensifier with a suitably tuned high-sensitive color CCD camera, higher sensitivity and wider dynamic range could be simultaneously attained than that of the conventional P20-phosphor-type image intensifier. The results of experiments at the JRR-3M neutron radiography irradiation port (flux: 1.5x10 8 n/cm 2 /s) showed that these neutron color image intensifiers can clearly image dynamic phenomena with a 30 frame/s video picture. It is expected that the color image intensifier will be used as a new two-dimensional neutron sensor in new application fields.

The neutron imaging system fielded at the National Ignition Facility

Directory of Open Access Journals (Sweden)

Fittinghoff D.N.

2013-11-01

Full Text Available We have fielded a neutron imaging system at the National Ignition Facility to collect images of fusion neutrons produced in the implosion of inertial confinement fusion experiments and scattered neutrons from (n, n′ reactions of the source neutrons in the surrounding dense material. A description of the neutron imaging system is presented, including the pinhole array aperture, the line-of-sight collimation, the scintillator-based detection system and the alignment systems and methods. Discussion of the alignment and resolution of the system is presented. We also discuss future improvements to the system hardware.

Microscopic cross-section measurements by thermal neutron activation

International Nuclear Information System (INIS)

Avila L, J.

1987-08-01

Microscopic cross sections measured by thermal neutron activation using RP-0 reactor at the Peruvian Nuclear Energy Institute. The method consists in measuring microscopic cross section ratios through activated samples, requiring being corrected in thermal and epithermal energetic range by Westcott formalism. Furthermore, the comptage ratios measured for each photopeak to its decay fraction should be normalized from interrelation between both processes above, activation microscopic cross sections are obtained

Transportable, Low-Dose Active Fast-Neutron Imaging

Energy Technology Data Exchange (ETDEWEB)

Mihalczo, John T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wright, Michael C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McConchie, Seth M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Palles, Blake A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.

Detection of plastic explosives using thermal neutron radiography

International Nuclear Information System (INIS)

Hacidume, Leo Ryoske

1999-12-01

The work aims to demonstrate the potentiality of the neutron radiography technique, allied to the computerized tomography by transmission, to both detect and visualize plastic explosive samples in several hidden conditions, using a simple scanner as a digitalisation instrument. Each tomographic essay was obtained in the J-9 channel of the Argonauta Research Reactor of IEN/CNEN, in groups of six neutron radiographic projections, performed with an angular increment of 30 deg C, in a period of time of 30 minutes for each projection. Two groups of tomographic reconstructions were generated, distinguished by the digitalisation process of the interested lines in the reconstruction plane coming from the projection groups, utilization a scanner and a microdensitometer, respectively. The reconstruction of the bi-dimensional image of the transverse section, in relation to this plane, was processed making use of the Image Reconstruction Algorithmic of an Image based on the Maximum Entropy principle (ARIEM). From the qualitative analysis of the images, we conclude that the neutron radiographic system was able to detect the explosive sample in a satisfactory way while the quantitative analysis confirmed the application effectiveness of a scanner to acquire the projection dates whose objective is only a reconnaissance. (author)

Production of a faithful realistic phantom to human head and thermal neutron flux measurement on the brain surface. Cooperative research

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Torii, Yoshiya; Uchiyama, Junzo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Endo, Kiyoshi; Yamamoto, Tetsuya; Matsumura, Akira; Nose, Tadao [Tsukuba Univ., Tsukuba, Ibaraki (Japan)

2002-12-01

Thermal neutron flux is determined using the gold wires in current BNCT irradiation, so evaluation of arbitrary points after the irradiation is limited in the quantity of these detectors. In order to make up for the weakness, dose estimation of a patient is simulated by a computational dose calculation supporting system. In another way without computer simulation, a medical irradiation condition can be replicate experimentally using of realistic phantom which was produced from CT images by rapid prototyping technique. This phantom was irradiated at a same JRR-4 neutron beam as clinical irradiation condition of the patient and the thermal neutron distribution on the brain surface was measured in detail. This experimental evaluation technique using a realistic phantom is applicable to in vitro cell irradiation experiments for radiation biological effects as well as in-phantom experiments for dosimetry under the nearly medical irradiation condition of patient. (author)

Production of a faithful realistic phantom to human head and thermal neutron flux measurement on the brain surface. Cooperative research

CERN Document Server

Yamamoto, K; Kishi, T; Kumada, H; Matsumura, A; Nose, T; Torii, Y; Uchiyama, J; Yamamoto, T

2002-01-01

Thermal neutron flux is determined using the gold wires in current BNCT irradiation, so evaluation of arbitrary points after the irradiation is limited in the quantity of these detectors. In order to make up for the weakness, dose estimation of a patient is simulated by a computational dose calculation supporting system. In another way without computer simulation, a medical irradiation condition can be replicate experimentally using of realistic phantom which was produced from CT images by rapid prototyping technique. This phantom was irradiated at a same JRR-4 neutron beam as clinical irradiation condition of the patient and the thermal neutron distribution on the brain surface was measured in detail. This experimental evaluation technique using a realistic phantom is applicable to in vitro cell irradiation experiments for radiation biological effects as well as in-phantom experiments for dosimetry under the nearly medical irradiation condition of patient.

Thermal neutron inelastic scattering and it's application to the material science

International Nuclear Information System (INIS)

Li Zhuqi

1986-01-01

A brief description of the elementary scattering theory of the interaction between the thermal neutrons and the condensed matter is given and the characteristics related to the experimental method of the thermal neutrons inelastic scattering is described. Expressions of the phonons dispersion, density of the phonon state and the self-diffusion coefficient at the some conditions are also introduced. Some examples of describing diagram of the phonon dispersion, density of the phonons state and selfdiffusion coefficient measured by different authors are given

Parity non-conservation in the capture of polarized thermal neutrons

DEFF Research Database (Denmark)

Warming, Inge Elisabeth

1969-01-01

The asymmetry in the intensity of γ-radiation following the capture of polarized thermal neutrons in 113Cd has been measured with Ge(Li) detectors. The result, A = (−0.6±1.8)×10−4, like that previously reported [1], gives no evidence for a non-zero effect.......The asymmetry in the intensity of γ-radiation following the capture of polarized thermal neutrons in 113Cd has been measured with Ge(Li) detectors. The result, A = (−0.6±1.8)×10−4, like that previously reported [1], gives no evidence for a non-zero effect....

Absolute efficiency calibration of 6LiF-based solid state thermal neutron detectors

Science.gov (United States)

Finocchiaro, Paolo; Cosentino, Luigi; Lo Meo, Sergio; Nolte, Ralf; Radeck, Desiree

2018-03-01

The demand for new thermal neutron detectors as an alternative to 3He tubes in research, industrial, safety and homeland security applications, is growing. These needs have triggered research and development activities about new generations of thermal neutron detectors, characterized by reasonable efficiency and gamma rejection comparable to 3He tubes. In this paper we show the state of the art of a promising low-cost technique, based on commercial solid state silicon detectors coupled with thin neutron converter layers of 6LiF deposited onto carbon fiber substrates. A few configurations were studied with the GEANT4 simulation code, and the intrinsic efficiency of the corresponding detectors was calibrated at the PTB Thermal Neutron Calibration Facility. The results show that the measured intrinsic detection efficiency is well reproduced by the simulations, therefore validating the simulation tool in view of new designs. These neutron detectors have also been tested at neutron beam facilities like ISIS (Rutherford Appleton Laboratory, UK) and n_TOF (CERN) where a few samples are already in operation for beam flux and 2D profile measurements. Forthcoming applications are foreseen for the online monitoring of spent nuclear fuel casks in interim storage sites.

In situ diagnostics of the crystal-growth process through neutron imaging

DEFF Research Database (Denmark)

Tremsin, Anton S.; Makowska, Malgorzata Grazyna; Perrodin, Didier

2016-01-01

Neutrons are known to be unique probes in situations where other types of radiation fail to penetrate samples and their surrounding structures. In this paper it is demonstrated how thermal and cold neutron radiography can provide time-resolved imaging of materials while they are being processed (e......, as limited by the resolution of the present experiments). It is also demonstrated that the dopant concentration can be quantified even for very low concentration levels (∼ 0.1%) in 10 mm thick samples. The interface between the solid and liquid phases can also be imaged, provided there is a sufficient change.......g. while growing single crystals). The processing equipment, in this case furnaces, and the scintillator materials are opaque to conventional X-ray interrogation techniques. The distribution of the europium activator within a BaBrCl:Eu scintillator (0.1 and 0.5% nominal doping concentrations per mole...

Thermal neutron capture and resonance integral cross sections of {sup 45}Sc