The design of a position-sensitive thermal-neutron detector

International Nuclear Information System (INIS)

Zhang Yi; Chen Ziyu; Shen Ji

2007-01-01

We design a type of position-sensitive thermal-neutron detector. The design is based on the nuclear reaction 10 B(n, α) 7 Li, and solid boron-10 is used as the target material while the alpha and lithium-7 particles from the reaction are caught as the source of position information of the original neutrons. With the help of MCNP software, we simulate the distribution of alpha particles in the boron target, which leads to the optimal thickness of target, physical efficiency and position resolution. (authors)

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

Microstructured silicon neutron detectors for security applications

International Nuclear Information System (INIS)

Esteban, S; Fleta, C; Jumilla, C; Pellegrini, G; Quirion, D; Rodriguez, J; Lozano, M; Guardiola, C

2014-01-01

In this paper we present the design and performance of a perforated thermal neutron silicon detector with a 6 LiF neutron converter. This device was manufactured within the REWARD project workplace whose aim is to develop and enhance technologies for the detection of nuclear and radiological materials. The sensor perforated structure results in a higher efficiency than that obtained with an equivalent planar sensor. The detectors were tested in a thermal neutron beam at the nuclear reactor at the Instituto Superior Técnico in Lisbon and the intrinsic detection efficiency for thermal neutrons and the gamma sensitivity were obtained. The Geant4 Monte Carlo code was used to simulate the experimental conditions, i.e. thermal neutron beam and the whole detector geometry. An intrinsic thermal neutron detection efficiency of 8.6%±0.4% with a discrimination setting of 450 keV was measured

A neutron activation detector

International Nuclear Information System (INIS)

Ambardanishvili, T.S.; Kolomiitsev, M.A.; Zakharina, T.Y.; Dundua, V.J.; Chikhladze, N.V.

1973-01-01

The present invention concerns a neutron activation detector made from a moulded and hardened composition. According to the invention, that composition contains an activable substance constituted by at least two chemical elements and/or compounds of at least two chemical elements. Each of these chemical elements is capable of reacting with the neutrons forming radio-active isotopes with vatious levels of energy during desintegration. This neutron detector is mainly suitable for measuring integral thermal neutron and fast neutron fluxes during irradiation of the sample, and also for measuring the intensities of neutron fields [fr

Neutron detector

Science.gov (United States)

Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN

2011-04-05

A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.

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.

Simulation and optimisation of a position sensitive scintillation detector with wavelength shifting fibers for thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Herzkamp, Matthias; Engels, Ralf; Kemmerling, Guenter [ZEA-2, Forschungszentrum Juelich (Germany); Brueckel, Thomas [JCNS, Forschungszentrum Juelich (Germany); Stahl, Achim [III. Physikalisches Institut B, RWTH Aachen (Germany); Waasen, Stefan van [ZEA-2, Forschungszentrum Juelich (Germany); Faculty of Engineering, University of Duisburg-Essen (Germany)

2015-07-01

In neutron scattering experiments it is important to have position sensitive large scale detectors for thermal neutrons. A detector based on a neutron scintillator with wave length shifting fibers is a new kind of such a detector. We present the simulation of the detector based on the microscopic structure of the scintillation material of the mentioned detector. It consists of a converter and a scintillation powder bound in a matrix. The converter in our case is lithium fluoride with enriched lithium 6, to convert thermal neutrons into high energetic alpha and triton particles. The scintillation material is silver doped zinc sulfide. We show that pulse height spectra obtained by these scintillators can be be explained by the simple model of randomly distributed spheres of zinc sulfide and lithium fluoride. With this model, it is possible to optimise the mass ratio of zinc sulfide to lithium fluoride with respect to detection efficiency and/or energy deposition in zinc sulfide.

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)

Properties of the lithium carbonate for to be used as thermal neutrons detector

International Nuclear Information System (INIS)

Herrera A, E.; Urena N, F.

2003-01-01

In this work the dosimetric properties of the lithium carbonate used as detecting of thermal neutrons and by means of free radicals is evaluated and presented. The studied parameters that were carried out for this detector were: intensity of the Electron paramagnetic resonance signal (EPR); reproducibility, fading of the signal to ambient temperature, stability of the signal to low temperature (0 degrees); answer of zero dose and homogeneity or reliability of the data of the detector, humidity, solar light, temperature and radio sensitivity. These parameters indicate the utility that have the detectors for the estimation of fields of neutron fluences that are applicable to capture therapies by neutron-boron and, nuclear reactors. (Author)

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)

Analytical modeling of thin film neutron converters and its application to thermal neutron gas detectors

Energy Technology Data Exchange (ETDEWEB)

Piscitelli, F; Esch, P Van, E-mail: piscitelli@ill.fr [Institut Laue-Langevin (ILL), 6, Jules Horowitz, 38042 Grenoble (France)

2013-04-15

A simple model is explored mainly analytically to calculate and understand the PHS of single and multi-layer thermal neutron detectors and to help optimize the design in different circumstances. Several theorems are deduced that can help guide the design.

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)

Dosimeter incorporating radiophotoluminescent detectors for thermal neutrons and γ-rays in n-γ fields

Energy Technology Data Exchange (ETDEWEB)

Salem, Y.O. [Groupe RaMsEs, Institut Pluridisciplinaire Hubert Curien (IPHC), UMR 7178 CNRS/IN2P3, 23 rue du Loess, BP 28, F-67037 Strasbourg Cedex 2 (France); Nachab, A., E-mail: a.nachab@uca.ma [Département de physique, Faculté Poly-disciplinaire, Université Cadi Ayyad, Route Sidi Bouzid BP 4162, 46000 Safi (Morocco); Roy, C.; Nourreddine, A. [Groupe RaMsEs, Institut Pluridisciplinaire Hubert Curien (IPHC), UMR 7178 CNRS/IN2P3, 23 rue du Loess, BP 28, F-67037 Strasbourg Cedex 2 (France)

2016-10-15

We have developed a dosimeter associating different neutron converters with two radiophotoluminescent detectors to measure thermal neutrons and γ-rays in a mixed n-γ field. Tests show that the H{sup ∗}(10) and H{sub p}(10) responses to thermal neutrons and γ-rays are linear with detection limits lower than 0.4 mSv. The angular dependence of the dosimeter response is satisfactory and the influence of a phantom on the results is examined.

Measure of thermal neutron flux in the IPEN/MB-01 reactor using 197 Au wire activation detectors

International Nuclear Information System (INIS)

Marques, Andre Luis Ferreira

1995-01-01

This dissertation has aimed at developing a neutron flux measurement technique by means of detectors activation analysis. The main task of this work was the implementation of this thermal neutron flux measurement technique, using gold wires as activation detectors in the IPEN/MB-01 reactor core. The neutron thermal flux spatial distribution was obtained by gold wire activation technique, with wire diameters of 0.125 mm and 0.250 mm in seven selected reactor experimental channels. The values of thermal flux were about 10 9 neutrons/cm 2 .s. This experiment has been the first one conducted with gold wires in the IPEN/MB-01 reactor, being this technique implemented for use by experiments in flux mapping of the core

Detection mechanisms in silicon diodes used as α-particle and thermal neutron detectors

International Nuclear Information System (INIS)

Cerofolini, G.F.; Ferla, G.; Foglio Para, A.

1981-01-01

Some common silicon devices (diodes, RAMs etc.) can be used as α and thermal neutron detectors. An α resolution of approx. equal to 3% can be obtained utilizing p + /n or n + /p diodes with no external bias. Thermal neutrons are detected by means of the reaction 10 B(n,α) 7 Li on the 10 B present in the devices. Neutron efficiency has been substantially improved by implantation of 10 B ions in the p + region of the diodes. Experimental results allow us to clarify the carrier collection mechanisms throughout the device. Some current opinions in the field are contradicted. (orig.)

Design and fabrication of 4π Clover Detector Array Assembly for gamma-spectroscopy studies using thermal neutrons

International Nuclear Information System (INIS)

Kumar, Manish; Kamble, S.R.; Chaudhari, A.T.; Sabharwal, T.P.; Pathak, Kavindra; Prasad, N.K.; Kinage, L.A.; Biswas, D.C.; Bhagwat, P.V.

2017-01-01

Nuclear spectroscopy has been studied earlier from the measurement of prompt gamma rays produced in reactions with thermal neutrons from CIRUS reactor. For studying the prompt γ-spectroscopy using thermal neutrons from Dhruva Reactor, BARC, the development of a dedicated beam line (R-3001) is in progress. In this beam line a detector assembly consisting of Clover Ge detectors will be used. This experimental setup will be utilized to investigate nuclear structure using prompt (n,γ) reactions and also to study the spectroscopy of neutron-rich fission-fragment nuclei

Study of a high spatial resolution {sup 10}B-based thermal neutron detector for application in neutron reflectometry: the Multi-Blade prototype

Energy Technology Data Exchange (ETDEWEB)

Piscitelli, F; Buffet, J C; Clergeau, J F; Cuccaro, S; Guérard, B; Khaplanov, A; Manna, Q La; Rigal, J M; Esch, P Van, E-mail: piscitelli@ill.fr [Institut Laue-Langevin (ILL), 6, Jules Horowitz, 38042, Grenoble (France)

2014-03-01

Although for large area detectors it is crucial to find an alternative to detect thermal neutrons because of the {sup 3}He shortage, this is not the case for small area detectors. Neutron scattering science is still growing its instruments' power and the neutron flux a detector must tolerate is increasing. For small area detectors the main effort is to expand the detectors' performances. At Institut Laue-Langevin (ILL) we developed the Multi-Blade detector which wants to increase the spatial resolution of {sup 3}He-based detectors for high flux applications. We developed a high spatial resolution prototype suitable for neutron reflectometry instruments. It exploits solid {sup 10}B-films employed in a proportional gas chamber. Two prototypes have been constructed at ILL and the results obtained on our monochromatic test beam line are presented here.

The alanine detector in BNCT dosimetry: dose response in thermal and epithermal neutron fields.

Science.gov (United States)

Schmitz, T; Bassler, N; Blaickner, M; Ziegner, M; Hsiao, M C; Liu, Y H; Koivunoro, H; Auterinen, I; Serén, T; Kotiluoto, P; Palmans, H; Sharpe, P; Langguth, P; Hampel, G

2015-01-01

The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a (60)Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes fluka and mcnp. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen & Olsen alanine response model. The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. The alanine detector can be used without

Properties of the lithium carbonate for to be used as thermal neutrons detector; Propiedades del carbonato de litio para ser usado como detector de neutrones termicos

Energy Technology Data Exchange (ETDEWEB)

Herrera A, E.; Urena N, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

In this work the dosimetric properties of the lithium carbonate used as detecting of thermal neutrons and by means of free radicals is evaluated and presented. The studied parameters that were carried out for this detector were: intensity of the Electron paramagnetic resonance signal (EPR); reproducibility, fading of the signal to ambient temperature, stability of the signal to low temperature (0 degrees); answer of zero dose and homogeneity or reliability of the data of the detector, humidity, solar light, temperature and radio sensitivity. These parameters indicate the utility that have the detectors for the estimation of fields of neutron fluences that are applicable to capture therapies by neutron-boron and, nuclear reactors. (Author)

A Polyethylene Moderator Design for Auxiliary Ex-core Neutron Detector

International Nuclear Information System (INIS)

Lee, Hwan Soo; Shin, Ho Cheol; Bae, Seong Man

2012-01-01

The moderator of detector assembly in ENFMS (Excore Neutron Flux Monitoring System) plays a key role for slowing down from fast neutron to thermal neutron at outside of reactor vessel. Since neutron monitoring detector such as BF3, fission chamber detectors mostly responds to thermal neutron, moderator should be included to neutron detector assembly to detect more efficiently. Generally, resin has been used for moderator of detector in ENFMS of OPR1000 and APR1400, because resin has stable thermal resistance, availability and high neutron moderation characteristics due to the light atomic materials. In case of an auxiliary ex-core neutron detector, the polyethylene is suggested that polyethylene has a better moderator rather than resin, then, the amounts of moderator are reduced. This is important thing for auxiliary ex-core detector equipment at reactor, because the auxiliary equipment should affect minimally to another system. In this study, polyethylene moderator is designed for auxiliary ex-core neutron detector. To find out the optimal thickness of polyethylene moderator, preliminary simulation and experiments are performed. And sensitivity simulation for detector moderator at actual reactor is performed by DORT code

filled neutron detectors

Indian Academy of Sciences (India)

Boron trifluoride (BF3) proportional counters are used as detectors for thermal neutrons. They are characterized by high neutron sensitivity and good gamma discriminating properties. Most practical BF3 counters are filled with pure boron trifluoride gas enriched up to 96% 10B. But BF3 is not an ideal proportional counter ...

Neutron detectors for the ESS diffractometers

Czech Academy of Sciences Publication Activity Database

Stefanescu, I.; Christensen, M.; Fenske, J.; Hall-Wilton, R.; Henry, P. F.; Kirstein, O.; Muller, M.; Nowak, G.; Pooley, D.; Raspino, D.; Rhodes, N.; Šaroun, Jan; Schefer, J.; Schooneveld, E.; Sykora, J.; Schweika, W.

2017-01-01

Roč. 12, JAN (2017), č. článku P01019. ISSN 1748-0221 R&D Projects: GA MŠk LM2015048 Institutional support: RVO:61389005 Keywords : instrumentation for neutron sources * neutron diffraction detectors * neutron detectors (cold, thermal, fast neutrons) Subject RIV: BM - Solid Matter Physics ; Magnetism OBOR OECD: Condensed matter physics (including formerly solid state physics, supercond.) Impact factor: 1.220, year: 2016

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

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

Using a Tandem Pelletron accelerator to produce a thermal neutron beam for detector testing purposes.

Science.gov (United States)

Irazola, L; Praena, J; Fernández, B; Macías, M; Bedogni, R; Terrón, J A; Sánchez-Nieto, B; Arias de Saavedra, F; Porras, I; Sánchez-Doblado, F

2016-01-01

Active thermal neutron detectors are used in a wide range of measuring devices in medicine, industry and research. For many applications, the long-term stability of these devices is crucial, so that very well controlled neutron fields are needed to perform calibrations and repeatability tests. A way to achieve such reference neutron fields, relying on a 3 MV Tandem Pelletron accelerator available at the CNA (Seville, Spain), is reported here. This paper shows thermal neutron field production and reproducibility characteristics over few days. Copyright © 2015 Elsevier Ltd. All rights reserved.

Thermal neutron detection using a silicon pad detector and {sup 6}LiF removable converters

Energy Technology Data Exchange (ETDEWEB)

Barbagallo, Massimo [Istituto Nazionale di Fisica Nucleare, Sezione di Bari (Italy); Cosentino, Luigi; Marchetta, Carmelo; Pappalardo, Alfio; Scire, Carlotta; Scire, Sergio; Schillaci, Maria; Vecchio, Gianfranco; Finocchiaro, Paolo [Istituto Nazionale di Fisica Nucleare, Laboratori Nazionali del Sud, Catania (Italy); Forcina, Vittorio; Peerani, Paolo [European Commission, Joint Research Centre, Institute of Transuranium Elements, Ispra (Italy); Vaccaro, Stefano [European Commission, Directorate-General for Energy (Luxembourg)

2013-03-15

A semiconductor detector coupled with a neutron converter is a good candidate for neutron detection, especially for its compactness and reliability if compared with other devices, such as {sup 3}He tubes, even though its intrinsic efficiency is rather lower. In this paper we show a neutron detector design consisting of a 3 cm Multiplication-Sign 3 cm silicon pad detector coupled with one or two external {sup 6}LiF layers, enriched in {sup 6}Li at 95%, placed in contact with the Si active surfaces. This prototype, first characterized and tested at INFN Laboratori Nazionali del Sud and then at JRC Ispra, was successfully shown to detect thermal neutrons with the expected efficiency and an outstanding gamma rejection capability.

Neutron detector development at Brookhaven

International Nuclear Information System (INIS)

Yu, B.; Harder, J.A.; Mead, J.A.; Radeka, V.; Schaknowski, N.A.; Smith, G.C.

2003-01-01

Two-dimensional thermal neutron detectors have been the subject of research and development at Brookhaven for over 20 years. Based primarily on multi-wire chambers filled with a gas mixture containing 3 He, these detectors have been used in wide-ranging studies of molecular biology and material science samples. At each phase of development, experimenters have sought improvements in key parameters such as position resolution, counting rate, efficiency, solid-angle coverage and stability. A suite of detectors has been developed with sensitive areas ranging from 5x5 to 50x50 cm 2 . These devices incorporate low-noise-position readout and the best position resolution for thermal neutron gas detectors. Recent developments include a 1.5 mx20 cm detector containing multiple segments with continuously sensitive readout, and detectors with unity gain for ultra-high rate capability and long-term stability

Improving the neutron-to-photon discrimination capability of detectors used for neutron dosimetry in high energy photon beam radiotherapy

International Nuclear Information System (INIS)

Irazola, L.; Terrón, J.A.; Bedogni, R; Pola, A.; Lorenzoli, M.; Sánchez-Nieto, B.; Gómez, F.; Sánchez-Doblado, F.

2016-01-01

The increasing interest of the medical community to radioinduced second malignancies due to photoneutrons in patients undergoing high-energy radiotherapy, has stimulated in recent years the study of peripheral doses, including the development of some dedicated active detectors. Although these devices are designed to respond to neutrons only, their parasitic photon response is usually not identically zero and anisotropic. The impact of these facts on measurement accuracy can be important, especially in points close to the photon field-edge. A simple method to estimate the photon contribution to detector readings is to cover it with a thermal neutron absorber with reduced secondary photon emission, such as a borated rubber. This technique was applied to the TNRD (Thermal Neutron Rate Detector), recently validated for thermal neutron measurements in high-energy photon radiotherapy. The positive results, together with the accessibility of the method, encourage its application to other detectors and different clinical scenarios. - Highlights: • Neutron-to-photon discrimination of a thermal neutron detector used in radiotherapy. • Photon and anisotropic response study with distance and beam incidence of thermal neutron detector. • Borated rubber for estimating photon contribution in any thermal neutron detector.

Neutron detection using Dy2O3 activation detectors

International Nuclear Information System (INIS)

Gomaa, M.A.; Mohamed, E.J.

1979-01-01

The aim of the present study is to examine the usefulness of Dy 2 O 3 not only as thermal neutron activation detector but also as a fast neutron detector. For thermal neutrons, the half life of 165 Dy is measured to be (141 +- 6) min, its response to thermal neutrons is (2.18 +- 0.01) cpm/ncm -2 s -1 for a 250 mg Dy 2 O 3 pellet. For fast neutrons the Dy 2 O 3 detector is placed within a 20 cm polyethylene sphere and its response is found to be (2.2 +- 0.1) cpm/ncm -2 s -1 for 4 MeV neutrons and (2.10 +- 0.04) cpm/ncm -2 s -1 for 14 MeV neutrons. For neutron dosimetry, its response is found to be (16.7 +- 0.4) cpm per mrem h -1 . (author)

The effect of incremental gamma-ray doses and incremental neutron fluences upon the performance of self-biased sup 1 sup 0 B-coated high-purity epitaxial GaAs thermal neutron detectors

CERN Document Server

Gersch, H K; Simpson, P A

2002-01-01

High-purity epitaxial GaAs sup 1 sup 0 B-coated thermal neutron detectors advantageously operate at room temperature without externally applied voltage. Sample detectors were systematically irradiated at fixed grid locations near the core of a 2 MW research reactor to determine their operational neutron dose threshold. Reactor pool locations were assigned so that fast and thermal neutron fluxes to the devices were similar. Neutron fluences ranged between 10 sup 1 sup 1 and 10 sup 1 sup 4 n/cm sup 2. GaAs detectors were exposed to exponential fluences of base ten. Ten detector designs were irradiated and studied, differentiated between p-i-n diodes and Schottky barrier diodes. The irradiated sup 1 sup 0 B-coated detectors were tested for neutron detection sensitivity in a thermalized neutron beam. Little damage was observed for detectors irradiated at neutron fluences of 10 sup 1 sup 2 n/cm sup 2 and below, but signals noticeably degraded at fluences of 10 sup 1 sup 3 n/cm sup 2. Catastrophic damage was appare...

Dual-sided microstructured semiconductor neutron detectors (DSMSNDs)

International Nuclear Information System (INIS)

Fronk, Ryan G.; Bellinger, Steven L.; Henson, Luke C.; Ochs, Taylor R.; Smith, Colten T.; Kenneth Shultis, J.; McGregor, Douglas S.

2015-01-01

Microstructured semiconductor neutron detectors (MSNDs) have in recent years received much interest as high-efficiency replacements for thin-film-coated thermal neutron detectors. The basic device structure of the MSND involves micro-sized trenches that are etched into a vertically-oriented pvn-junction diode that are backfilled with a neutron converting material. Neutrons absorbed within the converting material induce fission of the parent nucleus, producing a pair of energetic charged-particle reaction products that can be counted by the diode. The MSND deep-etched microstructures produce good neutron-absorption and reaction-product counting efficiencies, offering a 10× improvement in intrinsic thermal neutron detection efficiency over thin-film-coated devices. Performance of present-day MSNDs are nearing theoretical limits; streaming paths between the conversion-material backfilled trenches, allow a considerable fraction of neutrons to pass undetected through the device. Dual-sided microstructured semiconductor neutron detectors (DSMSNDs) have been developed that utilize a complementary second set of trenches on the back-side of the device to count streaming neutrons. DSMSND devices are theoretically capable of greater than 80% intrinsic thermal neutron detection efficiency for a 1-mm thick device. The first such prototype DSMSNDs, presented here, have achieved 29.48±0.29% nearly 2× better than MSNDs with similar microstructure dimensions.

Use of a newly developed active thermal neutron detector for in-phantom measurements in a medical LINAC

Energy Technology Data Exchange (ETDEWEB)

Bodogni, R.; Sanchez-Doblado, F.; Pola, A.; Gentile, A.; Esposito, A.; Gomez-ros, J. M.; Pressello, M. C.; Lagares, J. I.; Terron, J. A.; Gomez, F.

2013-07-01

In this work a newly developed active thermal neutron detector, based on a solid state analog device, was used to determine the thermal neutron fluence in selected positions of a simplified human phantom undergoing radiotherapy with a 15 MV LINAC. The results are compared with TLD, the predictions from a Monte Carlo simulation and with measurements indirectly performed with a digital device, located far from the phantom, inside the treatment room. In this work only TLD comparison is presented. Since active neutron instruments are usually affected by systematic deviations when used in a pulsed field with large photon background, the new detector offered in this work may represent an innovative and useful tool for neutron evaluations in accelerator-based radiotherapy. (Author)

SPECTRUM WEIGHTED RESPONSES OF SEVERAL DETECTORS IN MIXED FIELDS OF FAST AND THERMAL NEUTRONS

Directory of Open Access Journals (Sweden)

SANG IN KIM

2014-04-01

Full Text Available The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software ‘K-SWR’. The detectors’ response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403. The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the 241Am-Be sources held in a graphite pile, a bare 241Am-Be source, and a DT neutron generator. Fluence-average energy (Eave varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [H*(10/h] varied from 0.99 to 16.5 mSv/h.

Studies on neutron detection with solid state nuclear track detectors

International Nuclear Information System (INIS)

Khouri, M.C.; Vilela, E.C.; Andrade, C. de.

1993-03-01

The detection of thermal and fast neutrons was studied. For thermal neutrons, alpha sensitive plastic was used in order to register the products of nuclear reactions taking place in boron and /or lithium converters. Fast neutrons produce recoil tracks within the detector. In the present case, CR-39 and Makrofol E were used. Chemical and electrochemical etching processes were used for thermal and fast neutron detectors, respectively. (F.E.). 6 refs, 4 figs, 6 tabs

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)

Selective data analysis for diamond detectors in neutron fields

Directory of Open Access Journals (Sweden)

Weiss Christina

2017-01-01

Full Text Available Detectors based on synthetic chemical vapor deposition diamond gain importance in various neutron applications. The superior thermal robustness and the excellent radiation hardness of diamond as well as its excellent electronic properties make this material uniquely suited for rough environments, such as nuclear fission and fusion reactors. The intrinsic electronic properties of single-crystal diamond sensors allow distinguishing various interactions in the detector. This can be used to successfully suppress background of γ-rays and charged particles in different neutron experiments, such as neutron flux measurements in thermal nuclear reactors or cross-section measurements in fast neutron fields. A novel technique of distinguishing background reactions in neutron experiments with diamond detectors will be presented. A proof of principle will be given on the basis of experimental results in thermal and fast neutron fields.

Modeling of a Low-Background Spectroscopic Position-Sensitive Neutron Detector

Energy Technology Data Exchange (ETDEWEB)

Postovarova, Daria; Evsenin, Alexey; Gorshkov, Igor; Kuznetsov, Andrey; Osetrov, Oleg; Vakhtin, Dmitry; Yurmanov, Pavel [V.G. Khlopin Radium Institute, 194021, 28, 2nd Murinsky pr., Saint-Petersburg (Russian Federation)

2011-12-13

A new low-background spectroscopic direction-sensitive neutron detector that would allow one to reduce the neutron background component in passive and active neutron detection techniques is proposed. The detector is based on thermal neutron detectors surrounded by a fast neutron scintillation detector, which serves at the same time as a neutron moderator. Direction sensitivity is achieved by coincidence/anticoincidence analysis between different parts of the scintillator. Results of mathematical modeling of several detector configurations are presented.

Modeling of a Low-Background Spectroscopic Position-Sensitive Neutron Detector

International Nuclear Information System (INIS)

Postovarova, Daria; Evsenin, Alexey; Gorshkov, Igor; Kuznetsov, Andrey; Osetrov, Oleg; Vakhtin, Dmitry; Yurmanov, Pavel

2011-01-01

A new low-background spectroscopic direction-sensitive neutron detector that would allow one to reduce the neutron background component in passive and active neutron detection techniques is proposed. The detector is based on thermal neutron detectors surrounded by a fast neutron scintillation detector, which serves at the same time as a neutron moderator. Direction sensitivity is achieved by coincidence/anticoincidence analysis between different parts of the scintillator. Results of mathematical modeling of several detector configurations are presented.

Joint estimation of the fast and thermal components of a high neutron flux with a two on-line detector system

International Nuclear Information System (INIS)

Filliatre, P.; Oriol, L.; Jammes, C.; Vermeeren, L.

2009-01-01

A fission chamber with a 242 Pu deposit is the best suited detector for on-line measurements of the fast component of a high neutron flux (∼10 14 ncm -2 s -1 or more) with a significant thermal component. To get the fast flux, it is, however, necessary to subtract the contribution of the thermal neutrons, which increases with fluence because of the evolution of the isotopic content of the deposit. This paper presents an algorithm that permits, thanks to measurements provided by a 242 Pu fission chamber and a detector for thermal neutrons, to estimate the thermal and the fast flux at any time. An implementation allows to test it with simulated data.

Contribution to the study of position sensitive detectors with high spatial resolution for thermal neutron detection

International Nuclear Information System (INIS)

Idrissi Fakhr-Eddine, Abdellah.

1978-01-01

With a view to improving the spatial resolution of the localization of thermal neutrons, the work covers four position sensitive detectors: - 800 cell multi-detectors (1 dimension), - linear 'Jeu de Jacquet' detectors (1 dimension) - Multi-detector XYP 128x128 (2 dimensions), - 'Jeu de Jacquet' detector with 2 dimensions. Mention is made of the various position finding methods known so far, as well as the reasons for selecting BF 3 as detector gas. A study is then made of the parameters of the multiwire chamber whose principle will form the basis of most of the position detecting appliances subsequently dealt with. Finally, a description is given of the detection tests of the thermal neutrons in the multiwire chamber depending on the pressure, a parameter that greatly affects the accuracy of the position finding. The single dimension position tests on two kinds of appliance, the 800 cell multi-detector for the wide angle diffraction studies, and the linear 'Jeu de Jacquet' detector designed for small angle diffraction are mentioned. A description is then given of two position appliances with two dimensions; the multi-detector XYP 128x128 and the two dimensional 'Jeu de Jacquet' detector. In the case of this latter detector, only the hoped for characteristics are indicated [fr

Enriched Boron-Doped Amorphous Selenium Based Position-Sensitive Solid-State Thermal Neutron Detector for MPACT Applications

International Nuclear Information System (INIS)

Mandal, Krishna

2017-01-01

High-efficiency thermal neutron detectors with compact size, low power-rating and high spatial, temporal and energy resolution are essential to execute non-proliferation and safeguard protocols. The demands of such detector are not fully covered by the current detection system such as gas proportional counters or scintillator-photomultiplier tube combinations, which are limited by their detection efficiency, stability of response, speed of operation, and physical size. Furthermore, world-wide shortage of 3 He gas, required for widely used gas detection method, has further prompted to design an alternative system. Therefore, a solid-state neutron detection system without the requirement of 3 He will be very desirable. To address the above technology gap, we had proposed to develop new room temperature solidstate thermal neutron detectors based on enriched boron ( 10 B) and enriched lithium ( 6 Li) doped amorphous Se (As- 0.52%, Cl 5 ppm) semiconductor for MPACT applications. The proposed alloy materials have been identified for its many favorable characteristics - a wide bandgap (~2.2 eV at 300 K) for room temperature operation, high glass transition temperature (t g ~ 85°C), a high thermal neutron cross-section (for boron ~ 3840 barns, for lithium ~ 940 barns, 1 barn = 10 -24 cm 2 ), low effective atomic number of Se for small gamma ray sensitivity, and high radiation tolerance due to its amorphous structure.

Numerical Simulations of Pillar Structured Solid State Thermal Neutron Detector Efficiency and Gamma Discrimination

Energy Technology Data Exchange (ETDEWEB)

Conway, A; Wang, T; Deo, N; Cheung, C; Nikolic, R

2008-06-24

This work reports numerical simulations of a novel three-dimensionally integrated, {sup 10}boron ({sup 10}B) and silicon p+, intrinsic, n+ (PIN) diode micropillar array for thermal neutron detection. The inter-digitated device structure has a high probability of interaction between the Si PIN pillars and the charged particles (alpha and {sup 7}Li) created from the neutron - {sup 10}B reaction. In this work, the effect of both the 3-D geometry (including pillar diameter, separation and height) and energy loss mechanisms are investigated via simulations to predict the neutron detection efficiency and gamma discrimination of this structure. The simulation results are demonstrated to compare well with the measurement results. This indicates that upon scaling the pillar height, a high efficiency thermal neutron detector is possible.

Thermal neutron detector and gamma-ray spectrometer utilizing a single material

Science.gov (United States)

Stowe, Ashley; Burger, Arnold; Lukosi, Eric

2017-05-02

A combined thermal neutron detector and gamma-ray spectrometer system, including: a detection medium including a lithium chalcopyrite crystal operable for detecting thermal neutrons in a semiconductor mode and gamma-rays in a scintillator mode; and a photodetector coupled to the detection medium also operable for detecting the gamma rays. Optionally, the detection medium includes a .sup.6LiInSe.sub.2 crystal. Optionally, the detection medium comprises a compound formed by the process of: melting a Group III element; adding a Group I element to the melted Group III element at a rate that allows the Group I and Group III elements to react thereby providing a single phase I-III compound; and adding a Group VI element to the single phase I-III compound and heating; wherein the Group I element includes lithium.

Joint estimation of the fast and thermal components of a high neutron flux with a two on-line detector system

Energy Technology Data Exchange (ETDEWEB)

Filliatre, P. [CEA, DEN, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France)], E-mail: philippe.filliatre@cea.fr; Oriol, L.; Jammes, C. [CEA, DEN, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France)

2009-05-21

A fission chamber with a {sup 242}Pu deposit is the best suited detector for on-line measurements of the fast component of a high neutron flux ({approx}10{sup 14}ncm{sup -2}s{sup -1} or more) with a significant thermal component. To get the fast flux, it is, however, necessary to subtract the contribution of the thermal neutrons, which increases with fluence because of the evolution of the isotopic content of the deposit. This paper presents an algorithm that permits, thanks to measurements provided by a {sup 242}Pu fission chamber and a detector for thermal neutrons, to estimate the thermal and the fast flux at any time. An implementation allows to test it with simulated data.

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.

Enriched Boron-Doped Amorphous Selenium Based Position-Sensitive Solid-State Thermal Neutron Detector for MPACT Applications

Energy Technology Data Exchange (ETDEWEB)

Mandal, Krishna [Univ. of South Carolina, Columbia, SC (United States)

2017-09-29

High-efficiency thermal neutron detectors with compact size, low power-rating and high spatial, temporal and energy resolution are essential to execute non-proliferation and safeguard protocols. The demands of such detector are not fully covered by the current detection system such as gas proportional counters or scintillator-photomultiplier tube combinations, which are limited by their detection efficiency, stability of response, speed of operation, and physical size. Furthermore, world-wide shortage of ³He gas, required for widely used gas detection method, has further prompted to design an alternative system. Therefore, a solid-state neutron detection system without the requirement of ³He will be very desirable. To address the above technology gap, we had proposed to develop new room temperature solidstate thermal neutron detectors based on enriched boron (¹⁰B) and enriched lithium (⁶Li) doped amorphous Se (As- 0.52%, Cl 5 ppm) semiconductor for MPACT applications. The proposed alloy materials have been identified for its many favorable characteristics - a wide bandgap (~2.2 eV at 300 K) for room temperature operation, high glass transition temperature (t_g ~ 85°C), a high thermal neutron cross-section (for boron ~ 3840 barns, for lithium ~ 940 barns, 1 barn = 10^-24 cm²), low effective atomic number of Se for small gamma ray sensitivity, and high radiation tolerance due to its amorphous structure.

Study of the Li{sub 2}CO{sub 3} as thermal neutrons detector; Estudio del Li{sub 2}CO{sub 3} como detector de neutrones termicos

Energy Technology Data Exchange (ETDEWEB)

Herrera A, E.; Urena N, F.; Delfin L, A. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)] e-mail: eha@nuclear.inin.mx

2003-07-01

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 {sup 197}Au (n, {gamma} ) {sup 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{sub 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{sub 2}CO{sub 3} complete with the requirements to be used as detectors and/or dosemeters of thermal neutrons. (Author)

In-core neutron flux measurements at PARR using self powered neutron detector

International Nuclear Information System (INIS)

Hussain, A.; Ansari, S.A.

1989-10-01

This report describes experimental reactor physics measure ments at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response time determination and neutron flux measurements. The detectors were calibrated at low power using gold foils and full power by the thermal channel. Based on this calibration it was observed that the detector response remains almost linear throughout the power range. The self powered detectors were used for on-line determination of absolute neutron flux in the core as well as the spatial distribution of neutron flux or reactor power. The experimental, axial and horizontal flux mapping results at certain locations in the core are presented. The total response time of rhodium detector was experimentally determined to be about 5 minutes, which agree well with the theoretical results. Because of longer response time of SPND of the detectors it is not possible to use them in the reactor protection system. (author). 10 figs

Personal Dosemeter of Thermal Neutron Using A Cr-39 Detector with Filter Natural LiF

International Nuclear Information System (INIS)

Sofyan, Hasnel; Thamrin, M.Thoyib

1996-01-01

The research of personal dosemeter for thermal neutron using Cr-39 detector with different thicknesses of natural LiF filter was carried out. The irradiation of Cr-39 detector with neutron source from reactor research TRIGA mark II of Rikkyo University Tokyo, Japan. Nuclear track was counted by automatic method with ASPECT ver.4.22 Series A4T124 software and manual method for correction. The result of research, the maximum of nuclear tracks was obtained at 8 mm of LiF filter was 10 mm with 11,630x10E-5 track/neutron for air radiation. And on phantom radiation, the thickness of filter was 10 mm with 11,630x10E-5 track/neutron. Its values were 3.6 and 7.5 bigger than the response of Cr-39 non filter in air and on phantom radiation, respectively

Toward achieving flexible and high sensitivity hexagonal boron nitride neutron detectors

Science.gov (United States)

Maity, A.; Grenadier, S. J.; Li, J.; Lin, J. Y.; Jiang, H. X.

2017-07-01

Hexagonal boron nitride (h-BN) detectors have demonstrated the highest thermal neutron detection efficiency to date among solid-state neutron detectors at about 51%. We report here the realization of h-BN neutron detectors possessing one order of magnitude enhancement in the detection area but maintaining an equal level of detection efficiency of previous achievement. These 3 mm × 3 mm detectors were fabricated from 50 μm thick freestanding and flexible 10B enriched h-BN (h-10BN) films, grown by metal organic chemical vapor deposition followed by mechanical separation from sapphire substrates. Mobility-lifetime results suggested that holes are the majority carriers in unintentionally doped h-BN. The detectors were tested under thermal neutron irradiation from californium-252 (252Cf) moderated by a high density polyethylene moderator. A thermal neutron detection efficiency of ˜53% was achieved at a bias voltage of 200 V. Conforming to traditional solid-state detectors, the realization of h-BN epilayers with enhanced electrical transport properties is the key to enable scaling up the device sizes. More specifically, the present results revealed that achieving an electrical resistivity of greater than 1014 Ωṡcm and a leakage current density of below 3 × 10-10 A/cm2 is needed to fabricate large area h-BN detectors and provided guidance for achieving high sensitivity solid state neutron detectors based on h-BN.

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

Preliminary results on bubble detector as personal neutron dosemeter

International Nuclear Information System (INIS)

Ponraju, D.; Krishnan, H.; Viswanathan, S.; Indira, R.

2011-01-01

The bubble detector is demonstrated as one of the best suitable neutron detectors for neutron dose rate measurements in the presence of high-intense gamma fields. Immobilisation of a volatile liquid in a superheated state and achieving uniform distribution of tiny superheated droplets were a practical challenge. A compact and reusable bubble detector with high neutron sensitivity has been developed at the Indira Gandhi Centre for Atomic Research by immobilising the superheated droplets in a suitable polymer matrix. Two types of bubble detectors have been successfully developed, one by incorporating isobutane for measuring fast neutron and another by incorporating Freon-12 for both fast and thermal neutron. The performance of the detector has been tested using 5 Ci Am-Be neutron source and the results are described. (authors)

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

Improving differential die-away analysis via the use of neutron poisons in detectors

International Nuclear Information System (INIS)

Jordan, Kelly A.; Vujic, Jasmina; Phillips, Emmanuel; Gozani, Tsahi

2007-01-01

Differential Die-Away Analysis (DDAA) is an active interrogation technique to detect special nuclear material (SNM). In DDAA, a pulsed neutron generator produces pulses of neutrons that are directed into a cargo to be interrogated. As each pulse passes through the cargo, the neutrons are thermalized and absorbed. If SNM is present, the thermalized neutrons from the source will cause fissions that produce a new source of neutrons. The number of thermal neutrons decay exponentially with the diffusion decay time of the inspected medium, on the order of hundreds of μs. An external neutron detector which is designed to detect only epithermal neutrons, will measure only a single decaying exponential when there is no SNM present, and two exponentials when SNM is present. This paper shows that in many cases, a gain in detection sensitivity can be realized by introducing a thermal neutron poison (such as boron) into the detector. This poison will reduce the efficiency of the detector, but decrease its decay time. A decreased decay time will cause the separation between the detector and fission signal exponentials to occur at an earlier time. There is a balance between efficiency and time constant for a detector. The boron concentration to achieve the maximum sensitivity, and its magnitude, will be different for different detector designs

Experimental evaluation of scattered thermal neutrons from various jig materials for use in fixing detectors for the calibration

International Nuclear Information System (INIS)

Shimizu, Shigeru; Yoshizawa, Michio

2000-05-01

Some jigs to fix detectors are used when radiation measuring instruments are calibrated or reference fluence rates are measured in thermal neutron irradiation fields. In this case, scattered thermal neutrons from the jigs, in particular, which contain hydrogenous materials, may affect the results of the calibration and measurements. In this study, scattered thermal neutrons were measured and calculated to clarify the characteristics of the thermal neutron scattered from various materials which are frequently used for the jigs. A spherical BF 3 -counter of 2-inches in diameter was used in the experiment. Ratios of the fluence of scattered neutrons to primaries (hereinafter, scattering ratio) were evaluated as a function of thickness and size of the materials, as well as the distance from the surface of the materials. The scattering ratios of the jigs that were actually-used in the calibration were also measured in order to select appropriate materials and thickness for the jigs. It was found that the scattering ratios were saturated with increase of thickness and size of the materials. The higher values were observed in the case of PMMA (polymethylmethacrylates) and paraffin since these materials contain more number of hydrogen atoms than the others. The saturated value was obtained 130% for PMMA and paraffin with the thickness of more than 5 cm and the size of 40 cm x 40 cm. The results for the actually-used jigs show that the thinner plate of styrofoam and aluminum reduces the scattering ratio to the value of less than 1%. The obtained data will be useful to improve the accuracy of the calibration of thermal neutron detectors and the measurement of reference fluence rates in thermal neutron irradiation fields. (author)

Experimental evaluation of scattered thermal neutrons from various jig materials for use in fixing detectors for the calibration

Energy Technology Data Exchange (ETDEWEB)

Shimizu, Shigeru; Yoshizawa, Michio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nemoto, Hisashi; Kurosawa, Koji [Institute of Radiation Measurement, Tokai, Ibaraki (Japan)

2000-05-01

Some jigs to fix detectors are used when radiation measuring instruments are calibrated or reference fluence rates are measured in thermal neutron irradiation fields. In this case, scattered thermal neutrons from the jigs, in particular, which contain hydrogenous materials, may affect the results of the calibration and measurements. In this study, scattered thermal neutrons were measured and calculated to clarify the characteristics of the thermal neutron scattered from various materials which are frequently used for the jigs. A spherical BF{sub 3}-counter of 2-inches in diameter was used in the experiment. Ratios of the fluence of scattered neutrons to primaries (hereinafter, scattering ratio) were evaluated as a function of thickness and size of the materials, as well as the distance from the surface of the materials. The scattering ratios of the jigs that were actually-used in the calibration were also measured in order to select appropriate materials and thickness for the jigs. It was found that the scattering ratios were saturated with increase of thickness and size of the materials. The higher values were observed in the case of PMMA (polymethylmethacrylates) and paraffin since these materials contain more number of hydrogen atoms than the others. The saturated value was obtained 130% for PMMA and paraffin with the thickness of more than 5 cm and the size of 40 cm x 40 cm. The results for the actually-used jigs show that the thinner plate of styrofoam and aluminum reduces the scattering ratio to the value of less than 1%. The obtained data will be useful to improve the accuracy of the calibration of thermal neutron detectors and the measurement of reference fluence rates in thermal neutron irradiation fields. (author)

Energy response of neutron area monitor with silicon semiconductor detector

International Nuclear Information System (INIS)

Kitaguchi, Hiroshi; Izumi, Sigeru; Kobayashi, Kaoru; Kaihara, Akihisa; Nakamura, Takashi.

1993-01-01

A prototype neutron area monitor with a silicon semiconductor detector has been developed which has the energy response of 1 cm dose equivalent recommended by the ICRP-26. Boron and proton radiators are coated on the surface of the silicon semiconductor detector. The detector is set at the center of a cylindrical polyethylene moderator. This moderator is covered by a porous cadmium board which serves as the thermal neutron absorber. Neutrons are detected as α-particles generated by the nuclear reaction 10 B(n,α) 7 Li and as recoil protons generated by the interaction of fast neutrons with hydrogen. The neutron energy response of the monitor was measured using thermal neutrons and monoenergetic fast neutrons generated by an accelerator. The response was consistent with the 1 cm dose equivalent response required for the monitor within ±34% in the range of 0.025 - 15 Mev. (author)

Nuclear reactor ex-core startup neutron detector

International Nuclear Information System (INIS)

Wyvill, J.R.

1980-01-01

A sensitive ex-core neutron detector is needed to monitor the power level of reactors during startup. The neutron detector of this invention has a photomultiplier with window resistant to radiation darkening at the input end and an electrical connector at the output end. The photomultiplier receives light signals from a neutron-responsive scintillator medium, typically a cerium-doped lithium silicate glass, that responds to neutrons after they have been thermalized by a silicone resin moderator. Enclosing and shielding the photmultiplier, the scintillator medium and the moderator is a combined lead and borated silicone resin housing

NEUTRON SPECTRUM MEASUREMENTS USING MULTIPLE THRESHOLD DETECTORS

Energy Technology Data Exchange (ETDEWEB)

Gerken, William W.; Duffey, Dick

1963-11-15

From American Nuclear Society Meeting, New York, Nov. 1963. The use of threshold detectors, which simultaneously undergo reactions with thermal neutrons and two or more fast neutron threshold reactions, was applied to measurements of the neutron spectrum in a reactor. A number of different materials were irradiated to determine the most practical ones for use as multiple threshold detectors. These results, as well as counting techniques and corrections, are presented. Some materials used include aluminum, alloys of Al -Ni, aluminum-- nickel oxides, and magesium orthophosphates. (auth)

Measure of thermal neutron flux in the IPEN/MB-01 reactor using {sup 197} Au wire activation detectors; Medida do fluxo de neutrons termicos do reator IPEN/MB-01 com detectores de ativacao de fios de {sup 197} Au

Energy Technology Data Exchange (ETDEWEB)

Marques, Andre Luis Ferreira

1995-12-31

This dissertation has aimed at developing a neutron flux measurement technique by means of detectors activation analysis. The main task of this work was the implementation of this thermal neutron flux measurement technique, using gold wires as activation detectors in the IPEN/MB-01 reactor core. The neutron thermal flux spatial distribution was obtained by gold wire activation technique, with wire diameters of 0.125 mm and 0.250 mm in seven selected reactor experimental channels. The values of thermal flux were about 10{sup 9} neutrons/cm{sup 2}.s. This experiment has been the first one conducted with gold wires in the IPEN/MB-01 reactor, being this technique implemented for use by experiments in flux mapping of the core 73 refs., 60 figs., 31 tabs.

Measure of thermal neutron flux in the IPEN/MB-01 reactor using {sup 197} Au wire activation detectors; Medida do fluxo de neutrons termicos do reator IPEN/MB-01 com detectores de ativacao de fios de {sup 197} Au

Energy Technology Data Exchange (ETDEWEB)

Marques, Andre Luis Ferreira

1996-12-31

This dissertation has aimed at developing a neutron flux measurement technique by means of detectors activation analysis. The main task of this work was the implementation of this thermal neutron flux measurement technique, using gold wires as activation detectors in the IPEN/MB-01 reactor core. The neutron thermal flux spatial distribution was obtained by gold wire activation technique, with wire diameters of 0.125 mm and 0.250 mm in seven selected reactor experimental channels. The values of thermal flux were about 10{sup 9} neutrons/cm{sup 2}.s. This experiment has been the first one conducted with gold wires in the IPEN/MB-01 reactor, being this technique implemented for use by experiments in flux mapping of the core 73 refs., 60 figs., 31 tabs.

Measurement of neutron sensitivity of self powered neutron detectors

International Nuclear Information System (INIS)

Mahant, A.K.; Yeshuraja, V.; Ghodke, Shobha

2005-01-01

Self powered neutron detectors (SPNDs ) will form the part of Reactor Instrumentation in the upcoming 500 MWe power reactors. ECIL has developed Vanadium and Cobalt SPNDs for NPCIL to be used in regulation and protection channels. Experimental determination of neutron sensitivity of the vanadium and cobalt Self Powered Neutron Detectors (SPNDs) was carried out in A-l location of Apsara reactor at BARC. The measurements involved determination of total detector signal, its various components and the thermal neutron flux at the detector location. The paper describes the experimental techniques used to measure various parameters required to evaluate the neutron sensitivity of the SPNDs and also the parameters required to ascertain the integrity of SPNDs. Neutron flux measurement was done by gold foil irradiation technique. The predominant signal component from the vanadium SPND is Ib the current due to activation of the vanadium emitter, it forms about 85% of the total signal. The other components I n,γ due to the capture gamma rays of 52 V and I externalγ produced by the external reactor gamma rays contribute about 10% and 5% respectively to the total signal. Whereas in the cobalt SPND the main signal component is due to the capture gamma rays of 60 Co and accounts for about the 95% of the total signal. Remaining 5% signal is due to external reactor gamma rays. (author)

Evaluation of CdZnTe as neutron detector around medical accelerators

International Nuclear Information System (INIS)

Martin-Martin, A.; Iniguez, M. P.; Luke, P. N.; Barquero, R.; Lorente, A.; Morchon, J.; Gallego, E.; Quincoces, G.; Marti-Climent, J. M.

2009-01-01

The operation of electron linear accelerators (LINACs) and cyclotrons can produce a mixed gamma-neutron field composed of energetic neutrons coming directly from the source and scattered lower energy neutrons. The thermal neutron detection properties of a non-moderated coplanar-grid CdZnTe (CZT) gamma-ray detector close to an 18 MV electron LINAC and an 18 MeV proton cyclotron producing the radioisotope 18 F for positron emission tomography are investigated. The two accelerators are operated at conditions producing similar thermal neutron fluence rates of the order of 104 cm -2 s -1 at the measurement locations. The counting efficiency of the CZT detector using the prompt 558 keV photopeak following 113 Cd thermal neutron capture is evaluated and a good neutron detection performance is found at the two installations. (authors)

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.

Characterization of the Ljubljana TRIGA thermal column neutron radiographic facility

International Nuclear Information System (INIS)

Nemec, T.; Rant, J.; Kristof, E.; Glumac, B.

1995-01-01

An extensive characterization of the neutron beam of the existing neutron radiographic facility in the thermal column of the Ljubljana Triga Mark II research reactor is in progress. Neutron beam characteristics are needed to determine the effect of various neutron and gamma radiation on the neutron radiographic image. Commercially available medical scintillator converter screens based on Gd dioxy sulphite as well as Gd metal neutron converters are used to record neutron radiographic image. Thermal, epithermal and fast neutron fluxes were measured using Au and In activation detectors and cadmium ratio is determined. Neutron beam flux profiles are measured by film densitometry and by Au activation detector wires. By exposing films shielded by boral or lead plates individual contributions of thermal, epithermal neutrons and gamma radiation are estimated by densitometric measurements. By recording images of neutron image quality indicators BPI (Beam Purity Indicator) and SI (Sensitivity Indicator) produced by Riso, standard neutron radiography image characteristic are established. In gamma dosimetric measurements thermoluminescent detectors (CaF 2 Mn) are used. (author)

Design and spectrum calculation of 4H-SiC thermal neutron detectors using FLUKA and TCAD

Science.gov (United States)

Huang, Haili; Tang, Xiaoyan; Guo, Hui; Zhang, Yimen; Zhang, Yimeng; Zhang, Yuming

2016-10-01

SiC is a promising material for neutron detection in a harsh environment due to its wide band gap, high displacement threshold energy and high thermal conductivity. To increase the detection efficiency of SiC, a converter such as 6LiF or 10B is introduced. In this paper, pulse-height spectra of a PIN diode with a 6LiF conversion layer exposed to thermal neutrons (0.026 eV) are calculated using TCAD and Monte Carlo simulations. First, the conversion efficiency of a thermal neutron with respect to the thickness of 6LiF was calculated by using a FLUKA code, and a maximal efficiency of approximately 5% was achieved. Next, the energy distributions of both 3H and α induced by the 6LiF reaction according to different ranges of emission angle are analyzed. Subsequently, transient pulses generated by the bombardment of single 3H or α-particles are calculated. Finally, pulse height spectra are obtained with a detector efficiency of 4.53%. Comparisons of the simulated result with the experimental data are also presented, and the calculated spectrum shows an acceptable similarity to the experimental data. This work would be useful for radiation-sensing applications, especially for SiC detector design.

CVD polycrystalline diamond. A novel neutron detector and applications

International Nuclear Information System (INIS)

Mongkolnavin, R.

1998-01-01

Chemical Vapour Deposition (CVD) Polycrystalline Diamond film has been investigated as a low noise sensor for beta particles, gammas and neutrons using High Energy Physics technologies. Its advantages and disadvantages have been explored in comparison with other particle detectors such as silicon detector and other plastic scintillators. The performance and characteristic of the diamond detector have been fully studied and discussed. These studies will lead to a better understanding of how CVD diamonds perform as a detector and how to improve their performance under various conditions. A CVD diamond detector model has been proposed which is an attempt to explain the behaviour of such an extreme detector material. A novel neutron detector is introduced as a result of these studies. A good thermal and fast neutron detector can be fabricated with CVD diamond with new topologies. This detector will perform well without degradation in a high neutron radiation environment, as diamond is known to be radiation hard. It also offers better neutrons and gammas discrimination for high gamma background applications compared to other semiconductor detectors. A full simulation of the detector has also been done using GEANT, a Monte-Carlo simulation program for particle detectors. Simulation results show that CVD diamond detectors with this novel topology can detect neutrons with great directionality. Experimental work has been done on this detector in a nuclear reactor environment and accelerator source. A novel neutron source which offers a fast pulse high-energy neutrons has also been studied. With this detector, applications in neutron spectrometer for low-Z material have been pursued with various neutron detection techniques. One of these is a low-Z material identification system. The system has been designed and simulated for contraband luggage interrogation using the detector and the novel neutron source. Also other neutron related applications have been suggested. (author)

CVD polycrystalline diamond. A novel neutron detector and applications

International Nuclear Information System (INIS)

Mongkolnavin, R.

1998-07-01

Chemical Vapour Deposition (CVD) Polycrystalline Diamond film has been investigated as a low noise sensor for beta particles, gammas and neutrons using High Energy Physics technologies. Its advantages and disadvantages have been explored in comparison with other particle detectors such as silicon detector and other plastic scintillators. The performance and characteristic of the diamond detector have been fully studied and discussed. These studies will lead to a better understanding of how CVD diamonds perform as a detector and how to improve their performance under various conditions. A CVD diamond detector model has been proposed which is an attempt to explain the behaviour of such an extreme detector material. A novel neutron detector is introduced as a result of these studies. A good thermal and fast neutron detector can be fabricated with CVD diamond with new topologies. This detector will perform well without degradation in a high neutron radiation environment, as diamond is known to be radiation-hard. It also offers better neutrons and gammas discrimination for high gamma background applications compared to other semiconductor detectors. A full simulation of the detector has also been done using GEANT, a Monte Carlo simulation program for particle detectors. Simulation results show that CVD diamond detectors with this novel topology can detect neutrons with great directionality. Experimental work has been done on this detector in a nuclear reactor environment and accelerator source. A novel neutron source which offers a fast pulse high-energy neutrons has also been studied. With this detector, applications in neutron spectrometry for low-Z material have been pursued with various neutron detection techniques. One of these is a low-Z material identification system. The system has been designed and simulated for contraband luggage interrogation using the detector and the novel neutron source. (author)

Characteristics of Fabricated SiC Neutron Detectors for Neutron Flux Monitoring

Energy Technology Data Exchange (ETDEWEB)

Kim, Han Soo; Ha, Jang Ho; Park, Se Hwan; Lee, Kyu Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Ho [Hanyang University, Seoul (Korea, Republic of)

2011-05-15

An SPND (Self-powered Neutron Detector) is commonly used for neutron detection in NPP (Nuclear Power Plant) by virtue of un-reactivity for gamma-rays. But it has a drawback, which is that it cannot detect neutrons in real time due to beta emissions (about > 48 s) after reactions between neutrons and {sup 103}Rh in an SPND. And Generation IV reactors such as MSR (Molten-salt reactor), SFR (Sodium-cooled fast reactor), and GFR (Gas-cooled fast reactor) are designed to compact size and integration type. For GEN IV reactor, neutron monitor also must be compact-sized to apply such reactor easily and much more reliable. The wide band-gap semiconductors such as SiC, AlN, and diamond make them an attractive alternative in applications in harsh environments by virtue of the lower operating voltage, faster charge-collection times compared with gas-filled detectors, and compact size.1) In this study, two PIN-type SiC semiconductor neutron detectors, which are for fast neutron detection by elastic and inelastic scattering SiC atoms and for thermal neutron detection by charged particle emissions of 6LiF reaction, were designed and fabricated for NPP-related applications. Preliminary tests such as I-V and alpha response were performed and neutron responses at ENF in HANARO research reactor were also addressed. The application feasibility of the fabricated SiC neutron detector as an in-core neutron monitor was discussed

Neutron detection at jet using artificial diamond detectors

International Nuclear Information System (INIS)

Pillon, M.; Angelone, M.; Lattanzi, D.; Marinelli, M.; Milani, E.; Tucciarone, A.; Verona-Rinati, G.; Popovichev, S.; Montereali, R.M.; Vincenti, M.A.; Murari, A.

2007-01-01

Artificial diamond neutron detectors recently proved to be promising devices to measure the neutron production on large experimental fusion machines. Diamond detectors are very promising detectors to be used in fusion environment due to their radiation hardness, low sensitivity to gamma rays, fast response and high energy resolution. High quality 'electronic grade' diamond films are produced through microwave chemical vapour deposition (CVD) technique. Two CVD diamond detectors have been installed and operated at joint European torus (JET), Culham Science Centre, UK. One of these detectors was a polycrystalline CVD diamond film; about 12 mm 2 area and 30 μm thickness while the second was a monocrystalline film of about 5 mm 2 area and 20 μm thick. Both diamonds were covered with 2 μm of lithium fluoride (LiF) 95% enriched in 6 Li. The LiF layer works as a neutron-to-charged particle converter so these detectors can measure thermalized neutrons. Their output signals were compared to JET total neutron yield monitors (KN1 diagnostic) realized with a set of uranium fission chambers. Despite their small active volumes the diamond detectors were able to measure total neutron yields with good reliability and stability during the recent JET experimental campaign of 2006

Using Backscattering to Enhance Efficiency in Neutron Detectors

DEFF Research Database (Denmark)

Kittelmann, T.; Kanaki, K.; Klinkby, Esben Bryndt

2017-01-01

The principle of using strongly scattering materials to recover efficiency in detectors for neutron instruments, via backscattering of unconverted thermal neutrons, is discussed in general. The feasibility of the method is illustrated through Geant4-based simulations involving thermal neutrons im......, respectively, centimeters and tens of microseconds. Potential mitigation techniques to contain the impact on resolution are investigated and are found to alleviate the issues to some degree, at a cost of reduced gain in efficiency....

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

Earth formation pulsed neutron porosity logging system utilizing epithermal neutron and inelastic scattering gamma ray detectors

International Nuclear Information System (INIS)

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

1978-01-01

An improved pulsed neutron porosity logging system is provided in the present invention. A logging tool provided with a 14 MeV pulsed neutron source, an epithermal neutron detector and an inelastic scattering gamma ray detector is moved through a borehole. The detection of inelastic gamma rays provides a measure of the fast neutron population in the vicinity of the detector. repetitive bursts of neutrons irradiate the earth formation and, during the busts, inelastic gamma rays representative of the fast neutron population is sampled. During the interval between bursts the epithermal neutron population is sampled along with background gamma radiation due to lingering thermal neutrons. the fast and epithermal neutron population measurements are combined to provide a measurement of formation porosity

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.

The use of multi-energy-group neutron diffusion theory to numerically evaluate the relative utility of three dial-detector neutron porosity well logging tools

International Nuclear Information System (INIS)

Zalan, T.A.

1988-01-01

Multi-energy-group neutron diffusion theory is used to numerically evaluate the utility of two different dual-detector neutron porosity logging devices, a 14 MeV (accelerator) neutron source - epithermal neutron detector device and a 4 MeV neutron source - capture gamma-ray detector device, relative to the traditional 4 MeV neutron source - thermal neutron detector device. Fast and epithermal neutron diffusion parameters are calculated using Monte Carlo - derived neutron flux distributions. Thermal parameters are calculated from tabulated cross sections. An existing analytical method to describe the transport of gamma-rays through common earth materials is modified in order to accommodate the modeling of the 4 MeV neutron - capture gamma-ray device. The 14 MeV neutron - epithermal neutron device is found to be less sensitive to porosity than the 4 MeV neutron - capture gamma-ray device, which in turn is found to be less sensitive to porosity than the traditional 4 MeV neutron - thermal neutron device. Salinity effects are found to be comparable for the 4 MeV neutron - capture gamma-ray and 4 MeV neutron - thermal neutron devices. The 4 MeV neutron capture gamma-ray measurement is found to be deepest investigating

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)

Dose-equivalent response CR-39 track detector for personnel neutron dosimetry

International Nuclear Information System (INIS)

Oda, K.; Ito, M.; Yoneda, H.; Miyake, H.; Yamamoto, J.; Tsuruta, T.

1991-01-01

A dose-equivalent response detector based on CR-39 has been designed to be applied for personnel neutron dosimetry. The intrinsic detection efficiency of bare CR-39 was first evaluated from irradiation experiments with monoenergetic neutrons and theoretical calculations. In the second step, the radiator effect was investigated for the purpose of sensitization to fast neutrons. A two-layer radiator consisting of deuterized dotriacontane (C 32 D 66 ) and polyethylene (CH 2 ) was designed. Finally, we made the CR-39 detector sensitive to thermal neutrons by doping with orthocarbone (B 10 H 12 C 2 ), and also estimated the contribution of albedo neutrons. It was found that the new detector - boron-doped CR-39 with the two-layer radiator - would have a flat response with an error of about 70% in a wide energy region, ranging from thermal to 15 MeV. (orig.)

Development of a modular directional and spectral neutron detection system using solid-state detectors

Energy Technology Data Exchange (ETDEWEB)

Weltz, A., E-mail: weltza3@gmail.com; Torres, B.; McElwain, L.; Dahal, R.; Huang, J.; Bhat, I.; Lu, J.; Danon, Y.

2015-08-21

A detection system using room-temperature, microstructured solid-state thermal neutron detectors with very low leakage current has been developed at Rensselaer Polytechnic Institute (RPI) with the ability to provide positional and spectral information about an unknown neutron source. The Directional and Spectral Neutron Detection System (DSNDS) utilizes a set of small-but-scalable, zero-bias solid-state thermal neutron detectors which have demonstrated high thermal neutron efficiency and adequate gamma insensitivity. The DSNDS can gather spectral information about an unknown neutron source with a relatively small number of detectors, simplifying the detector electronics and minimizing cost; however, the DSNDS is modular in design, providing the capability to increase the detection efficiency and angular resolution. The system used in this paper was comprised of a stack of five high-density polyethylene (HDPE) disks with a thickness of 5 cm and a diameter of 30 cm, the middle disk containing 16 detectors positioned as one internal (moderated) and one external (unmoderated) ring of solid-state neutron detectors. These two detector rings provide the ability to determine the directionality of a neutron source. The system gathers spectral information about a neutron source in two ways: by measuring the relative responses of the internal ring of detectors as well as measuring the ratio of the internal-to-external detector responses. Experiments were performed with variable neutron spectra: a {sup 252}Cf spontaneous fission neutron source which was HDPE moderated, HDPE reflected, lead (Pb) shielded, and bare in order to benchmark the system for spectral sensitivity. Simulations were performed in order to characterize the neutron spectra corresponding to each of the source configurations and showed agreement with experimental measurements. The DSNDS demonstrates the ability to determine the relative angle of the source and the hardness of the neutron spectrum. By using the

Thermal, intermediate and fast neutron flux measurements using activation detectors; Mesure des flux de neutrons thermiques, intermediaires et rapides au moyen de detecteurs par activation

Energy Technology Data Exchange (ETDEWEB)

Brisbois, J; Lott, M; Manent, G [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

The problem of neutron flux measurements using activation detectors is studied in the particular case of protection research. It is shown how it possible, it is possible, using a known thermal flux, to organise a coherent calibration system for all the detectors. The rapid neutron detectors are calibrated with respect to a reference detector (phosphorus) in a natural uranium converter; the intermediate neutron detectors with respect to gold in the axial channel of ZOE. This method makes it possible to minimise the errors due to the activation cross-sections. A brief description is given of the counting room of the Pile Safety Study Service, as well of the practical utilisation characteristics of the counters employed. (authors) [French] Le probleme de la mesure des flux de neutrons au moyen de detecteurs par activation est etudie dans le cas particulier des etudes de protections. On montre comment, a partir d'un flux thermique connu, on peut organiser un systeme coherent d'etalonnage de tous les detecteurs. Les detecteurs de neutrons rapides sont etalonnes par rapport a un detecteur de reference (phosphore) dans un convertisseur en uranium naturel; les detecteurs de neutrons intermediaires, par rapport a l'or dans le canal axial de ZOE, Cette methode permet de minimiser les erreurs dues aux sections efficaces d'activation. On decrit sommairement la salle de comptage du Service d'Etudes de Protections de Piles et on indique les caracteristiques d'emploi pratique des detecteurs utilises. (auteurs)

Exotic geophysical phenomena observed in an environmental neutron flux study using EAS PRISMA detectors

Directory of Open Access Journals (Sweden)

Alekseenko Victor

2017-01-01

Full Text Available Some exotic geophysical events are observed by a global net of electron-neutron detectors (en-detectors developed in the framework of the PRISMA EAS project. Our en-detectors running both on the Earth's surface and underground are continuously measuring the environmental thermal neutron flux. Thermal neutrons are in equilibrium with media and are therefore sensitive to many geophysical phenomena, which are exotic for people studying ultra high-energy cosmic rays or carrying out low background experiments deep underground.

A new detector for the measurement of neutron flux in nuclear reactors

International Nuclear Information System (INIS)

Koch, L.; Labeyrie, J.; Tarassenko, S.

1958-01-01

The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high γ ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their β and γ disintegration. Thermal neutron fluxes as low as few neutrons per cm 2 per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm 2 ) for fluxes higher than 10 8 n/cm 2 /s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author) [fr

Calibration of detector efficiency of neutron detector

International Nuclear Information System (INIS)

Guo Hongsheng; He Xijun; Xu Rongkun; Peng Taiping

2001-01-01

BF 3 neutron detector has been set up. Detector efficiency is calibrated by associated particle technique. It is about 3.17 x 10 -4 (1 +- 18%). Neutron yield of neutron generator per pulse (10 7 /pulse) is measured by using the detector

Laminated Amorphous Silicon Neutron Detector (pre-print)

International Nuclear Information System (INIS)

McHugh, Harry; Branz, Howard; Stradins, Paul; Xu, Yueqin

2009-01-01

An internal R and D project was conducted at the Special Technologies Laboratory (STL) of National Security Technologies, LLC (NSTec), to determine the feasibility of developing a multi-layer boron-10 based thermal neutron detector using the amorphous silicon (AS) technology currently employed in the manufacture of liquid crystal displays. The boron-10 neutron reaction produces an alpha that can be readily detected. A single layer detector, limited to an approximately 2-micron-thick layer of boron, has a theoretical sensitivity of about 3%; hence a thin multi-layer device with high sensitivity can theoretically be manufactured from single layer detectors. Working with National Renewable Energy Laboratory (NREL), an AS PiN diode alpha detector was developed and tested. The PiN diode was deposited on a boron-10 coated substrate. Testing confirmed that the neutron sensitivity was nearly equal to the theoretical value of 3%. However, adhesion problems with the boron-10 coating prevented successful development of a prototype detector. Future efforts will include boron deposition work and development of integrated AS signal processing circuitry.

Multicomponent activation detector measurements of reactor neutron spectra

International Nuclear Information System (INIS)

Sandberg, J.; Aarnio, P. A.; Routti, J. T.

1984-01-01

Information on the neutron flux is required in many applications of research reactors, such as activation analysis or radiation damage measurements. Flux spectrum measurements are commonly carried out with activation foils. The reaction types used are threshold reactions in the fast energy region, resonance reactions in the intermediate region and neutron capture reactions with l/v-cross section in the thermal region. It has been shown that it is possible to combine several detector elements into homogeneous multicomponent detectors. The activities of all detector reaction products can be determined with a single gamma spectrum measurement. The multicomponent principle sets some restrictions on the choice of detector reactions, for example, each product nuclide may be produced in one reaction only. Separate multicomponent threshold and resonance detectors were designed for the fast and intermediate regions, respectively. The detectors were fabricated in polyethylene irradiation capsules or quartz glass ampoules, and they were irradiated in a cadmium cover. The detectors were succesfully used in the irradiation ring and in the core of a Triga reactor. The intermediate and fast neutron spectrum was unfolded with the least-squares unfolding program LOUHI. According to the preliminary results multicomponent activation detectors might constitute a convenient means for carrying out routine neutron spectrum measurements in research reactors. (orig.)

A new online detector for estimation of peripheral neutron equivalent dose in organ

Energy Technology Data Exchange (ETDEWEB)

Irazola, L., E-mail: leticia@us.es; Sanchez-Doblado, F. [Departamento de Fisiología Médica y Biofísica, Universidad de Sevilla, Sevilla 41009, Spain and Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41007 (Spain); Lorenzoli, M.; Pola, A. [Departimento di Ingegneria Nuclear, Politecnico di Milano, Milano 20133 (Italy); Bedogni, R. [Laboratori Nazionali di Frascati, Istituto Nazionale di Fisica Nucleare (INFN), Frascati Roma 00044 (Italy); Terrón, J. A. [Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41007 (Spain); Sanchez-Nieto, B. [Instituto de Física, Pontificia Universidad Católica de Chile, Santiago 4880 (Chile); Expósito, M. R. [Departamento de Física, Universitat Autònoma de Barcelona, Bellaterra 08193 (Spain); Lagares, J. I.; Sansaloni, F. [Centro de Investigaciones Energéticas y Medioambientales y Tecnológicas (CIEMAT), Madrid 28040 (Spain)

2014-11-01

Purpose: Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. Methods: To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. Results: The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. Conclusions: The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an

A new online detector for estimation of peripheral neutron equivalent dose in organ

International Nuclear Information System (INIS)

Irazola, L.; Sanchez-Doblado, F.; Lorenzoli, M.; Pola, A.; Bedogni, R.; Terrón, J. A.; Sanchez-Nieto, B.; Expósito, M. R.; Lagares, J. I.; Sansaloni, F.

2014-01-01

Purpose: Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. Methods: To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. Results: The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. Conclusions: The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an

Thermal and epithermal neutron fluence rate gradient measurements by PADC detectors in LINAC radiotherapy treatments-field

Energy Technology Data Exchange (ETDEWEB)

Barrera, M. T., E-mail: mariate9590@gmail.com; Barros, H.; Pino, F.; Sajo-Bohus, L. [Universidad Simón Bolívar, Nuclear Physics Laboratory, Sartenejas, Caracas (Venezuela, Bolivarian Republic of); Dávila, J. [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

2015-07-23

LINAC VARIAN 2100 is where energetic electrons produce Bremsstrahlung radiation, with energies above the nucleon binding energy (E≈5.5MeV). This radiation induce (γ,n) and (e,e’n) reactions mainly in the natural tungsten target material (its total photoneutron cross section is about 4000 mb in a energy range from 9-17 MeV). These reactions may occur also in other components of the system (e.g. multi leaf collimator). During radiation treatment the human body may receive an additional dose inside and outside the treated volume produced by the mentioned nuclear reactions. We measured the neutron density at the treatment table using nuclear track detectors (PADC-NTD). These covered by a boron-converter are employed, including a cadmium filter, to determine the ratio between two groups of neutron energy, i.e. thermal and epithermal. The PADC-NTD detectors were exposed to the radiation field at the iso-center during regular operation of the accelerator. Neutron are determined indirectly by the converting reaction {sup 10}B(n,α){sup 7}Li the emerging charged particle leave their kinetic energy in the PADC forming a latent nuclear track, enlarged by chemical etching (6N, NaOH, 70°C). Track density provides information on the neutron density through calibration coefficient (∼1.6 10{sup 4} neutrons /track) obtained by a californium source. We report the estimation of the thermal and epithermal neutron field and its gradient for photoneutrons produced in radiotherapy treatments with 18 MV linear accelerators. It was obsered that photoneutron production have higher rate at the iso-center.

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

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.)

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

Development of neutron detectors for neutron scattering experiments

Energy Technology Data Exchange (ETDEWEB)

Moon, Myungkook; Kim, Jongyul; Kim, Jeong ho; Lee, Suhyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Changhwy [Korea Research Institute of Ships and Ocean Engineering, Daejeon (Korea, Republic of)

2015-10-15

Various kinds of detectors are used in accordance with the experimental purpose, such as zero dimensional detector, 1-D or 2-D position-sensitive detectors. Most of neutron detectors use He-3 gas because of its high neutron sensitivity. Since the He-3 supply shortage took place in early 2010, various He-3 alternative detectors have been developed even for the other neutron application. We have developed a new type alternative detector on the basis of He-3 detector technology. Although B- 10 has less neutron detection efficiency compared with He-3, it can be covered by the use of multiple B-10 layers. In this presentation, we would like to introduce the neutron detectors under development and developed detectors. Various types of detector were successfully developed and result of the technical test performance is promising. Even though the detection efficiency of the B-10 detector lower than He-3 one, the continuous research and development is needed for currently not available He-3.

Development of gamma-ray-suppression type of small-sized neutron detector based on a 6Li-glass scintillator

International Nuclear Information System (INIS)

Matsumoto, T.; Harano, H.; Shimoyama, T.; Kudo, K.; Uritani, A.

2005-01-01

A small-sized thermal neutron detector based on a 6 Li-glass scintillator and a plastic optical fiber was developed for measurement of a dose distribution of thermal neutrons in a thermal neutron standard field. A contribution of gamma rays can not be neglected in the neutron measurement with this detector, although the 6 Li-glass scintillator can be distinguishable for the neutrons and the gamma rays by difference of each pulse height. Moreover, to reduce an uncertainty of neutron counts caused by the gamma ray background around a discrimination level, we suggested a gamma-ray-suppression type of small-sized thermal neutron detector with a 6 Li-glass scintillator, a hollow CsI(Tl) scintillator and plastic optical fibers. The detector can reject signals due to the gamma rays with an anti-coincidence method. In the present paper, we evaluated an ability of a gamma-ray suppression of the detector using the EGS4 electron-photon transport Monte-Carlo code with the PRESTA routine. As the results, the sufficient gamma-ray suppression effect was shown. (author)

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)

Layered semiconductor neutron detectors

Science.gov (United States)

Mao, Samuel S; Perry, Dale L

2013-12-10

Room temperature operating solid state hand held neutron detectors integrate one or more relatively thin layers of a high neutron interaction cross-section element or materials with semiconductor detectors. The high neutron interaction cross-section element (e.g., Gd, B or Li) or materials comprising at least one high neutron interaction cross-section element can be in the form of unstructured layers or micro- or nano-structured arrays. Such architecture provides high efficiency neutron detector devices by capturing substantially more carriers produced from high energy .alpha.-particles or .gamma.-photons generated by neutron interaction.

New thermal neutron solid-state electronic detector based on HgI2 crystals

International Nuclear Information System (INIS)

Melamud, M.; Burshtein, Z.

1983-07-01

We describe the development of a new solid-state electronic neutron detector, based on HgI 2 single crystals. Incident neutrons are absorbed in high neutron absorbing foils, such as cadmium or gadolinium, which are placed in front of a HgI 2 detector. Gamma rays, emitted as a result of the neutron absorbtion, are then absorbed in the HgI 2 , generating free charge carriers, which are collected by the electric field. The advantage of this system lies in it's manufacturing simplicity, low weight and small physical dimensions, compared to gas-filled conventional neutron detectors. The disadvantage is that the system does not discriminate between gamma rays and neutrons. A method to minimize this disadvantage is pointed out. It is as well possible to count neutrons by direct exposure of the HgI 2 to neutrons. The neutron-to-gamma transformation in that case takes place by the material nuclei themselves. This method, however, is impractical due to the interference of delayed radioactivity whose origin are 129 I nuclei. They are generated from 128 I by absorbing a neutron, and decay with a 25 min half lifetime involving gamma emissions. (author)

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

International Nuclear Information System (INIS)

Angelone, M.; Klix, A.; Pillon, M.; Batistoni, P.; Fischer, U.; Santagata, A.

2014-01-01

Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

Energy Technology Data Exchange (ETDEWEB)

Angelone, M., E-mail: maurizio.angelone@enea.it [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Klix, A. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pillon, M.; Batistoni, P. [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Fischer, U. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Santagata, A. [ENEA C.R. Casaccia, via Anguillarese Km. 1,300, 00100 Roma (Italy)

2014-10-15

Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.

ATLAS-TPX: a two-layer pixel detector setup for neutron detection and radiation field characterization

International Nuclear Information System (INIS)

Bergmann, B.; Caicedo, I.; Pospisil, S.; Vykydal, Z.; Leroy, C.

2016-01-01

A two-layer pixel detector setup (ATLAS-TPX), designed for thermal and fast neutron detection and radiation field characterization is presented. It consists of two segmented silicon detectors (256 × 256 pixels, pixel pitch 55 μm, thicknesses 300 μm and 500 μm) facing each other. To enhance the neutron detection efficiency a set of converter layers is inserted in between these detectors. The pixelation and the two-layer design allow a discrimination of neutrons against γs by pattern recognition and against charged particles by using the coincidence and anticoincidence information. The neutron conversion and detection efficiencies are measured in a thermal neutron field and fast neutron fields with energies up to 600 MeV. A Geant4 simulation model is presented, which is validated against the measured detector responses. The reliability of the coincidence and anticoincidence technique is demonstrated and possible applications of the detector setup are briefly outlined.

To the use of bubble detectors in personal neutron dosimetry

Energy Technology Data Exchange (ETDEWEB)

Spurny, F; Vlcek, B [Academy of the Sciences of Czech Republic, Prague (Czech Republic). Nuclear Physics Institute, Department of Radiation Dosimetry

1996-12-31

In this paper the commercially available bubble neutron detectors (bubble damage neutron detectors (BDNDs*) from Bubble Technology Industries, Chalk River; and superheated drop detectors (SDDs*) from APFEL Industries, New Haven) for lowest limit of detection of an individual neutron dosimeter were tested. They were tested with the different neutron sources. BDNDs* tested had the sensitivity about 1 bubble per 1 Sv of H*(10) of AmBe neutrons, they were evaluated by eye counting (20 to 30 bubbles per detector). Two types of reusable BDNDs* were tested: BD-100R without and with temperature compensation, both with neutron energy threshold about 100 keV. SDDs* tested had the sensitivity about 3 bubbles per 1 {mu}Sv of H*10 from AmBe neutrons, they were evaluated using APFEL Survey Meter Model 202. SDDs* with three different energy thresholds have been used: 0.1, 1 and 6 MeV. For energetical dependence of BDNDs* the general conclusions were formulated in the following way: (1) With the exception of thermal neutron source SIGMA (50% of H*(10) from thermal neutrons) and high energy reference fields there is a reasonable agreement of data measured with BDNDs* and expected values; (2) the new lots to have a little different energetic dependence. The relative responses for `soft` fields are for them systematically higher than for previous samples. The response to energies between 0.01 and 1 MeV is for these lots relatively higher. (3) The underestimation of high energy neutrons is typical for any LET-threshold type detectors.It should be kept in mind when BDNDs* are used as dosemeters in high energy neutron environment. For energetical dependence of SDDs* was concluded: (1) The energetical dependence of SDD 100 is comparable with the dependencies of BD-100R and PND, the underestimation of high energy neutrons included; (2) The use of SDD with different energy thresholds can provide interesting spectrometric information; (Abstract Truncated)

Self-Powered Neutron Detector Qualification for Absolute On-Line In-Pile Neutron Flux Measurements in BR2

Science.gov (United States)

Vermeeren, L.; Wéber, M.

2003-06-01

A set of ten Self-Powered Neutron Detectors with Co, Rh and Ag emitters has been irradiated in several channels of the BR2 research reactor at SCK•CEN aiming at a comparison of their performance as thermal neutron flux detectors under various conditions. To allow for a correct interpretation of their signals, all detector sensitivity contributions (prompt and delayed) were calculated using a dedicated Monte Carlo model. The various contributions were also measured separately; the agreement between calculated and experimental data, including data from activation dosimetry, was excellent. Detailed neutron flux profiles were obtained from the SPND data, after correction for the finite detector lengths and for the slow response of delayed SPNDs.

Shielded regenerative neutron detector

International Nuclear Information System (INIS)

Terhune, J.H.; Neissel, J.P.

1978-01-01

An ion chamber type neutron detector is disclosed which has a greatly extended lifespan. The detector includes a fission chamber containing a mixture of active and breeding material and a neutron shielding material. The breeding and shielding materials are selected to have similar or substantially matching neutron capture cross-sections so that their individual effects on increased detector life are mutually enhanced

Development of real time personal neutron dosimeter with two silicon detectors

Energy Technology Data Exchange (ETDEWEB)

Nakamura, T.; Tsujimura, N. [Tohoku Univ., Cyclotron and Radioisotope Center, Aoba, Aramaki, Aoba-ku (Japan); Yamano, T. [Tokyo Factory, Fuji Electric Co. Ltd., Tokyo (Japan)

1992-07-01

We developed a real time personal neutron dosimeter by using two types of silicon p-n junction detectors, thermal neutron sensor and fast neutron sensor. The thermal neutron sensor which is {sup 10}B doped n-type silicon with a polyethylene radiator mainly counts neutrons of energy front thermal to I MeV, and the fast neutron sensor which is p-type silicon with a polyethylene radiator is sensitive to neutrons above I MeV. The neutron sensitivity measurements revealed that the dosimeter has a rather flat response for dose equivalent from thermal to 15 MeV, excluding a drop from 50 keV to I MeV. In order to get conversion factor from counts to dose equivalent as accurately as possible, we performed the field test of the dosimeter calibration in several neutron-generating fields. By introducing the two-group dose estimation method, this dosimeter can give the neutron dose equivalent within about 50% errors. (author)

Optimization of a neutron detector design using adjoint transport simulation

International Nuclear Information System (INIS)

Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

2012-01-01

A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

Comparison of various stopping gases for 3He-based position sensitive neutron detectors

International Nuclear Information System (INIS)

Doumas, A.; Smith, G.C.

2012-01-01

A range of solid state, scintillator and gas based detectors are being developed for use at the next generation of high flux neutron facilities. Since gas detectors are expected to continue to play a key role in future specific thermal neutron experiments, a comparison of the performance characteristics of prospective stopping gases is beneficial. Gas detectors typically utilize the reaction 3 He(n,p)t to detect thermal neutrons; the 3 He gas is used in a mixture containing a particular stopping gas in order to maintain relatively short ranges for the proton and triton pair emitted from the n- 3 He reaction. Common stopping gases include hydrocarbons (e.g. propane), carbon tetrafluoride, and noble gases such as argon and xenon. For this study, we utilized the Monte Carlo simulation code “Stopping and Range of Ions in Matter” to analyze the expected behavior of argon, xenon, carbon dioxide, difluoroethane and octafluoropropane as stopping gases for thermal neutron detectors. We also compare these findings to our previously analyzed performance of propane, butane and carbon tetrafluoride. A discussion of these gases includes their behavior in terms of proton and triton range, ionization distribution and straggle.

Comparison of various stopping gases for 3He-based position sensitive neutron detectors

Science.gov (United States)

Doumas, A.; Smith, G. C.

2012-05-01

A range of solid state, scintillator and gas based detectors are being developed for use at the next generation of high flux neutron facilities. Since gas detectors are expected to continue to play a key role in future specific thermal neutron experiments, a comparison of the performance characteristics of prospective stopping gases is beneficial. Gas detectors typically utilize the reaction 3He(n,p)t to detect thermal neutrons; the 3He gas is used in a mixture containing a particular stopping gas in order to maintain relatively short ranges for the proton and triton pair emitted from the n-3He reaction. Common stopping gases include hydrocarbons (e.g. propane), carbon tetrafluoride, and noble gases such as argon and xenon. For this study, we utilized the Monte Carlo simulation code "Stopping and Range of Ions in Matter" to analyze the expected behavior of argon, xenon, carbon dioxide, difluoroethane and octafluoropropane as stopping gases for thermal neutron detectors. We also compare these findings to our previously analyzed performance of propane, butane and carbon tetrafluoride. A discussion of these gases includes their behavior in terms of proton and triton range, ionization distribution and straggle.

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)

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.

Hexagonal boron nitride neutron detectors with high detection efficiencies

Science.gov (United States)

Maity, A.; Grenadier, S. J.; Li, J.; Lin, J. Y.; Jiang, H. X.

2018-01-01

Neutron detectors fabricated from 10B enriched hexagonal boron nitride (h-10BN or h-BN) epilayers have demonstrated the highest thermal neutron detection efficiency among solid-state neutron detectors to date at about 53%. In this work, photoconductive-like vertical detectors with a detection area of 1 × 1 mm2 were fabricated from 50 μm thick free-standing h-BN epilayers using Ni/Au and Ti/Al bilayers as ohmic contacts. Leakage currents, mobility-lifetime (μτ) products under UV photoexcitation, and neutron detection efficiencies have been measured for a total of 16 different device configurations. The results have unambiguously identified that detectors incorporating the Ni/Au bilayer on both surfaces as ohmic contacts and using the negatively biased top surface for neutron irradiation are the most desired device configurations. It was noted that high growth temperatures of h-10BN epilayers on sapphire substrates tend to yield a higher concentration of oxygen impurities near the bottom surface, leading to a better device performance by the chosen top surface for irradiation than by the bottom. Preferential scattering of oxygen donors tends to reduce the mobility of holes more than that of electrons, making the biasing scheme with the ability of rapidly extracting holes at the irradiated surface while leaving the electrons to travel a large average distance inside the detector at a preferred choice. When measured against a calibrated 6LiF filled micro-structured semiconductor neutron detector, it was shown that the optimized configuration has pushed the detection efficiency of h-BN neutron detectors to 58%. These detailed studies also provided a better understanding of growth-mediated impurities in h-BN epilayers and their effects on the charge collection and neutron detection efficiencies.

Monitor for reactor neutron detector

International Nuclear Information System (INIS)

Shirakami, Hisayuki; Shibata, Masatoshi

1992-01-01

The device of the present invention judges as to whether a neutron detector is normal or not while considering the change of indication value depending on the power change of a reactor core. That is, the device of the present invention comprises a standard value setting device for setting the standard value for calibrating the neutron detector and an abnormality judging device for comparing the standard value with a measured value of the neutron detector and judging the abnormality when the difference is greater than a predetermined value. The measured value upon initialization of each of the neutron detectors is determined as a quasi-standard value. An average value of the difference between the measured value and the quasi-standard value of a plurality of effective neutron detectors at a same level for the height of the reactor core is multiplied to a power rate based on the reactor core power at a position where the neutron detector is disposed upon calibration. The value obtained by adding the multiplied value and the quasi-standard value is determined as a standard value. The abnormality judging device compares the standard value with the measured value of the neutron detector and, if the difference is greater than a predetermined value, the neutron detector is determined as abnormal. As a result, judgement can be conducted more accurately than conventional cases. (I.S.)

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...

Neutron--neutron logging

International Nuclear Information System (INIS)

Allen, L.S.

1977-01-01

A borehole logging tool includes a steady-state source of fast neutrons, two epithermal neutron detectors, and two thermal neutron detectors. A count rate meter is connected to each neutron detector. A first ratio detector provides an indication of the porosity of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two epithermal neutron detectors. A second ratio detector provides an indication of both porosity and macroscopic absorption cross section of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two thermal neutron detectors. By comparing the signals of the two ratio detectors, oil bearing zones and salt water bearing zones within the formation being logged can be distinguished and the amount of oil saturation can be determined. 6 claims, 2 figures

Physical motivations for thermal detectors

International Nuclear Information System (INIS)

Fiorini, E.

1993-01-01

Low temperature bolometers can be complementary and sometimes superior to open-quotes classicalclose quotes ionization detectors in many experiments without accelerators in nuclear, subnuclear and astroparticle physics. After a short review of the open-quotes toolsclose quotes that cryogenics offer for the detection of particles the author first considers a few practical applications of bolometers in the spectroscopy of α, γ and X rays, in the detection of neutrons, and in measurements of weak radioactive contaminations. Searches with this technique on single and double beta decay, of which some are already being carried out, are then considered and discussed. The various properties which make thermal detectors particularly suitable for searches on dark matter are reviewed, stressing the potentiality of this technique. The promising, but still far, potentiality of thermal detectors in solar neutrino experiments is finally discussed

Self powered neutron detectors

International Nuclear Information System (INIS)

Passe, J.; Petitcolas, H.; Verdant, R.

1975-01-01

The self-powered neutron detectors (SPND) enable to measure continuously high fluxes of thermal neutrons. They are particularly suitable for power reactor cores because of their robustness. Description of two kinds of SPND's characterized by the electrical current production way is given here: the first SPND's which present a V, Ag or Rh emitter are sensitive enough but they offer a few minute delay time: the second SPND's which are depending on the gamma activation have a short delay time. The emitter is made of Co or Pt. In any case, the signal is linear with reaction rates. Finally, the applications are briefly repeated here: irradiation facility monitor in research reactors, and flux map and space instability control in power reactors [fr

Analysis of the sensitivity concept of self-powered neutron detector (SPND)

International Nuclear Information System (INIS)

Moreira, O.; Lescano, H.

2012-01-01

Self powered neutron detectors (SPND) are widely used to monitor the neutron flux, either in nuclear as in irradiation facilities and medical treatments. However, the physical meaning of the parameter that is used to relate the detector signal (an electrical current) with the neutron flux, i.e., the sensitivity of the detector, has not been sufficiently analyzed. Since the definition of sensitivity, ε=i/φ is calculated for particular reactor conditions, i.e., for thermal neutrons at room temperature, it does not take into account the deviation originated from other conditions of temperature (above ambient), as found for example in nuclear power plants. In this work we calculated the microscopic cross section weighted with the neutron flux, defined in the usual way. This weighted microscopic cross section reveals the no proportionality between the absorption rate and the neutron flux, exhibiting the problem that the SPND current signal has to properly represent the neutron flux (author)

Neutron Fluence Evaluation using an Am-Be Neutron Sources Assembly and P ADC Detectors

International Nuclear Information System (INIS)

Seddik, U.

2008-01-01

An assembly of four 241 Am-Be sources has been constructed at Nuclear Reactions Unit (NRU) of Nuclear Research Center (NRU) to perform analysis of different materials using thermal and fast neutrons. In the present paper, we measure the value of transmittance (T) in percentage of etched CR-39 detectors using a spectrophotometer at different neutron fluences ,to relate the transmittance of the detector with the neutron fluence values. The exposed samples to neutrons with accumulated fluence of order between 10 10 and 10 12 cm -2 were etched for 15 time intervals between 10-600 min in 6.25 N NaOH at 70 degree C. The etched samples were analyzed using Tech 8500 II spectrophotometer. A trend of the sample transmission and the etching time is observed which is different for each fluence value. A linear relation between the transmittance decay constant and the neutron fluence is observed which could be used as a calibration to determine unknown neutron fluence

The influence of rhodium burn-up on the sensitivity of rhodium self-powered neutron detectors

International Nuclear Information System (INIS)

Erben, O.

1980-01-01

Depression and self-shielding coefficients are presented for thermal and epithermal neutron flux densities. Functions are shown describing the distribution of beta particle sources on the emitter cross section for 0 to 50% rhodium burnup. The values are calculated of detector sensitivity to thermal and epithermal neutron flux densities for the said burnup for main types of rhodium SPN detectors made by SODERN. (J.B.)

Studies of the dependence on oxidation thermal processes of effects on the electrical properties of silicon detectors by fast neutron radiation

International Nuclear Information System (INIS)

Li, Zheng; Kraner, H.W.

1991-11-01

High resistivity silicon detectors along with MOS capacitors made on five silicon dioxides with different thermal conditions (975 degree C to 1200 degree C) have been exposed to fast neutron irradiation up to the fluence of a few times 10 14 n/cm 2 . New measurement techniques such as capacitance-voltage (C-V) of MOS capacitors and current-voltage (I-V) of back to back diodes (p + -n - - p + if n - is not inverted to p) or resistors (p + -p-p + if inverted) have been introduced in this study in monitoring the possible type-inversion (n→p) under high neutron fluence. No type-inversion in the material underneath SiO 2 and the p + contact has been observed so far in this work for detectors made on the five oxides up to the neutron fluence of a few times 10 13 n/cm 2 . However, it has been found that detectors made on higher temperature oxides (T≤ 1100 degree C) exhibit less leakage current increase at high neutron fluence (φ ≤ 10 13 n/cm 2 )

Using a Borated Panel to Form a Dual Neutron-Gamma Detector

Energy Technology Data Exchange (ETDEWEB)

Scott Wilde; Raymond Keegan

2008-06-20

A borated polyethylene plane placed between a neutron source and a gamma spectrometer is used to form a dual neutron-gamma detection system. The polyethylene thermalizes the source neutrons so that they are captured by {sup 10}B to produce a flux of 478 keV gamma-rays that radiate from the plane. This results in a buildup of count rate in the detector over that from a disk of the same diameter as the detector crystal (same thickness as the panel). Radiation portal systems are a potential application of this technique.

Preliminary studies of microchannel plate photomultiplier tube neutron detectors for flight test applications

International Nuclear Information System (INIS)

Dolan, K.W.

1978-10-01

Electrical, mechanical, thermal, and neutron response data indicate that microchannel plate photomultiplier tubes are viable candidates as miniature, ruggedized neutron detectors for flight test applications in future weapon systems

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)

ESR-dosimetry in thermal and epithermal neutron fields for application in boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Schmitz, Tobias

2016-01-22

Dosimetry is essential for every form of radiotherapy. In Boron Neutron Capture Therapy (BNCT) mixed neutron and gamma fields have to be considered. Dose is deposited in different neutron interactions with elements in the penetrated tissue and by gamma particles, which are always part of a neutron field. The therapeutic dose in BNCT is deposited by densely ionising particles, originating from the fragmentation of the isotope boron-10 after capture of a thermal neutron. Despite being investigated for decades, dosimetry in neutron beams or fields for BNCT remains complex, due to the variety in type and energy of the secondary particles. Today usually ionisation chambers combined with metal foils are used. The applied techniques require extensive effort and are time consuming, while the resulting uncertainties remain high. Consequently, the investigation of more effective techniques or alternative dosimeters is an important field of research. In this work the possibilities of ESR-dosimeters in those fields have been investigated. Certain materials, such as alanine, generate stable radicals upon irradiation. Using Electron Spin Resonance (ESR) spectrometry the amount of radicals, which is proportional to absorbed dose, can be quantified. Different ESR detector materials have been irradiated in the thermal neutron field of the research reactor TRIGA research reactor in Mainz, Germany, with five setups, generating different secondary particle spectra. Further irradiations have been conducted in two epithermal neutron beams. The detector response, however, strongly depends on the dose depositing particle type and energy. It is hence necessary to accompany measurements by computational modelling and simulation. In this work the Monte Carlo code FLUKA was used to calculate absorbed doses and dose components. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using amorphous track models. For the simulation, detailed models of

Design and fabrication of a novel self-powered solid-state neutron detector

Science.gov (United States)

LiCausi, Nicholas

There is a strong interest in intercepting special nuclear materials (SNM) at national and international borders and ports for homeland security applications. Detection of SNM such as U and Pu is often accomplished by sensing their natural or induced neutron emission. Such detector systems typically use thermal neutron detectors inside a plastic moderator. In order to achieve high detection efficiency gas filled detectors are often used; these detectors require high voltage bias for operation, which complicates the system when tens or hundreds of detectors are deployed. A better type of detector would be an inexpensive solid-state detector that can be mass-produced like any other computer chip. Research surrounding solid-state detectors has been underway since the late 1990's. A simple solid-state detector employs a planar solar-cell type p-n junction and a thin conversion material that converts incident thermal neutrons into detectable alpha-particles and 7Li ions. Existing work has typically used 6LiF or 10B as this conversion layer. Although a simple planar detector can act as a highly portable, low cost detector, it is limited to relatively low detection efficiency (˜10%). To increase the efficiency, 3D perforated p-i-n silicon devices were proposed. To get high efficiency, these detectors need to be biased, resulting in increased leakage current and hence detector noise. In this research, a new type of detector structure was proposed, designed and fabricated. Among several detector structures evaluated, a honeycomb-like silicon p-n structure was selected, which is filled with natural boron as the neutron converter. A silicon p+-n diode formed on the thin silicon wall of the honeycomb structure detects the energetic alpha-particles emitted from the boron conversion layer. The silicon detection layer is fabricated to be fully depleted with an integral step during the boron filling process. This novel feature results in a simplified fabrication process. Three

Comparison of various stopping gases for {sup 3}He-based position sensitive neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Doumas, A. [United States Merchant Marine Academy, Steamboat Road, Kings Point, NY 11024 (United States); Smith, G.C., E-mail: gsmith@bnl.gov [Instrumentation Division, Brookhaven National Laboratory, Upton, NY 11973 (United States)

2012-05-21

A range of solid state, scintillator and gas based detectors are being developed for use at the next generation of high flux neutron facilities. Since gas detectors are expected to continue to play a key role in future specific thermal neutron experiments, a comparison of the performance characteristics of prospective stopping gases is beneficial. Gas detectors typically utilize the reaction {sup 3}He(n,p)t to detect thermal neutrons; the {sup 3}He gas is used in a mixture containing a particular stopping gas in order to maintain relatively short ranges for the proton and triton pair emitted from the n-{sup 3}He reaction. Common stopping gases include hydrocarbons (e.g. propane), carbon tetrafluoride, and noble gases such as argon and xenon. For this study, we utilized the Monte Carlo simulation code 'Stopping and Range of Ions in Matter' to analyze the expected behavior of argon, xenon, carbon dioxide, difluoroethane and octafluoropropane as stopping gases for thermal neutron detectors. We also compare these findings to our previously analyzed performance of propane, butane and carbon tetrafluoride. A discussion of these gases includes their behavior in terms of proton and triton range, ionization distribution and straggle.

New electronically black neutron detectors

International Nuclear Information System (INIS)

Drake, D.M.; Feldman, W.C.; Hurlbut, C.

1986-03-01

Two neutron detectors are described that can function in a continuous radiation background. Both detectors identify neutrons by recording a proton recoil pulse followed by a characteristic capture pulse. This peculiar signature indicates that the neutron has lost all its energy in the scintillator. Resolutions and efficiencies have been measured for both detectors

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

Directional epithermal neutron detector

International Nuclear Information System (INIS)

Givens, W.W.; Mills, W.R. Jr.

1986-01-01

A borehole tool for epithermal neutron die-away logging of subterranean formations surrounding a borehole is described which consists of: (a) a pulsed source of fast neutrons for irradiating the formations surrounding a borehole, (b) at least one neutron counter for counting epithermal neutrons returning to the borehole from the irradiated formations, (c) a neutron moderating material, (d) an outer thermal neutron shield providing a housing for the counter and the moderating material, (e) an inner thermal neutron shield dividing the housing so as to provide a first compartment bounded by the inner thermal neutron shield and a first portion of the outer thermal neutron shield and a second compartment bounded by the inner thermal neutron shield and a second portion of the outer thermal neutron shield, the counter being positioned within the first compartment and the moderating material being positioned within the second compartment, and (f) means for positioning the borehole tool against one side of the borehole wall and azimuthally orienting the borehole tool such that the first chamber is in juxtaposition with the borehole wall, the formation epithermal neutrons penetrating into the first chamber through the first portion of the outer thermal neutron shield are detected by the neutron counter for die-away measurement, thereby maximizing the directional sensitivty of the neutron counter to formation epithermal neutrons, the borehole fluid epithermal neutrons penetrating into the second chamber through the second chamber through the second portion of the outer thermal neutron shield are largely slowed down and lowered in energy by the moderating material and absorbed by the inner thermal neutron shield before penetrating into the first chamber, thereby minimizing the directional sensitivity of the neutron counter to borehole fluid epithermal neutrons

A novel track density measurement method by thermal neutron activation of DYECETs

International Nuclear Information System (INIS)

Sohrabi, M.; Mahdi, Sh.

1995-01-01

A novel track density evaluation method based on thermal neutron activation of some elements of dyed electrochemically etched tracks (DYECETs) of charged particles in detectors like polycarbonate (PC) followed by measurements of gamma activity of the activated detectors is introduced. In this method, the tracks of charged particles like fast neutron-induced recoils in PC detectors were electrochemically etched, dyed by ''QuicDYECET'' methods as recently introduced by us, activated by thermal neutrons and counted for gamma activity determination to be correlated with track density. The activities of elements such as bromine-82 ( 82 Br) and sodium-24 ( 24 Na) on dyes such as Eosin Yellowish, Eosin Bluish, etc. determined by a hyper-pure germanium detector, were found to be in good correlation with DYECET density and thus particle fluence or dose. The effects of different types of dyes and their elements, dye concentration, neutron fluences and ECE durations on the DYECET density responses were studied. This new development is a method of scientific interest, potentially possessing some interesting features, as an alternative method for ECE track density determination using a gamma activity measuring system. It also has the drawback of being applicable only in centres having thermal neutron facilities. The results of the above studies are presented and discussed. (Author)

Method and apparatus for formation logging using position sensitive neutron detectors

International Nuclear Information System (INIS)

Gadken, L.L.

1986-01-01

This patent describes a method for logging earth formations using position sensitive neutron detectors. The method consists of: 1) Irradiation of earth formations in the vicinity of a well borehole with a source of fast neutrons. 2) At four longitudinally spaced distances from the neutron source in the borehole, the epithermal neutron population is detected. Each of the four separate populations is detected in an epithermally sensitive and substantially thermally insensitive portion of the same position sensitive neutron detector. A representative signal from each is then individually generated. 3) First, second, third, and fourth neutron population representative signals are combined. They derive a simultaneous measurement signal. This signal is functionally related to the porosity and also a signal functionally related to a neutron characteristic length of the earth formations in the vicinity of the borehole

Neutron sensitivity of prompt-response self-powered neutron detectors and the interval rule

International Nuclear Information System (INIS)

Molina Avila, J.; Carmolopes, M.

1989-01-01

This paper is devoted to the calculation of thermal s th and epithermal s epi sensitivities of cobalt prompt-response Self-Powered Neutron Detectors (SPNDs). The thermal sensitivity was obtained for a Maxwellian neutron field, and the effect of scattering on the self-shielding correction was taken into consideration in the second-collision approximation. The dependence of s th on the emitter radius R was studied in a wide region of R (0.025 to 0.2 cm). The differential and global epithermal sensitivities were calculated using a simple expression for the first-collision neutron absorption probability. Finally, a criterion to evaluate the accuracy of the parameters of the model was established in the form of some Interval Rule which is very sensitive to the radial dependence of the flux perturbation correction and other parameters of the model in both the thermal and epithermal regions

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

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.

Modeling of displacement damage in silicon carbide detectors resulting from neutron irradiation

Science.gov (United States)

Khorsandi, Behrooz

There is considerable interest in developing a power monitor system for Generation IV reactors (for instance GT-MHR). A new type of semiconductor radiation detector is under development based on silicon carbide (SiC) technology for these reactors. SiC has been selected as the semiconductor material due to its superior thermal-electrical-neutronic properties. Compared to Si, SiC is a radiation hard material; however, like Si, the properties of SiC are changed by irradiation by a large fluence of energetic neutrons, as a consequence of displacement damage, and that irradiation decreases the life-time of detectors. Predictions of displacement damage and the concomitant radiation effects are important for deciding where the SiC detectors should be placed. The purpose of this dissertation is to develop computer simulation methods to estimate the number of various defects created in SiC detectors, because of neutron irradiation, and predict at what positions of a reactor, SiC detectors could monitor the neutron flux with high reliability. The simulation modeling includes several well-known---and commercial---codes (MCNP5, TRIM, MARLOWE and VASP), and two kinetic Monte Carlo codes written by the author (MCASIC and DCRSIC). My dissertation will highlight the displacement damage that may happen in SiC detectors located in available positions in the OSURR, GT-MHR and IRIS. As extra modeling output data, the count rates of SiC for the specified locations are calculated. A conclusion of this thesis is SiC detectors that are placed in the thermal neutron region of a graphite moderator-reflector reactor have a chance to survive at least one reactor refueling cycle, while their count rates are acceptably high.

Simulation of a silicon neutron detector coated with TiB2 absorber

International Nuclear Information System (INIS)

Krapohl, D; Nilsson, H-E; Petersson, S; Slavicek, T; Thungström, G; Pospisil, S

2012-01-01

Neutron radiation cannot be directly detected in semiconductor detectors and therefore needs converter layers. Planar clean-room processing can be used in the manufacturing process of semiconductor detectors with metal layers to produce a cost-effective device. We used the Geant4 Monte-Carlo toolkit to simulate the performance of a semiconductor neutron detector. A silicon photo-diode was coated with vapour deposited titanium, aluminium thin films and a titaniumdiboride (TiB 2 ) neutron absorber layer. The neutron capture reaction 10B(n, alpha)7Li is taken advantage of to create charged particles that can be counted. Boron-10 has a natural abundance of about SI 19.8%. The emitted alpha particles are absorbed in the underlying silicon detector. We varied the thickness of the converter layer and ran the simulation with a thermal neutron source in order to find the best efficiency of the TiB 2 converter layer and optimize the clean room process.

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.)

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....

Recent Developments in GEM-Based Neutron Detectors

International Nuclear Information System (INIS)

Saenboonruang, K.

2014-01-01

The gas electron multiplier (GEM) detector is a relatively new gaseous detector that has been used for less than 20 years. Since the discovery in 1997 by F. Sauli, the GEM detector has shown excellent properties including high rate capability, excellent resolutions, low discharge probability, and excellent radiation hardness. These promising properties have led the GEM detector to gain popularity and attention amongst physicists and researchers. In particular, the GEM detector can also be modified to be used as a neutron detector by adding appropriate neutron converters. With properties stated above and the need to replace the expensive 3 He-based neutron detectors, the GEM-based neutron detector will be one of the most powerful and affordable neutron detectors. Applications of the GEM-based neutron detectors vary from researches in nuclear and particle physics, neutron imaging, and national security. Although several promising progresses and results have been shown and published in the past few years, further improvement is still needed in order to improve the low neutron detection efficiency (only a few percent) and to widen the possibilities for other uses.

Passive detectors for neutron fluence measurement

International Nuclear Information System (INIS)

Holt, P.D.

1985-01-01

The use of neutron activation detectors (slow neutron detectors and threshold detectors) and fission track detectors for radiological protection purposes, principally in criticality dosimetry, dosimetry of pulsed accelerators and calibration of neutron fluxes is discussed. References are given to compilations of cross sections. For the determination of the activity induced, either beta ray or gamma ray counting may be used. For beta-ray counting, thin foils are usually necessary which result in low neutron sensitivity. When fission track detectors are used, it is necessary to know the efficiency of track registration. Alternatively, a detector-counter system may be calibrated by exposure to a known flux of monoenergetic neutrons. Usually, the sensitivity of activation detectors is low because small foils are used. For criticality dosimetry, calibration work and shielding studies on accelerators, low sensitivity is acceptable. However, there are some instances where, by the use of long integration times, or very large quantities of detector material with gamma ray detection, neutron fluences in operational areas have been measured. (author)

Design and fabrication of a self-powered neutron detector

International Nuclear Information System (INIS)

Garcia Garcia, Florencio.

1979-01-01

Self powered neutron detectors are becoming more and more popular in reactor instrumentation. A fast response detector of this type was made at the Reactor Division, La Reina Nuclear Center in Santiago. Cobalt wire was the emitter, teflon the insulator and a stainless steel tubing was the collector. The overall dimensions of the detector are 6 mms diameter and 700 mms length. The irradiation tests, carried out at the Center's 5 Mw research reactor showed a very reasonably linear relation between current supplied by the detector and thermal neutron flux, over a range extending from 10 10 to 10 13 n/cm 2 x seg. These tests also showed a good agreement between calculated and measured current. The models used for the calculation of current are fully explained and they include some improvements over those that have been published recently. An important conclusion for the case of the cobalt detectors is that the wire's diameter must be at least 1 mm. in order to have a neutron induced current bigger than the parasitic components generated by indirect processes. Calculations for other emitters such as vanadium, silver and rhodum are also included. (EC)

Scintillation neutron detector with dynamic threshold

International Nuclear Information System (INIS)

Kornilov, N.; Massey, T.; Grimes, S.

2014-01-01

Scintillation neutron detectors with hydrogen are a common tool for neutron spectroscopy. They provide good time resolution, neutron-gamma discrimination and high efficiency of neutron counting. The real open problems connected with application of these detectors are in the energy range >10 MeV. There are no standard neutron spectra known with high accuracy for this energy range. Therefore, traditional methods for experimental investigation of the efficiency function fail for these neutrons. The Monte Carlo simulation cannot provide reasonable accuracy due to unknown characteristics of the reactions for charged particle production (p, α and so on, light output, reaction cross-sections). The application of fission chamber with fissile material as a neutron detector did not help to solve the problem. We may avoid many problems if we use the traditional neutron detector with non-traditional data analysis. In this report we give main relations, and demonstrate the method for Cf-source. Experimental detector efficiency is compared with MC simulation. (authors)

Black and grey neutron detectors

International Nuclear Information System (INIS)

Gabbard, F.

1977-01-01

Recent progress in the development and use of ''black'' and ''grey'' detectors is reviewed. Such detectors are widely used for counting neutrons in (p,n) and (α,n) experiments and in neutron cross section measurements. Accuracy of each detector is stressed. 19 figures

Measurements of thermal and fast neutron fluxes at the TRIGA reactor

International Nuclear Information System (INIS)

Zerdin, F.; Grabovsek, Z.; Klinc, T.; Solinc, H.

1966-01-01

Gold foils were placed at different positions in the TRIGA reactor core and in the experimental devices. Absolute values of the thermal neutron flux at these positions were obtained by coincidence method. Preliminary fast neutron spectrum was measured by threshold detector and by 'Li 6 sandwich' detector. A short description of the applied method and obtained measurements results are included [sl

Self-powered neutron detector of high sensitivity

International Nuclear Information System (INIS)

Brixy, H.; Spillekothen, H.G.; Benninghofen, G.; Serafin, N.

1983-01-01

A self-powered neutron detector is proposed, consisting of three concentrically arranged electrically conducting tubes; where the central one forms the emitter and the inner and outer ones form the collector and where the tubes are electrically insulated from each other by insulating material. The emitter consists of a material with a high absorption cross-section for thermal neutrons, particularly of gadolinium, and is provided with an auxiliary emitter layer on the inside or the outside. With suitable dimensions and material, the auxiliary emitter layer increases the yield of electrons. (orig./HP) [de

Automatic neutron dosimetry system based on fluorescent nuclear track detector technology

International Nuclear Information System (INIS)

Akselrod, M.S.; Fomenko, V.V.; Bartz, J.A.; Haslett, T.L.

2014-01-01

For the first time, the authors are describing an automatic fluorescent nuclear track detector (FNTD) reader for neutron dosimetry. FNTD is a luminescent integrating type of detector made of aluminium oxide crystals that does not require electronics or batteries during irradiation. Non-destructive optical readout of the detector is performed using a confocal laser scanning fluorescence imaging with near-diffraction limited resolution. The fully automatic table-top reader allows one to load up to 216 detectors on a tray, read their engraved IDs using a CCD camera and optical character recognition, scan and process simultaneously two types of images in fluorescent and reflected laser light contrast to eliminate false-positive tracks related to surface and volume crystal imperfections. The FNTD dosimetry system allows one to measure neutron doses from 0.1 mSv to 20 Sv and covers neutron energies from thermal to 20 MeV. The reader is characterised by a robust, compact optical design, fast data processing electronics and user-friendly software. The first table-top automatic FNTD neutron dosimetry system was successfully tested for LLD, linearity and ability to measure neutrons in mixed neutron-photon fields satisfying US and ISO standards. This new neutron dosimetry system provides advantages over other technologies including environmental stability of the detector material, wide range of detectable neutron energies and doses, detector re-readability and re-usability and all-optical readout. A new adaptive image processing algorithm reliably removes false-positive tracks associated with surface and bulk crystal imperfections. (authors)

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)

Long-term performance of the CANDU-type of vanadium self-powered neutron detectors in NRU

Energy Technology Data Exchange (ETDEWEB)

Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)]. E-mail: leungt@aecl.ca

2007-07-01

The CANDU-type of in-core vanadium self-powered neutron flux detectors have been installed in NRU to monitor the axial neutron flux distributions adjacent to the loop fuel test sites since 1996. This paper describes how the thermal neutron fluxes were measured at two monitoring sites, and presents a method of correcting the vanadium burn-up effect, which can be up to 2 to 3% per year, depending on the detector locations in the reactor. It also presents the results of measurements from neutron flux detectors that have operated for over eight-years in NRU. There is good agreement between the measured and simulated neutron fluxes, to within {+-} 6.5%, and the long-term performance of the CANDU-type of vanadium neutron flux detectors in NRU is satisfactory. (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)

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.

Dosimetric evaluation of semiconductor detectors for application in neutron dosimetry and microdosimetry in nuclear reactor and radiosurgical facilities

International Nuclear Information System (INIS)

Cardenas, Jose Patricio Nahuel

2010-01-01

The main objective of this research is the dosimetric evaluation of semiconductor components (surface barrier detectors and PIN photodiodes) for applications in dose equivalent measurements on low dose fields (fast and thermal fluxes) using an AmBe neutron source, the IEA-R1 reactor neutrongraphy facility (epithermal and thermal fluxes) and the Critical Unit facility IPEN/MB-01 (fast fluxes). As moderator compound to fast neutrons flux from the AmBe source was used paraffin and boron and polyethylene as converter for thermal and fast neutrons measurements. The resulting fluxes were used to the irradiation of semiconductor components (SSB - Surface Barrier Detector and PIN photodiodes). A mixed converter made of a borated polyethylene foil (Kodak) was also used. Monte Carlo simulation methodology was employed to evaluate analytically the optimal paraffin thickness. The obtained results were similar to the experimental data and allowed the evaluation of emerging neutron flux from moderator, as well as the fast neutron flux reaching the polyethylene covering the semiconductor sensitive surface. Gamma radiation levels were evaluated covering the whole detector with cadmium foil 1 mm thick, allowing thermal neutrons blockage and gamma radiation measurements. The IPEN/MB-01 facility was employed to evaluate the detector response for high neutron flux. The results were in good agreement with other studies published. Using the obtained spectra an approach to dose equivalent calculation was established. (author)

A study of dissipative phenomena using Orion, a 4 π sectorized neutron detector

International Nuclear Information System (INIS)

Galin, J.; Guerreau, D.; Morjean, M.; Pouthas, J.; Saint-Laurent, F.; Sokolov, A.; Wang, X.M.; Piasecki, E.; Charvet, J.L.; CEA Centre d'Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette

1990-01-01

When studying the behavior of hot nuclei, the challenge is twofold: how are they formed in nucleus-nucleus collisions and how do they decay. For heavy and, thus neutron rich systems a large fraction of the thermalized energy is evacuated by neutron evaporation. Therefore the numbering, event-wise, of neutrons, over 4 π, gives a strong handle on energy dissipation for the different reaction channels. The first neutron measurements of this kind were performed using spherical detectors made of two hemispheres. Since then, a new and larger 4 π detector, ORION, has been designed in order to get information on the spatial distribution of the neutrons. The main characteristics of ORION are described and a few examples are given in order to illustrate the capabilities of such a detector in the study of dissipative collisions

Effect of high gamma background on neutron sensitivity of fission detectors

International Nuclear Information System (INIS)

Balagi, V.; Prasad, K.R.; Kataria, S.K.

2004-01-01

Tests were performed on two parallel plate and two cylindrical fission detectors in pulse and dc mode. The effect of gamma background on neutron sensitivity was studied in thermal neutron flux from 30 nv to 60 nv over which gamma field intensity ranging from 230 kR/h to 3.7 MR/h was superposed. In the case of one of the parallel plate detectors the fall in neutron sensitivity was observed to be 3.7% at 1 MR/h and negligible below 1 MR/h. In the case of one of the cylindrical counters the fall in neutron sensitivity was negligible below 500 kR/h and 37% at 1 MR/h. The data was used to derive the design parameters for a wide range fission detector to be procured for PFBR instrumentation for operation at 600 degC and gamma background of 1 MR/h. (author)

A Neutron Sensitive Microchannel Plate Detector with Cross Delay Line Readout

International Nuclear Information System (INIS)

Berry, Kevin D.; Bilheux, Hassina Z.; Crow, Lowell; Diawara, Yacouba; Feller, W. Bruce; Iverson, Erik B.; Martin, Adrian; Robertson, J. Lee

2012-01-01

Microchannel plates containing neutron absorbing elements such as boron and gadolinium in the bulk glass are used as the sensing element in high spatial resolution, high rate neutron imaging systems. In this paper we describe one such device, using both 10 B and natural Gd, which employs cross delay line signal readout, with time-of-flight capability. This detector has a measured spatial resolution under 40 m FWHM, thermal neutron efficiency of 19%, and has recorded rates in excess of 500 kHz. A physical and functional description is presented, followed by a discussion of measurements of detector performance and a brief survey of some practical applications.

Development of high efficiency neutron detectors

International Nuclear Information System (INIS)

Pickrell, M.M.; Menlove, H.O.

1993-01-01

The authors have designed a novel neutron detector system using conventional 3 He detector tubes and composites of polyethylene and graphite. At this time the design consists entirely of MCNP simulations of different detector configurations and materials. These detectors are applicable to low-level passive and active neutron assay systems such as the passive add-a-source and the 252 Cf shuffler. Monte Carlo simulations of these neutron detector designs achieved efficiencies of over 35% for assay chambers that can accommodate 55-gal. drums. Only slight increases in the number of detector tubes and helium pressure are required. The detectors also have reduced die-away times. Potential applications are coincident and multiplicity neutron counting for waste disposal and safeguards. The authors will present the general design philosophy, underlying physics, calculation mechanics, and results

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)

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.

The All Boron Carbide Diode Neutron Detector: Experiment and Modeling Approach

International Nuclear Information System (INIS)

Sabirianov, Ildar F.; Brand, Jennifer I.; Fairchild, Robert W.

2008-01-01

Boron carbide diode detectors, fabricated from two different polytypes of semiconducting boron carbide, will detect neutrons in reasonable agreement with theoretical expectations. The performance of the all boron carbide neutron detector differs, as expected, from devices where a boron rich neutron capture layer is distinct from the diode charge collection region (i.e. a conversion layer solid state detector). Diodes were fabricated from natural abundance boron (20% 10 B and 80% 11 B.) directly on the metal substrates and metal contacts applied to the films as grown. The total boron depth was on the order of 2 microns. This is clearly not a conversion-layer configuration. The diodes were exposed to thermal neutrons generated from a paraffin moderated plutonium-beryllium source in moderated and un-moderated, as well as shielded and unshielded experimental configurations, where the expected energy peaks at at 2.31 MeV and 2.8 MeV were clearly observed, albeit with some incomplete charge collection typical of thinner diode structures. The results are compared with other boron based thin film detectors and literature models. (authors)

SU-E-T-249: Neutron Model Upgrade for Radiotherapy Patients Monitoring Using a New Online Detector

Energy Technology Data Exchange (ETDEWEB)

Irazola, L; Sanchez Doblado, F. [Departamento de Fisiologia Medica y Biofisica, Universidad de Sevilla (Spain); Servicio de Radiofisica, Hospital Universitario Virgen Macarena, Sevilla (Spain); Lorenzoli, M; Pola, A. [Politecnico di Milano, Departimento di Ingegneria Nuclear, Milano (Italy); Terron, J.A. [Servicio de Radiofisica, Hospital Universitario Virgen Macarena, Sevilla (Spain); Bedogni, R. [Laboratori Nazionali di Frascati, Istituto Nazionale di Fisica Nucleare (INFN) (Italy); Sanchez Nieto, B. [Instituto de Fisica, Pontificia Universidad Catolica de Chile, Santiago (Chile); Romero-Exposito, M. [Departamento de Fisica, Universitat Autonoma de Barcelona (Spain)

2014-06-01

Purpose: The purpose of this work is to improve the existing methodology to estimate neutron equivalent dose in organs during radiotherapy treatments, based on a Static Random Access Memory neutron detector (SRAMnd) [1]. This is possible thanks to the introduction of a new digital detector with improved characteristics, which is able to measure online the neutron fluence rate in the presence of an intense photon background [2]. Its reduced size, allows the direct estimation of doses in specific points inside an anthropomorphic phantom (NORMA) without using passive detectors as TLD or CR-39. This versatility will allow not only to improve the existing models (generic abdomen and H and N [1]) but to generate more specific ones for any technique. Methods: The new Thermal Neutron Rate Detector (TNRD), based on a diode device sensitized to thermal neutrons, have been inserted in 16 points of the phantom. These points are distributed to infer doses to specific organs. Simultaneous measurements of these devices and a reference one, located in front of the gantry, have been performed for the mentioned generic treatments, in order to improve the existing model. Results: These new devices have shown more precise since they agree better with Monte Carlo simulations. The comparison of the thermal neutron fluence, measured with TNRD, and the existing models, converted from events to fluence, shows an average improvement of (3.90±3.37) % for H and N and (12.61±9.43) % for abdomen, normalized to the maximum value. Conclusion: This work indicates the potential of these new devices for more precise neutron equivalent dose estimation in organs, as a consequence of radiotherapy treatments. The simplicity of the process makes possible to establish more specific models that will provide a better dose estimation. References[1] Phys Med Biol 2012; 57:6167–6191.[2] A new active thermal neutron detector. Radiat. Prot. Dosim. (in press)

Connection factor calculation for isotopic neutron flux measurements with foil detectors

International Nuclear Information System (INIS)

Avila L, J.

1987-01-01

Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given

CdZnTe γ detector for deep inelastic neutron scattering on the VESUVIO spectrometer

Science.gov (United States)

Andreani, C.; D'Angelo, A.; Gorini, G.; Imberti, S.; Pietropaolo, A.; Rhodes, N. J.; Schooneveld, E. M.; Senesi, R.; Tardocchi, M.

In this paper it is shown that solid-state cadmium-zinc-telluride (CZT) is a promising photon detector for neutron spectroscopy in a wide energy interval, ranging from thermal ( 25 meV) to epithermal ( 70 eV) neutron energies. In the present study two CZT detectors were tested as part of the inverse-geometry neutron spectrometer VESUVIO operating at the ISIS pulsed neutron source. The response of the CZT detector to photon emission from radiative neutron capture in 238U was determined by biparametric measurements of neutron time of flight and photon energy. The scattering response function F(y) from a Pb sample has been derived using both CZT and conventional 6Li-glass scintillator detectors. The former showed both an improved signal to background ratio and higher efficiency as compared to 6Li glass, allowing us to measure F(y) up to the fourth 238U absorption energy (Er=66.02 eV). Due to the small size of CZT detectors, their use is envisaged in arrays, with high spatial resolution, for neutron-scattering studies at high energy (ω>1 eV) and low wavevector (q <10 Å-1) transfers.

CdZnTe γ detector for deep inelastic neutron scattering on the VESUVIO spectrometer

International Nuclear Information System (INIS)

Andreani, C.; Pietropaolo, A.; Senesi, R.; D'Angelo, A.; Gorini, G.; Imberti, S.; Tardocchi, M.; Rhodes, N.J.; Schooneveld, E.M.

2004-01-01

In this paper it is shown that solid-state cadmium-zinc-telluride (CZT) is a promising photon detector for neutron spectroscopy in a wide energy interval, ranging from thermal (∝25 meV) to epithermal (∝70 eV) neutron energies. In the present study two CZT detectors were tested as part of the inverse-geometry neutron spectrometer VESUVIO operating at the ISIS pulsed neutron source. The response of the CZT detector to photon emission from radiative neutron capture in 238 U was determined by biparametric measurements of neutron time of flight and photon energy. The scattering response function F(y) from a Pb sample has been derived using both CZT and conventional 6 Li-glass scintillator detectors. The former showed both an improved signal to background ratio and higher efficiency as compared to 6 Li glass, allowing us to measure F(y) up to the fourth 238 U absorption energy (E r =66.02 eV). Due to the small size of CZT detectors, their use is envisaged in arrays, with high spatial resolution, for neutron-scattering studies at high energy (ℎω>1 eV) and low wavevector (q -1 ) transfers. (orig.)

Dual detector neutron lifetime log: theory and practical applications

International Nuclear Information System (INIS)

Serpas, C.J.; Wichmann, P.A.; Fertl, W.H.; DeVries, M.R.; Rndall, R.R.

1977-01-01

The Neutron Lifetime Log instrumentation has continued to evolve and now is equipped with dual detectors for increased ease in gas detection and also a ratio response for a simultaneous porosity determination. A good deal of experimentation was involved to minimize both lithology and salinity effects on the porosity indication. This paper contains a discussion of the theory and concepts related to the application of the Dual Detector Neutron Lifetime Log (DNLL). It is important to note that with these advances the recording of thermal neutron capture cross section (Σ) remains consistent with the past measurements of earlier generations of instruments as the most accurate determination of this parameter. A number of field examples of the newly logged results are shown. These field cases include Dual Detector NLL's run thru the drill strings of highly deviated holes when difficulties were encountered in getting conventional open hole logs to bottom, logs thru open perforations and hot radioactive zones, comparisons of the large and small diameter instruments, logs with anomalous fluids in the annulus, logs thru multiple casing strings, and a number of other examples

A Micromegas Detector for Neutron Beam Imaging at the n_TOF Facility at CERN

CERN Document Server

Belloni, F; Berthoumieux, E; Calviani, M; Chiaveri, E; Colonna, N; Giomataris, Y; Guerrero, C; Gunsing, F; Iguaz, F J; Kebbiri, M; Pancin, J; Papaevangelou, T; Tsinganis, A; Vlachoudis, V; Altstadt, S; Andrzejewski, J; Audouin, L; Barbagallo, M; Bécares, V; Bečvář, F; Billowes, J; Boccone, V; Bosnar, D; Brugger, M; Calviño, F; Cano-Ott, D; Carrapiço, C; Cerutti, F; Chiaveri, E; Chin, M; Cortés, G; Corté-Giraldo, M A; Diakaki, M; Domingo-Pardo, C; Duran, I; Dzysiuk, N; Eleftheriadis, C; Ferrari, A; Fraval, K; Ganesan, S; García, A R; Giubrone, G; Gómez-Hornillos, M B; Gonçalves, I F; González-Romero, E; Griesmayer, E; Gurusamy, P; Jenkins, D G; Jericha, E; Kadi, Y; Käppeler, F; Karadimos, D; Koehler, P; Kokkoris, M; Krtička, M; Kroll, J; Langer, C; Lederer, C; Leeb, H; Leong, L S; Losito, R; Manousos, A; Marganiec, J; Marítnez, T; Massimi, C; Mastinu, P F; Mastromarco, M; Meaze, M; Mendoza, E; Mengoni, A; Milazzo, P M; Mingrone, F; Mirea, M; Mondalaers, W; Paradela, C; Pavlik, A; Perkowski, J; Plompen, A J M; Praena, J; Quesada, J M; Rauscher, T; Reifarth, R; Riego, A; Roman, F; Rubbia, C; Sarmento, R; Schillebeeckx, P; Schmidt, S; Tagliente, G; Tain, J L; Tarrío, D; Tassan-Got, L; Valenta, S; Vannini, G; Variale, V; Vaz, P; Ventura, A; Versaci, R; Vermeulen, M J; Vlastou, R; Wallner, A; Ware, T; Weigand, M; Weiss, C; Wright, T J; Žugec, P

2014-01-01

Micromegas (Micro-MEsh Gaseous Structure) detectors are gas detectors consisting of a stack of one ionization and one proportional chamber. A micromesh separates the two communicating regions, where two different electric fields establish respectively a charge drift and a charge multiplication regime. The n\\_TOF facility at CERN provides a white neutron beam (from thermal up to GeV neutrons) for neutron induced cross section measurements. These measurements need a perfect knowlodge of the incident neutron beam, in particular regarding its spatial profile. A position sensitive micromegas detector equipped with a B-10 based neutron/charged particle converter has been extensively used at the n\\_TOF facility for characterizing the neutron beam profile and extracting the beam interception factor for samples of different size. The boron converter allowed to scan the energy region of interest for neutron induced capture reactions as a function of the neutron energy, determined by the time of flight. Experimental ...

A neutron detector for measurement of total neutron production cross sections

International Nuclear Information System (INIS)

Sekharan, K.K.; Laumer, H.; Kern, B.D.; Gabbard, F.

1976-01-01

A neutron detector has been constructed and calibrated for the accurate measurement of total neutron production cross sections. The detector consists of a polyethylene sphere of 60 cm diameter in which eight 10 BF 3 counters have been installed radially. The relative efficiency of this detector has been determined for average neutron energies from 30 keV to 1.5 MeV by counting neutrons from 7 Li(p, n) 7 Be. By adjusting the radial positions of the BF 3 counters in the polyethylene sphere the efficiency for neutron detection was made nearly constant for this energy range. Measurement of absolute efficiency for the same neutron energy range has been done by counting the neutrons from 51 V(p, n) 51 Cr and 57 Fe(p, n) 57 Co reactions and determining the absolute number of residual nuclei produced during the measurement of neutron yield. Details of absolute efficiency measurements and the use of the detector for determination of neutron production cross sections are given. (Auth.)

Development of a large area thermal neutron detector based on a scintillator

International Nuclear Information System (INIS)

Engels, Ralf

2012-01-01

In the present work, the development and construction of a detector prototype based on wavelength shifting fiber in combination with a scintillator has been investigated and optimized. This development aims at an alternative for large area neutron detectors based on "3He detectors, which was the main construction in the past. After the study of the components and assemblies, such as: the scintillator, the wavelength-shifting-fibers and available photomultiplier tubes, the construction of the first prototype module begun. The neutron converter was selected as a "6LiF/ZnS scintillator, which produces a big light yield per absorbed neutron. The prototype itself is square and has an edge length of 30 cm in combination with two orthogonal layers of crossed wavelength-shifting-fibers. The top fiber layer, which is closer to the "6LiF/ZnS top scintillator produces the x-coordinates and the lower layer produces the y-coordinates for each event. In the prototype, MSJ-fibers from the company Kuraray were used with 1 mm diameter and spacing in the top layer of 1.5 mm and 1 mm in the lower layer. Due to the orthogonal arrangement of the wires in the two layers, one may identify where the neutron was absorbed in the scintillator and produced the light yield. In order to reduce the light loss of the absorbed photons inside the fibers, a bending radius of greater than 20 mm was used and achieved by warming up the fibers to 80 C during the bending process. The increased temperature reduces the crack formation in the fibers which increases the light loss. At this time it is expected that a photomultiplier from Hamamatsu with 256 individual pixels for readout will be used. This H9500 flat panel photomultiplier has the advantage of readout of all fibers of the prototype in one photomultiplier housing. In combination with integrated readout electronics one can minimize the homogeneity/gain differences of the photocathode pixels, the different light loss in each fiber, and the gain

Thermal neutron moderating device

International Nuclear Information System (INIS)

Takigami, Hiroyuki.

1995-01-01

In a thermal neutron moderating device, superconductive coils for generating magnetic fields capable of applying magnetic fields vertical to the longitudinal direction of a thermal neutron passing tube, and superconductive coils for magnetic field gradient for causing magnetic field gradient in the longitudinal direction of the thermal neutron passing tube are disposed being stacked at the outside of the thermal neutron passing tube. When magnetic field gradient is present vertically to the direction of a magnetic moment, thermal neutrons undergo forces in the direction of the magnetic field gradient in proportion to the magnetic moment. Then, the magnetic moment of the thermal neutrons is aligned with the direction vertical to the passing direction of the thermal neutrons, to cause the magnetic field gradient in the passing direction of the thermal neutrons. The speed of the thermal neutrons can be optionally selected and the wavelength can freely be changed by applying forces to the thermal neutrons and changing the extent and direction of the magnetic field gradient. Superconductive coils are used as the coils for generating magnetic fields and the magnetic field gradient in order to change extremely high energy of the thermal neutrons. (N.H.)

Neutron and X-ray Detectors

Energy Technology Data Exchange (ETDEWEB)

Carini, Gabriella [SLAC National Accelerator Lab., Menlo Park, CA (United States); Denes, Peter [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Gruener, Sol [Cornell Univ., Ithaca, NY (United States); Lessner, Elianne [Dept. of Energy (DOE), Washington DC (United States). Office of Science Office of Basic Energy Sciences

2012-08-01

The Basic Energy Sciences (BES) X-ray and neutron user facilities attract more than 12,000 researchers each year to perform cutting-edge science at these state-of-the-art sources. While impressive breakthroughs in X-ray and neutron sources give us the powerful illumination needed to peer into the nano- to mesoscale world, a stumbling block continues to be the distinct lag in detector development, which is slowing progress toward data collection and analysis. Urgently needed detector improvements would reveal chemical composition and bonding in 3-D and in real time, allow researchers to watch “movies” of essential life processes as they happen, and make much more efficient use of every X-ray and neutron produced by the source The immense scientific potential that will come from better detectors has triggered worldwide activity in this area. Europe in particular has made impressive strides, outpacing the United States on several fronts. Maintaining a vital U.S. leadership in this key research endeavor will require targeted investments in detector R&D and infrastructure. To clarify the gap between detector development and source advances, and to identify opportunities to maximize the scientific impact of BES user facilities, a workshop on Neutron and X-ray Detectors was held August 1-3, 2012, in Gaithersburg, Maryland. Participants from universities, national laboratories, and commercial organizations from the United States and around the globe participated in plenary sessions, breakout groups, and joint open-discussion summary sessions. Sources have become immensely more powerful and are now brighter (more particles focused onto the sample per second) and more precise (higher spatial, spectral, and temporal resolution). To fully utilize these source advances, detectors must become faster, more efficient, and more discriminating. In supporting the mission of today’s cutting-edge neutron and X-ray sources, the workshop identified six detector research challenges

The pin pixel detector--neutron imaging

CERN Document Server

Bateman, J E; Derbyshire, G E; Duxbury, D M; Marsh, A S; Rhodes, N J; Schooneveld, E M; Simmons, J E; Stephenson, R

2002-01-01

The development and testing of a neutron gas pixel detector intended for application in neutron diffraction studies is reported. Using standard electrical connector pins as point anodes, the detector is based on a commercial 100 pin connector block. A prototype detector of aperture 25.4 mmx25.4 mm has been fabricated, giving a pixel size of 2.54 mm which matches well to the spatial resolution typically required in a neutron diffractometer. A 2-Dimensional resistive divide readout system has been adapted to permit the imaging properties of the detector to be explored in advance of true pixel readout electronics. The timing properties of the device match well to the requirements of the ISIS-pulsed neutron source.

Research of boron conversion coating in neutron detector with boron deposited GEM

International Nuclear Information System (INIS)

Ye Di; Sun Zhijia; Zhou Jianrong; Wang Yanfeng; Yang Guian; Xu Hong; Chen Yuanbai; Xiao Yu; Diao Xungang

2014-01-01

GEM is a flourishing new gas detector and nowadays its technology become more mature. It has outstanding properties, such as excellent position resolution, high counting rate, radiation resistance, simple and flexible signal readout, can be large-area detector, wide application range. Detector with boron deposited GEM uses multilayer GEM with deposited boron film as neutron conversion carrier which reads out the information of neutron shot from the readout electrode with gas amplification from every GEM layer. The detector is high performance which can meet the demands of neutron detector of a new generation. Boron deposited neutron conversion electrode with boron deposited cathode and GEM included is the core part of the detector. As boron is a high-melting-point metalloid (> 2 000 ℃), electroplating and thermal evaporation are inappropriate ways. So finding a way to deposit boron on electrode which can meet the demands become a key technology in the development of neutron detector with boron deposited GEM. Compared with evaporation, sputtering has features such as low deposition temperature, high film purity, nice adhesive, thus is appropriate for our research. Magnetron sputtering is a improved way of sputtering which can get lower sputtering air pressure and higher target voltage, so that we can get better films. Through deposit process, the research uses magnetron sputtering to deposit pure boron film on copper electrode and GEM film. This method can get high quality, nice adhere, high purity, controllable uniformity, low cost film with high speed film formation. (authors)

Self powered neutron detectors

International Nuclear Information System (INIS)

Gopalan, C.S.; Ramachandra Rao, M.N.; Ingale, A.D.

1976-01-01

Two types of self powered neutron detectors used for in-core flux measurements are described. The characteristics of the various detectors, with emitters Rh, V, Co, Py are presented. Details about the fabrication of these detectors are given. (A.K.)

{sup 10}B multi-grid proportional gas counters for large area thermal neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Andersen, K. [ESS, P.O. Box 176, SE-221 00 Lund (Sweden); Bigault, T. [ILL, BP 156, 6, rue Jules Horowitz, 38042 Grenoble Cedex 9 (France); Birch, J. [Linköping University, SE-581, 83 Linköping (Sweden); Buffet, J. C.; Correa, J. [ILL, BP 156, 6, rue Jules Horowitz, 38042 Grenoble Cedex 9 (France); Hall-Wilton, R. [ESS, P.O. Box 176, SE-221 00 Lund (Sweden); Hultman, L. [Linköping University, SE-581, 83 Linköping (Sweden); Höglund, C. [ESS, P.O. Box 176, SE-221 00 Lund (Sweden); Linköping University, SE-581, 83 Linköping (Sweden); Guérard, B., E-mail: guerard@ill.fr [ILL, BP 156, 6, rue Jules Horowitz, 38042 Grenoble Cedex 9 (France); Jensen, J. [Linköping University, SE-581, 83 Linköping (Sweden); Khaplanov, A. [ILL, BP 156, 6, rue Jules Horowitz, 38042 Grenoble Cedex 9 (France); ESS, P.O. Box 176, SE-221 00 Lund (Sweden); Kirstein, O. [Linköping University, SE-581, 83 Linköping (Sweden); Piscitelli, F.; Van Esch, P. [ILL, BP 156, 6, rue Jules Horowitz, 38042 Grenoble Cedex 9 (France); Vettier, C. [ESS, P.O. Box 176, SE-221 00 Lund (Sweden)

2013-08-21

{sup 3}He was a popular material in neutrons detectors until its availability dropped drastically in 2008. The development of techniques based on alternative convertors is now of high priority for neutron research institutes. Thin films of {sup 10}B or {sup 10}B{sub 4}C have been used in gas proportional counters to detect neutrons, but until now, only for small or medium sensitive area. We present here the multi-grid design, introduced at the ILL and developed in collaboration with ESS for LAN (large area neutron) detectors. Typically thirty {sup 10}B{sub 4}C films of 1 μm thickness are used to convert neutrons into ionizing particles which are subsequently detected in a proportional gas counter. The principle and the fabrication of the multi-grid are described and some preliminary results obtained with a prototype of 200 cm×8 cm are reported; a detection efficiency of 48% has been measured at 2.5 Å with a monochromatic neutron beam line, showing the good potential of this new technique.

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)

Fission-product yields for thermal-neutron fission of 243Cm determined from measurements with a high-resolution low-energy germanium gamma-ray detector

International Nuclear Information System (INIS)

Merriman, L.D.

1984-04-01

Cumulative fission-product yields have been determined for 13 gamma rays emitted during the decay of 12 fission products created by thermal-neutron fission of 243 Cm. A high-resolution low-energy germanium detector was used to measure the pulse-height spectra of gamma rays emitted from a 77-nanogram sample of 243 Cm after the sample had been irradiated by thermal neutrons. Analysis of the data resulted in the identification and matching of gamma-ray energies and half-lives to individual radioisotopes. From these results, 12 cumulative fission product yields were deduced for radionuclides with half-lives between 4.2 min and 84.2 min. 7 references

A neutron detector for measurement of total neutron production cross sections

Energy Technology Data Exchange (ETDEWEB)

Sekharan, K K; Laumer, H; Kern, B D; Gabbard, F [Kentucky Univ., Lexington (USA). Dept. of Physics and Astronomy

1976-03-01

A neutron detector has been constructed and calibrated for the accurate measurement of total neutron production cross sections. The detector consists of a polyethylene sphere of 60 cm diameter in which eight /sup 10/BF/sub 3/ counters have been installed radially. The relative efficiency of this detector has been determined for average neutron energies from 30 keV to 1.5 MeV by counting neutrons from /sup 7/Li(p, n)/sup 7/Be. By adjusting the radial positions of the BF/sub 3/ counters in the polyethylene sphere the efficiency for neutron detection was made nearly constant for this energy range. Measurement of absolute efficiency for the same neutron energy range has been done by counting the neutrons from /sup 51/V(p, n)/sup 51/Cr and /sup 57/Fe(p, n)/sup 57/Co reactions and determining the absolute number of residual nuclei produced during the measurement of neutron yield. Details of absolute efficiency measurements and the use of the detector for determination of neutron production cross sections are given.

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)

A large solid angle multiparameter neutron detector

International Nuclear Information System (INIS)

Ricco, G.; Anghinolfi, M.; Corvisiero, P.; Durante, E.; Maggiolo, S.; Prati, P.; Rottura, A.; Taiuti, M.

1991-01-01

A 4π neutron detector has been realized using organic scintillators: the detector is suitable for high efficiency, low background measurements of very low neutron rates in the 0.6-5 MeV energy range. Gamma-neutron discrimination has been performed by pulse shape, energy and neutron lifetime analysis and backgrounds have been reduced by anticoincidence detectors and paraffin-lead shielding. Tests of efficiency, energy resolution and radiation identification have been made with a low intensity Am-Be neutron source. (orig.)

Neutron and gamma sensitivities of self-powered detectors: Monte Carlo modelling

Energy Technology Data Exchange (ETDEWEB)

Vermeeren, Ludo [SCK-CEN, Nuclear Research Centre, Boeretang 200, B-2400 Mol, (Belgium)

2015-07-01

This paper deals with the development of a detailed Monte Carlo approach for the calculation of the absolute neutron sensitivity of SPNDs, which makes use of the MCNP code. We will explain the calculation approach, including the activation and beta emission steps, the gamma-electron interactions, the charge deposition in the various detector parts and the effect of the space charge field in the insulator. The model can also be applied for the calculation of the gamma sensitivity of self-powered detectors and for the radiation-induced currents in signal cables. The model yields detailed information on the various contributions to the sensor currents, with distinct response times. Results for the neutron sensitivity of various types of SPNDs are in excellent agreement with experimental data obtained at the BR2 research reactor. For typical neutron to gamma flux ratios, the calculated gamma induced SPND currents are significantly lower than the neutron induced currents. The gamma sensitivity depends very strongly upon the immediate detector surroundings and on the gamma spectrum. Our calculation method opens the way to a reliable on-line determination of the absolute in-pile thermal neutron flux. (authors)

CdZnTe {gamma} detector for deep inelastic neutron scattering on the VESUVIO spectrometer

Energy Technology Data Exchange (ETDEWEB)

Andreani, C.; Pietropaolo, A.; Senesi, R. [Dipartimento di Fisica, Universita degli Studi di Roma ' Tor Vergata' , Via della Ricerca Scientifica 1, 00133, Roma (Italy); Istituto Nazionale per la Fisica della Materia, UdR, Tor Vergata (Italy); D' Angelo, A. [Dipartimento di Fisica, Universita degli Studi di Roma ' Tor Vergata' , Via della Ricerca Scientifica 1, 00133, Roma (Italy); Istituto Nazionale di Fisica Nucleare, Sezione, Roma II (Italy); Gorini, G.; Imberti, S.; Tardocchi, M. [Dipartimento di Fisica G. Occhialini, Universita degli Studi di Milano-Bicocca, Piazza della Scienza 3, 20126, Milano (Italy); Istituto Nazionale per la Fisica della Materia, UdR, Milano-Bicocca (Italy); Rhodes, N.J.; Schooneveld, E.M. [Isis Facility, Rutherford Appleton Laboratory, Chilton, Didcot, OX11 0QX, Oxfordshire (United Kingdom)

2004-03-01

In this paper it is shown that solid-state cadmium-zinc-telluride (CZT) is a promising photon detector for neutron spectroscopy in a wide energy interval, ranging from thermal ({proportional_to}25 meV) to epithermal ({proportional_to}70 eV) neutron energies. In the present study two CZT detectors were tested as part of the inverse-geometry neutron spectrometer VESUVIO operating at the ISIS pulsed neutron source. The response of the CZT detector to photon emission from radiative neutron capture in {sup 238}U was determined by biparametric measurements of neutron time of flight and photon energy. The scattering response function F(y) from a Pb sample has been derived using both CZT and conventional {sup 6}Li-glass scintillator detectors. The former showed both an improved signal to background ratio and higher efficiency as compared to {sup 6}Li glass, allowing us to measure F(y) up to the fourth {sup 238}U absorption energy (E{sub r}=66.02 eV). Due to the small size of CZT detectors, their use is envisaged in arrays, with high spatial resolution, for neutron-scattering studies at high energy ({Dirac_h}{omega}>1 eV) and low wavevector (q <10 A{sup -1}) transfers. (orig.)

Characterization of plastic nuclear track detectors on solid state, CR-39 and LR-115 and its possibilities application on thermal and fast neutron dosimetry

International Nuclear Information System (INIS)

Vallejo Delgado, L.R.

1989-01-01

This work is an study about the use feasibility of plastic nuclear track detectors, LR 115, II-B (of Eastmann Kodak Co) and CR-39 (of American Acrylics and Plastics), for thermal and fast neutron dosimetry, respectively. The LR-115 with converter (n, alpha) was exposed to thermal neutrons with energy of 0,046 e V, proceeding from nuclear reactor RECH-1 of Nuclear Energy Chilean Commission. The irradiated films were submited to a chemical etching with NaOH, plus a washing and brushing. The CR-39 with polyethylene irradiator, was exposed to fast neutrons proceeding of calibrated sources of Am-Se. The irradiated plates were submited to a chemical pre-etching with KOH and a electrochemical post-etching. (author)

Digital signal processing for a thermal neutron detector using ZnS(Ag):{sup 6}LiF scintillating layers read out with WLS fibers and SiPMs

Energy Technology Data Exchange (ETDEWEB)

Mosset, J.-B., E-mail: jean-baptiste.mosset@psi.ch; Stoykov, A.; Greuter, U.; Hildebrandt, M.; Schlumpf, N.

2016-07-11

We present a digital signal processing system based on a photon counting approach which we developed for a thermal neutron detector consisting of ZnS(Ag):{sup 6}LiF scintillating layers read out with WLS fibers and SiPMs. Three digital filters have been evaluated: a moving sum, a moving sum after differentiation and a digital CR-RC{sup 4} filter. The performances of the detector with these filters are presented. A full analog signal processing using a CR-RC{sup 4} filter has been emulated digitally. The detector performance obtained with this analog approach is compared with the one obtained with the best performing digital approach. - Highlights: • Application of digital signal processing for a SiPM-based ZnS:6LiF neutron detector. • Optimisation of detector performances with 3 different digital filters. • Comparison with detector performances with a full analog signal processing.

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)

Estimate of the neutron fields in ATLAS based on ATLAS-MPX detectors data

International Nuclear Information System (INIS)

Bouchami, J; Dallaire, F; Gutierrez, A; Idarraga, J; Leroy, C; Picard, S; Scallon, O; Kral, V; PospIsil, S; Solc, J; Suk, M; Turecek, D; Vykydal, Z; Zemlieka, J

2011-01-01

The ATLAS-MPX detectors are based on Medipix2 silicon devices designed by CERN for the detection of different types of radiation. These detectors are covered with converting layers of 6 LiF and polyethylene (PE) to increase their sensitivity to thermal and fast neutrons, respectively. These devices allow the measurement of the composition and spectroscopic characteristics of the radiation field in ATLAS, particularly of neutrons. These detectors can operate in low or high preset energy threshold mode. The signature of particles interacting in a ATLAS-MPX detector at low threshold are clusters of adjacent pixels with different size and form depending on their type, energy and incidence angle. The classification of particles into different categories can be done using the geometrical parameters of these clusters. The Medipix analysis framework (MAFalda) - based on the ROOT application - allows the recognition of particle tracks left in ATLAS-MPX devices located at various positions in the ATLAS detector and cavern. The pattern recognition obtained from the application of MAFalda was configured to distinguish the response of neutrons from other radiation. The neutron response at low threshold is characterized by clusters of adjoining pixels (heavy tracks and heavy blobs) left by protons and heavy ions resulting from neutron interactions in the converting layers of the ATLAS-MPX devices. The neutron detection efficiency of ATLAS-MPX devices has been determined by the exposure of two detectors of reference to radionuclide sources of neutrons ( 252 Cf and 241 AmBe). With these results, an estimate of the neutrons fields produced at the devices locations during ATLAS operation was done.

Geant4 Analysis of a Thermal Neutron Real-Time Imaging System

Science.gov (United States)

Datta, Arka; Hawari, Ayman I.

2017-07-01

Thermal neutron imaging is a technique for nondestructive testing providing complementary information to X-ray imaging for a wide range of applications in science and engineering. Advancement of electronic imaging systems makes it possible to obtain neutron radiographs in real time. This method requires a scintillator to convert neutrons to optical photons and a charge-coupled device (CCD) camera to detect those photons. Alongside, a well collimated beam which reduces geometrical blurriness, the use of a thin scintillator can improve the spatial resolution significantly. A representative scintillator that has been applied widely for thermal neutron imaging is 6LiF:ZnS (Ag). In this paper, a multiphysics simulation approach for designing thermal neutron imaging system is investigated. The Geant4 code is used to investigate the performance of a thermal neutron imaging system starting with a neutron source and including the production of charged particles and optical photons in the scintillator and their transport for image formation in the detector. The simulation geometry includes the neutron beam collimator and sapphire filter. The 6LiF:ZnS (Ag) scintillator is modeled along with a pixelated detector for image recording. The spatial resolution of the system was obtained as the thickness of the scintillator screen was varied between 50 and 400 μm. The results of the simulation were compared to experimental results, including measurements performed using the PULSTAR nuclear reactor imaging beam, showing good agreement. Using the established model, further examination showed that the resolution contribution of the scintillator screen is correlated with its thickness and the range of the neutron absorption reaction products (i.e., the alpha and triton particles). Consequently, thinner screens exhibit improved spatial resolution. However, this will compromise detection efficiency due to the reduced probability of neutron absorption.

MCNPX Simulation Study of STRAW Neutron Detectors - Summary Paper

International Nuclear Information System (INIS)

Mukhopadhyay, Sanjoy; Maurer, Richard; Mitchell, Stephen

2010-01-01

A novel prototype fission meter is being designed at National Security Technologies, LLC, using a thin uniform coating (only 1 micron thick) of 10 B as a neutron converter inside a large array of thin (4 mm diameter) copper tubes. The copper tubes are only 2 mils thick, and each holds the stretched anode wire under tension and high voltage. The tubes are filled with proportional counter gas (a mixture of 90%/10% of Ar/CO 2 ). The tubes operate in proportional counter mode and attract mobile charged particles (α's) created in the nuclear interaction 10 B(n, 4 He) 7 Li. However, a single tube has about 1/7th the sensitivity of a 3 He tube. Modeling is required to determine if enough such tubes could be placed in a neutron detection assembly of the current size to give comparable sensitivity to 3 He. Detectors lined with 10 B lie between 3 He and 10 BF 3 proportional counters and fission chambers in terms of neutron detection efficiency and gamma ray insensitivity. The mean free path of thermal neutrons in 10 B is about 18 (micro)m. It takes about 60 (micro)m of 10 B layer to completely stop thermal neutrons, but the energetic α-particles generated in the reaction have a range of only 3.3 (micro)m in 10 B environment - hence the thin layer of boron coating on the copper tube. The prototype design is shown in Figure 1. It consists of two panels of three staggered rows of 500-mm-long, 4-mm-diameter straws, with 20 in each row, embedded in 30-mm-thick high density polyethylene (HDPE). The project demonstrates a new application of thin neutron and gamma converter technique (1 micron thin 10 B coated copper tube). It exploits fast timing from multiple straw detectors to count multiplicity of both gamma and neutrons from fissioning materials. The objective is to find a near-term replacement of 3 He gas in neutron detection and measurement (with a very large neutron detection area). All the solid-state detectors developed thus far are small and inefficient. The thin size

Neutron detector assembly

International Nuclear Information System (INIS)

Hanai, Koi; Shirayama, Shinpei.

1978-01-01

Purpose: To prevent gamma-ray from leaking externally passing through the inside of a neutron detector assembly. Constitution: In a neutron detector assembly having a protection pipe formed with an enlarged diameter portion which serves also as a spacer, partition plates with predetermined width are disposed at the upper and the lower portions in this expanded portion. A lot of metal particles are filled into spaces formed by the partition plates. In such a structure, the metal particles well-absorb the gamma-rays from above and convert them into heat to provide shielding for the gamma-rays. (Horiuchi, T.)

Implantable self-powered detector for on-line determination of neutron flux in patients during NCT treatment.

Science.gov (United States)

Miller, M E; Mariani, L E; Gonçalves-Carralves, M L Sztejnberg; Skumanic, M; Thorp, S I

2004-11-01

A novel system to determine thermal neutron flux in real time during NCT treatments was developed in the National Atomic Energy Commission of Argentina. The system is based on a special self-powered detector that can be implanted in patients owing to its small size and biocompatibility. High voltage is not required to operate this kind of detectors, which is a considerable advantage in terms of medical uses. By choosing the appropriate materials, it was possible to obtain a prototype with thermal neutron sensitivity providing for an adequate signal level in typical NCT thermal fluxes. It was also possible to minimize gamma response in order to neglect its contribution.

Implantable self-powered detector for on-line determination of neutron flux in patients during NCT treatment

Energy Technology Data Exchange (ETDEWEB)

Miller, M.E. E-mail: miller@cae.cnea.gov.ar; Mariani, L.E.; Sztejnberg Goncalves-Carralves, M.L.; Skumanic, M.; Thorp, S.I

2004-11-01

A novel system to determine thermal neutron flux in real time during NCT treatments was developed in the National Atomic Energy Commission of Argentina. The system is based on a special self-powered detector that can be implanted in patients owing to its small size and biocompatibility. High voltage is not required to operate this kind of detectors, which is a considerable advantage in terms of medical uses. By choosing the appropriate materials, it was possible to obtain a prototype with thermal neutron sensitivity providing for an adequate signal level in typical NCT thermal fluxes. It was also possible to minimize gamma response in order to neglect its contribution.

Detection of boron in metal alloys with solid state nuclear track detector by neutron induced autoradiography

International Nuclear Information System (INIS)

Ali Nabipour; Hosseini, A.; Afarideh, H.

2002-01-01

Neutron induced autoradiography is very useful technique for detection as well as measurement of Boron densities in metal alloys. The method is relatively simple and quite sensitive in comparison with other techniques with resolution in the range of PPM. Using this technique with it is also possible to investigate microscopic scattering of Boron in metal alloys. In comparison with most techniques neutron induced autoradiography has its own difficulties and limitations. In this research measurement of Boron densities and investigation of that diffusion in metal alloys has been carried out. A flat nicely polished Boron doped metal samples is covered with a track detecting plastic (CR-39 solid state nuclear track detector) and exposed to thermal neutron dose. After irradiation the plastic detector have been removed and put in an etching solution. Since the diffusion rate of corrosive solution in those area, which heavy ions have been, produces as the result of nuclear reaction with thermal neutron are more than the other areas, some cavities are formed. The diameter of cavities or tracks cross section are increased with increasing the etching time, to some extent that it is possible to observe the cavities with optical microscopes. The density of tracks on the detector surface is directly related to the Boron concentration in the sample and thermal neutron dose. So by measuring the number of tracks on surface of the detector it would possible to calculate the concentration of Boron in metal samples. (Author)

Thermal neutron imaging in an active interrogation environment

International Nuclear Information System (INIS)

Vanier, P.E.; Forman, L.; Norman, D.R.

2009-01-01

We have developed a thermal-neutron coded-aperture imager that reveals the locations of hydrogenous materials from which thermal neutrons are being emitted. This imaging detector can be combined with an accelerator to form an active interrogation system in which fast neutrons are produced in a heavy metal target by means of excitation by high energy photons. The photo-induced neutrons can be either prompt or delayed, depending on whether neutronemitting fission products are generated. Provided that there are hydrogenous materials close to the target, some of the photo-induced neutrons slow down and emerge from the surface at thermal energies. These neutrons can be used to create images that show the location and shape of the thermalizing materials. Analysis of the temporal response of the neutron flux provides information about delayed neutrons from induced fission if there are fissionable materials in the target. The combination of imaging and time-of-flight discrimination helps to improve the signal-to-background ratio. It is also possible to interrogate the target with neutrons, for example using a D-T generator. In this case, an image can be obtained from hydrogenous material in a target without the presence of heavy metal. In addition, if fissionable material is present in the target, probing with fast neutrons can stimulate delayed neutrons from fission, and the imager can detect and locate the object of interest, using appropriate time gating. Operation of this sensitive detection equipment in the vicinity of an accelerator presents a number of challenges, because the accelerator emits electromagnetic interference as well as stray ionizing radiation, which can mask the signals of interest.

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)

Investigation of gamma-ray sensitivity of neutron detectors based on thin converter films

Energy Technology Data Exchange (ETDEWEB)

Khaplanov, A; Hall-Wilton, R [European Spallation Source, P.O Box 176, SE-22100 Lund (Sweden); Piscitelli, F; Buffet, J-C; Clergeau, J-F; Correa, J; Esch, P van; Ferraton, M; Guerard, B [Institute Laue Langevin, Rue Jules Horowitz, FR-38042 Grenoble (France)

2013-10-15

Currently, many detector technologies for thermal neutron detection are in development in order to lower the demand for the rare {sup 3}He gas. Gas detectors with solid thin film neutron converters readout by gas proportional counter method have been proposed as an appropriate choice for applications where large area coverage is necessary. In this paper, we investigate the probability for {gamma}-rays to generate a false count in a neutron measurement. Simulated results are compared to measurement with {sup 10}B thin film prototypes and a {sup 3}He detector. It is demonstrated that equal {gamma}-ray rejection to that of {sup 3}He tubes is achieved with the new technology. The arguments and results presented here are also applicable to gas detectors with converters other than solid {sup 10}B layers, such as {sup 6}Li layers and {sup 10}BF{sub 3} gas.

Methods and apparatus for environmental correction of thermal neutron logs

International Nuclear Information System (INIS)

Preeg, W.E.; Scott, H.D.

1983-01-01

An on-line environmentally-corrected measurement of the thermal neutron decay time (tau) of an earth formation traversed by a borehole is provided in a two-detector, pulsed neutron logging tool, by measuring tau at each detector and combining the two tau measurements in accordance with a previously established empirical relationship of the general form: tau = tausub(F) +A(tausub(F) + tausub(N)B) + C, where tausub(F) and tausub(N) are the tau measurements at the far-spaced and near-spaced detectors, respectively, A is a correction coefficient for borehole capture cross section effects, B is a correction coefficient for neutron diffusion effects, and C is a constant related to parameters of the logging tool. Preferred numerical values of A, B and C are disclosed, and a relationship for more accurately approximating the A term to specific borehole conditions. (author)

Radiation hard detectors from silicon enriched with both oxygen and thermal donors improvements in donor removal and long-term stability with regard to neutron irradiation

CERN Document Server

Li, Z; Eremin, V; Dezillie, B; Chen, W; Bruzzi, M

2002-01-01

Detectors made on the silicon wafers with high concentration of thermal donors (TD), which were introduced during the high temperature long time (HTLT) oxygenation procedure, have been investigated in the study of radiation hardness with regard to neutron irradiation and donor removal problems in irradiated high resistivity Si detectors. Two facts have been established as the evidence of radiation hardness improvement of HTLT(TD) Si detectors irradiated below approx 10 sup 1 sup 4 n/cm sup 2 compared to detectors made on standard silicon wafers: the increase of space charge sign inversion fluence (of 1 MeV neutrons) due to lower initial Si resistivity dominated by TD, and the gain in the reverse annealing time constant tau favourable for this material. Coupled with extremely high radiation tolerance to protons observed earlier ('beta zero' behaviour in a wide range of fluence), detectors from HTLT(TD) Si may be unique for application in the experiments with multiple radiations. The changes in the effective sp...

Indigenous development of diamond detectors for monitoring neutrons

International Nuclear Information System (INIS)

Singh, Arvind; Amit Kumar; Topkar, Anita; Pithawa, C.K.

2013-01-01

High purity synthetic chemically vapor deposited (CVD) diamond has several outstanding characteristics that make it as an important material for detector applications specifically for extreme environmental conditions like high temperature, high radiation, and highly corrosive environments. Diamond detectors are especially considered promising for monitoring fast neutrons produced by the D-T nuclear fusion reactions in next generation fusion facilities such as ITER. When fast neutrons interact with carbon, elastic, inelastic and (n,α) type reactions can occur. These reactions can be employed for the detection of fast neutrons using diamond. We have initiated the development of diamond detectors based on synthetic CVD substrates. In this paper, the first test of a polycrystalline CVD diamond detector with fast neutrons is reported. The test results demonstrate that this detector can be used for monitoring fast neutrons. The diamond detectors have been fabricated using 5 mm x 5 mm, 300 μm polycrystalline diamond substrates. Aluminum metallization has been used on both sides of the detector to provide electrical contacts. The performance of fabricated detectors was first evaluated using current and capacitance measurements. The leakage current was observed to be stable and about a few pAs for voltages up to 300V. The capacitance-voltage characteristics showed a constant capacitance which is as expected. To confirm the response of the detector to charged particles, the pulse height spectrum (PHS) was obtained using 238 Pu- 239 Pu dual α- source. The PHS showed a continuum without any peak due to polycrystalline nature of diamond film. The response of the detector to fast neutrons has been studied using the fast neutron facility at NXF, BARC. The PHS obtained for a neutron yield of 4 x 10 8 n/s is shown. The average counts per second (cps) measured for diamond detector for different neutron yields is shown. The plot shows linearity with coefficient of determination R

The 4π neutron detector CARMEN

Energy Technology Data Exchange (ETDEWEB)

Ledoux, X., E-mail: Xavier.ledoux@ganil.fr [CEA/DAM/DIF, F-91297 Arpajon (France); GANIL, CEA/DRF-CNRS/IN2P3, Caen, F-14076 France (France); Laborie, J.-M.; Pras, P.; Lantuéjoul-Thfoin, I.; Varignon, C. [CEA/DAM/DIF, F-91297 Arpajon (France)

2017-02-01

CARMEN is a 4π neutron detector filled with a gadolinium-loaded liquid scintillator built to measure neutron multiplicity distributions. It is used to study fission and (n,xn) reactions. In addition to neutron multiplicity measurements, CARMEN can be used to measure neutron energy spectra with the time-of-flight technique, thanks to the time properties of the prompt signal. The detector, detection technique and efficiency determination are presented in detail. Two examples are also presented: the measurement of {sup 252}Cf spontaneous fission neutron multiplicity probability distribution and the measurement of the neutron energy spectrum emitted by an Am-Be radioactive source.

Activation neutron detector

International Nuclear Information System (INIS)

Ambardanishvili, T.S.; Kolomiitsev, M.A.; Zakharina, T.Y.; Dundua, V.J.; Chikhladze, N.V.

1976-01-01

An activation neutron detector made as a moulded and cured composition of a material capable of being neutron-activated is described. The material is selected from a group consisting of at least two chemical elements, a compound of at least two chemical elements and their mixture, each of the chemical elements and their mixture, each of the chemical elements being capable of interacting with neutrons to form radioactive isotopes having different radiation energies when disintegrating. The material capable of being neutron-activated is distributed throughout the volume of a polycondensation resin inert with respect to neutrons and capable of curing. 17 Claims, No Drawings

Estimate of the neutron fields in ATLAS based on ATLAS-MPX detectors data

Energy Technology Data Exchange (ETDEWEB)

Bouchami, J; Dallaire, F; Gutierrez, A; Idarraga, J; Leroy, C; Picard, S; Scallon, O [Universite de Montreal, Montreal, Quebec H3C 3J7 (Canada); Kral, V; PospIsil, S; Solc, J; Suk, M; Turecek, D; Vykydal, Z; Zemlieka, J, E-mail: scallon@lps.umontreal.ca [Institute of Experimental and Applied Physics of the CTU in Prague, Horska 3a/22, CZ-12800 Praha2 - Albertov (Czech Republic)

2011-01-15

The ATLAS-MPX detectors are based on Medipix2 silicon devices designed by CERN for the detection of different types of radiation. These detectors are covered with converting layers of {sup 6}LiF and polyethylene (PE) to increase their sensitivity to thermal and fast neutrons, respectively. These devices allow the measurement of the composition and spectroscopic characteristics of the radiation field in ATLAS, particularly of neutrons. These detectors can operate in low or high preset energy threshold mode. The signature of particles interacting in a ATLAS-MPX detector at low threshold are clusters of adjacent pixels with different size and form depending on their type, energy and incidence angle. The classification of particles into different categories can be done using the geometrical parameters of these clusters. The Medipix analysis framework (MAFalda) - based on the ROOT application - allows the recognition of particle tracks left in ATLAS-MPX devices located at various positions in the ATLAS detector and cavern. The pattern recognition obtained from the application of MAFalda was configured to distinguish the response of neutrons from other radiation. The neutron response at low threshold is characterized by clusters of adjoining pixels (heavy tracks and heavy blobs) left by protons and heavy ions resulting from neutron interactions in the converting layers of the ATLAS-MPX devices. The neutron detection efficiency of ATLAS-MPX devices has been determined by the exposure of two detectors of reference to radionuclide sources of neutrons ({sup 252}Cf and {sup 241}AmBe). With these results, an estimate of the neutrons fields produced at the devices locations during ATLAS operation was done.

Estimate of the neutron fields in ATLAS based on ATLAS-MPX detectors data

Science.gov (United States)

Bouchami, J.; Dallaire, F.; Gutiérrez, A.; Idarraga, J.; Král, V.; Leroy, C.; Picard, S.; Pospíšil, S.; Scallon, O.; Solc, J.; Suk, M.; Turecek, D.; Vykydal, Z.; Žemlièka, J.

2011-01-01

The ATLAS-MPX detectors are based on Medipix2 silicon devices designed by CERN for the detection of different types of radiation. These detectors are covered with converting layers of 6LiF and polyethylene (PE) to increase their sensitivity to thermal and fast neutrons, respectively. These devices allow the measurement of the composition and spectroscopic characteristics of the radiation field in ATLAS, particularly of neutrons. These detectors can operate in low or high preset energy threshold mode. The signature of particles interacting in a ATLAS-MPX detector at low threshold are clusters of adjacent pixels with different size and form depending on their type, energy and incidence angle. The classification of particles into different categories can be done using the geometrical parameters of these clusters. The Medipix analysis framework (MAFalda) — based on the ROOT application — allows the recognition of particle tracks left in ATLAS-MPX devices located at various positions in the ATLAS detector and cavern. The pattern recognition obtained from the application of MAFalda was configured to distinguish the response of neutrons from other radiation. The neutron response at low threshold is characterized by clusters of adjoining pixels (heavy tracks and heavy blobs) left by protons and heavy ions resulting from neutron interactions in the converting layers of the ATLAS-MPX devices. The neutron detection efficiency of ATLAS-MPX devices has been determined by the exposure of two detectors of reference to radionuclide sources of neutrons (252Cf and 241AmBe). With these results, an estimate of the neutrons fields produced at the devices locations during ATLAS operation was done.

Large-area self-powered neutron-detectors for neutron-flux measurements in HTRs. Status of developmental work

International Nuclear Information System (INIS)

Brixy, H.; Hecker, R.; Serpekian, T.; Benninghofen, G.; Serafin, N.; Spillekothen, H.G.

1982-06-01

The development is described of the large-area SPN-detector as an out of core power monitoring system. Gadolinium or cobalt was used as the emitter. Response functions of the gadolinium SPN-detector were found with regard to the reactor power, the effect of the gamma field, its short-term behaviour following reactor shutdown and long-term behaviour during reactor operation. It was shown that a detector of 0.1 mm emitter thickness can withstand an integral thermal neutron flux of 2.10 20 nvt almost without efficiency loss thus indicating that the large-area gadolinium SPN-detector is a suitable means for power monitoring in large HTGR's

The EUROBALL neutron wall - design and performance tests of neutron detectors

CERN Document Server

Skeppstedt, Ö; Lindström, L; Wadsworth, R; Hibbert, I; Kelsall, N; Jenkins, D; Grawe, H; aGórska, M; Moszynski, M; Sujkowski, Z; Wolski, D; Kapusta, M; Hellström, M; Kalogeropoulos, S; Oner, D; Johnson, A; Cederkäll, J; Klamra, W; Nyberg, J; Weiszflog, M; Kay, J; Griffiths, R; Garces-Narro, J; Pearson, C; Eberth, J

1999-01-01

The mechanical design of the EUROBALL neutron wall and neutron detectors, and their performance measured with a sup 2 sup 4 sup 6 sup , sup 2 sup 4 sup 8 Cm fission source are described. The array consists of 15 pseudohexaconical detector units subdivided into three, 149 mm high, hermetically separated segments and a smaller central pentagonal unit subdivided into five segments. The detectors are filled with Bicron BC501A liquid scintillator. Each section of the hexaconical detectors is viewed by a 130 mm diameter Philips XP4512PA photomultiplier while the sections of pentagonal detectors are viewed by Philips XP4312B PMTs. The tests of n-gamma discrimination performed by zero-crossing and time-of-flight methods show a full separation of gamma- and neutron events down to 50 keV recoil electron energy. These tests demonstrate the excellent timing properties of the detectors and an average time resolution of 1.56 ns. The factors determining the efficiency of neutron detectors are discussed. The total efficiency...

Optimization of a neutron detector system for the fourier RTOF-diffractometer facility at the ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Maayouf, R M.A.; El-Kady, A S.I.; El-Shaer, Y H [Reactor and Neutron physics Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

1997-12-31

This work deals with optimization of a detector system for use with the fourier reverse time of flight (RTOF) diffractometer facility recently installed at the ETRR-1 reactor. The detector system of the diffractometer was designed for detecting neutrons scattered from the sample. It consists of 4 independent scintillation detector elements arranged according to the time focusing geometry in order to increase the luminosity of the diffractometer for the given resolution. Each of the detector elements is made of 1 mm thick {sup 6} Li-glass scintillator (NE-912) whose surface area is 200 mm{sup 2}. The present detector system ensures reliable identification and separation of thermal neutron pulses from fast neutrons and gamma- quanta. It was found from measurements with Fe sample that it is possible to make neutron diffraction measurement, within one hour using the present detector system and with resultion 0.5%. 6 FIGS.

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)

Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

Energy Technology Data Exchange (ETDEWEB)

LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

2014-07-01

Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

Fusion neutron detector calibration using a table-top laser generated plasma neutron source

International Nuclear Information System (INIS)

Hartke, R.; Symes, D.R.; Buersgens, F.; Ruggles, L.E.; Porter, J.L.; Ditmire, T.

2005-01-01

Using a high intensity, femtosecond laser driven neutron source, a high-sensitivity neutron detector was calibrated. This detector is designed for observing fusion neutrons at the Z accelerator in Sandia National Laboratories. Nuclear fusion from laser driven deuterium cluster explosions was used to generate a clean source of nearly monoenergetic 2.45 MeV neutrons at a well-defined time. This source can run at 10 Hz and was used to build up a clean pulse-height spectrum on scintillating neutron detectors giving a very accurate calibration for neutron yields at 2.45 MeV

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

Ultra Low Level Environmental Neutron Measurements Using Superheated Droplet Detectors

Energy Technology Data Exchange (ETDEWEB)

Fernandes, A.C. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Estrada Nacional 10 - km 139.7, 2695-066 Bobadela LRS (Portugal); Centro de Fisica Nuclear, Universidade de Lisboa. Av. Prof. Gama Pinto, 2, 1649- 003 Lisboa (Portugal); Felizardo, M.; Girard, T.A.; Kling, A.; Ramos, A.R. [Centro de Fisica Nuclear, Universidade de Lisboa. Av. Prof. Gama Pinto, 2, 1649- 003 Lisboa (Portugal); Marques, J.G.; Prudencio, M.I.; Marques, R.; Carvalho, F.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Estrada Nacional 10 - km 139.7, 2695-066 Bobadela LRS (Portugal)

2015-07-01

Through the application of superheated droplet detectors (SDDs), the SIMPLE project for the direct search for dark matter (DM) reached the most restrictive limits on the spin-dependent sector to date. The experiment is based on the detection of recoils following WIMP-nuclei interaction, mimicking those from neutron scattering. The thermodynamic operation conditions yield the SDDs intrinsically insensitive to radiations with linear energy transfer below ∼150 keVμm{sup -1} such as photons, electrons, muons and neutrons with energies below ∼40 keV. Underground facilities are increasingly employed for measurements in a low-level radiation background (DM search, gamma-spectroscopy, intrinsic soft-error rate measurements, etc.), where the rock overburden shields against cosmic radiation. In this environment the SDDs are sensitive only to α-particles and neutrons naturally emitted from the surrounding materials. Recently developed signal analysis techniques allow discrimination between neutron and α-induced signals. SDDs are therefore a promising instrument for low-level neutron and α measurements, namely environmental neutron measurements and α-contamination assays. In this work neutron measurements performed in the challenging conditions of the latest SIMPLE experiment (1500 mwe depth with 50-75 cm water shield) are reported. The results are compared with those obtained by detailed Monte Carlo simulations of the neutron background induced by {sup 238}U and {sup 232}Th traces in the facility, shielding and detector materials. Calculations of the neutron energy distribution yield the following neutron fluence rates (in 10{sup -8} cm{sup -2}s{sup -1}): thermal (<0.5 eV): 2.5; epithermal (0.5 eV-100 keV): 2.2; fast (>1 MeV): 3.9. Signal rates were derived using standard cross sections and codes routinely employed in reactor dosimetry. The measured and calculated neutron count rates per unit of active mass were 0.15 ct/kgd and 0.33 ct/kg-d respectively. As the major

Array detector for neutron pre-emission investigations

International Nuclear Information System (INIS)

Petrascu, M.; Cruceru, I.; Bordeanu, C.

1999-01-01

It was predicted that in a fusion experiment induced by 11 Li halo nuclei on light targets, due to the very large dimension of 11 Li, one may expect that the valence neutrons will not be absorbed together with the 9 Li core, but will be emitted in the early stage of the fusion process. The experiment aiming at checking this expectation was performed at the RIKEN-RIPS facility. It was found from neutron energy spectra measurements, that an important number of fusions, more than 30%, are preceded by the pre-emission of one or two neutrons. In the position spectra measurements a very narrow neutron component has been found. This component is much narrower than that calculated by using the Cluster Shell Model Approximation (COSMA). The recent results of time- position coincidence measurements show that within the narrow component the neutrons are pre-emitted predominantly as neutron pairs. The Program Advisory Committee of RIKEN has approved a new measurement at RIKEN Ring Cyclotron aiming at investigation of neutron-neutron coincidences by using a new neutron array detector. This detector has been recently accomplished within the collaboration existing between IFIN-HH, Romania and RIKEN, Japan. The array system consists of 81 4 x 4 x 12 cm 3 BC400 plastic scintillators each coupled to XP2972 Phototubes. The mounting and the testing of the new neutron array detector will be done at RIKEN. The components of one of the 81 elements of the array detector are shown in a photo. The Monte Carlo calculated neutron detection efficiencies as a function of energy are shown. This detector will be used for the investigation of neutron-neutron coincidences in the case of Si( 11 Li, fusion) reaction. The cross- talk between adjacent and non adjacent detectors will be determined by using a 9 Li beam. As it is known in the case of Si( 9 Li, fusion) the neutrons are of evaporation origin, and since these neutrons are emitted in 4 π the chance for detecting 2 coincident neutrons in the

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.

Development of resonant detectors for epithermal neutron spectroscopy at pulsed neutron sources

International Nuclear Information System (INIS)

Tardocchi, M.; Pietropaolo, A.; Senesi, R.; Andreani, C.; Gorini, G.

2004-01-01

New perspectives for epithermal neutron spectroscopy are opened by the development of new detectors for inverse geometry time of flight spectrometers at pulsed neutron sources. One example is the Very Low Angle Detector (VLAD) bank planned to be delivered, within the next 4 years, within the eVERDI project, on the neutron spectrometer VESUVIO, at the ISIS pulsed neutron source (UK). VLAD will extend the (q,ω) kinematical region for neutron scattering to low wavefactor transfer (q -1 ) still keeping energy transfer >1 eV, thus allowing the investigations of new experimental studies in condensed matter systems. The technique being developed for detection of epithermal neutrons, within this low q and high-energy transfer region, is the Resonance Detection Technique. In this work, the state of the detector development will be presented with special focus on the results obtained with some prototype detectors, namely YAP scintillators and cadmium-zinc-telluride semiconductors

Geometric optimization of a neutron detector based on a lithium glass–polymer composite

Energy Technology Data Exchange (ETDEWEB)

Mayer, M., E-mail: mike.f.mayer@gmail.com [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Nattress, J. [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Trivelpiece, C. [Materials Research Institute, The Pennsylvania State University, University Park, PA 16802 (United States); Jovanovic, I. [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States)

2015-06-01

We report on the simulation and optimization of a neutron detector based on a glass–polymer composite that achieves high gamma rejection. Lithium glass is embedded in polyvinyltoluene in three geometric forms: disks, rods, and spheres. Optimal shape, geometric configuration, and size of the lithium glass fragments are determined using Geant4 simulations. All geometrical configurations maintain an approximate 7% glass to polymer mass ratio. Results indicate a 125-mm diameter as the optimal detector size for initial prototype design achieving a 10% efficiency for the thermalization of incident fission neutrons from {sup 252}Cf. The geometrical features of a composite detector are shown to have little effect on the intrinsic neutron efficiency, but a significant effect on the gamma rejection is observed. The sphere geometry showed the best overall performance with an intrinsic neutron efficiency of approximately 6% with a gamma rejection better than 10{sup −7} for 280-μm diameter spheres. These promising results provide a motivation for prototype composite detector development based on the simulated designs. - Highlights: • Composite polymer–lithium glass scintillation detector is simulated. • Polymer is considered to be non-scintillating in the simulation. • Three forms of lithium glass are considered: disks, rods, and spheres. • Glass shape has a small effect on neutron efficiency. • Glass shape has a significant effect on gamma rejection.

Use of GEANT4 vs. MCNPX for the characterization of a boron-lined neutron detector

Energy Technology Data Exchange (ETDEWEB)

Ende, B.M. van der; Atanackovic, J.; Erlandson, A.; Bentoumi, G.

2016-06-01

This work compares GEANT4 with MCNPX in the characterization of a boron-lined neutron detector. The neutron energy ranges simulated in this work (0.025 eV to 20 MeV) are the traditional domain of MCNP simulations. This paper addresses the question, how well can GEANT4 and MCNPX be employed for detailed thermal neutron detector characterization? To answer this, GEANT4 and MCNPX have been employed to simulate detector response to a {sup 252}Cf energy spectrum point source, as well as to simulate mono-energetic parallel beam source geometries. The {sup 252}Cf energy spectrum simulation results demonstrate agreement in detector count rate within 3% between the two packages, with the MCNPX results being generally closer to experiment than are those from GEANT4. The mono-energetic source simulations demonstrate agreement in detector response within 5% between the two packages for all neutron energies, and within 1% for neutron energies between 100 eV and 5 MeV. Cross-checks between the two types of simulations using ISO-8529 {sup 252}Cf energy bins demonstrates that MCNPX results are more self-consistent than are GEANT4 results, by 3–4%.

Development of Optical Fiber Detector for Measurement of Fast Neutron

International Nuclear Information System (INIS)

YAGI, Takahiro; KAWAGUCHI, Shinichi; MISAWA, Tsuyoshi; PYEON, Cheol Ho; UNESAKI, Hironobu; SHIROYA, Seiji; OKAJIMA, Shigeaki; TANI, Kazuhiro

2008-01-01

Measurement of fast neutron flux is important for investigation of characteristic of fast reactors. In order to insert a neutron detector in a narrow space such as a gap of between fuel plates and measure the fast neutrons in real time, a neutron detector with an optical fiber has been developed. This detector consists of an optical fiber whose tip is covered with mixture of neutron converter material and scintillator such as ZnS(Ag). The detector for fast neutrons uses ThO 2 as converter material because 232 Th makes fission reaction with fast neutrons. The place where 232 Th can be used is limited by regulations because 232 Th is nuclear fuel material. The purpose of this research is to develop a new optical fiber detector to measure fast neutrons without 232 Th and to investigate the characteristic of the detector. These detectors were used to measure a D-T neutron generator and fast neutron flux distribution at Fast Critical Assembly. The results showed that the fast neutron flux distribution of the new optical fiber detector with ZnS(Ag) was the same as it of the activation method, and the detector are effective for measurement of fast neutrons. (authors)

Development of SiC Neutron Detector Assembly to Measure the Neutron Flux of the Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Se Hwan; Park, June Sic; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of)

2012-05-15

At present, the conventional detector to measure the neutron at harsh environment is a Self Powered Neutron Detector (SPND). Rhodium(Rh)-103 is in the SPND. When neutron is incident on the Rhodium, the neutron capture reaction occurs, and the Rh-103 is converted to Rh-104. The Rh-104 is decayed to Pd-104 by {beta}-decay, and electrons are generated as the decay products. Because of the half life of Rh-104, approximately 5 minutes are required for the SPND output to reach the equilibrium condition. Therefore the on-line monitoring of the nuclear reactor state is limited if the neutron flux in the reactor core is monitored with the SPND. Silicon carbide (SiC) has the possibility to be developed as neutron detector at harsh environment, because the SiC can be operative at high temperature and high neutron flux conditions. Previously, the basic operation properties of the SiC detector were studied. Also, the radiation response of the SiC detector was studied at high neutron and gamma dose rate. The measurement results for an ex-core neutron flux monitor or a neutron flux monitor of the spent fuel were published. The SiC detector was also developed as neutron detector to measure the fissile material with active interrogation method. However, the studies about the development of SiC detector are still limited. In the present work, the radiation damage effect of the SiC detector was studied. The detector structure was determined based on the study, and a neutron detector assembly was made with the SiC detectors. The neutron and gamma-ray response of the detector assembly is presented in this paper. The detector assembly was positioned in the HANARO research reactor core, the performance test was done. The preliminary results are also included in this paper

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

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.

Preparation of the in-house neutron detectors and the software needed to process experimental data

International Nuclear Information System (INIS)

Haddad, Kh.; Haj-Hassan, H.; Helal, W.

2007-04-01

In - house neutron activation detectors were prepared in this work using pure commercial gold. The neutron self-shielding factors in the foils for both thermal and epithermal neutrons have been determined experimentally. The work shows good results repeatability and good agreement with certified activation monitors. the software KHW for neutron flux measurements using local and standards gold foils was designed and performed locally. it deals as well with irradiated uranium spectrums to calculate some important fission product ratios for neutron flux measurement. Some experiments were performed to investigate the possibility of using uranium, produced in the pilot plant, as fission neutron detector. The results shows the possibility of using fission product ratios to determine the cooling time of the samples. It shows also the possibility of using fission and activation product ratios as an indicators of neutron fluences ratios.(author)

Neutron detector using sol-gel absorber

Science.gov (United States)

Hiller, John M.; Wallace, Steven A.; Dai, Sheng

1999-01-01

An neutron detector composed of fissionable material having ions of lithium, uranium, thorium, plutonium, or neptunium, contained within a glass film fabricated using a sol-gel method combined with a particle detector is disclosed. When the glass film is bombarded with neutrons, the fissionable material emits fission particles and electrons. Prompt emitting activated elements yielding a high energy electron contained within a sol-gel glass film in combination with a particle detector is also disclosed. The emissions resulting from neutron bombardment can then be detected using standard UV and particle detection methods well known in the art, such as microchannel plates, channeltrons, and silicon avalanche photodiodes.

Application of nuclear pumped laser to an optical self-powered neutron detector

Science.gov (United States)

Yamanaka, N.; Takahashi, H.; Iguchi, T.; Nakazawa, M.; Kakuta, T.; Yamagishi, H.; Katagiri, M.

1996-05-01

A Nuclear Pumped Laser (NPL) using 3He/Ne/Ar gas mixture is investigated for a purpose of applying to an optical self-powered neutron detector. Reactor experiments and simulations on lasing mechanism have been made to estimate the best gas pressure and mixture ratios on the threshold input power density (or thermal neutron flux) in 3He/Ne/Ar mixture. Calculational results show that the best mixture pressure is 3He/Ne/Ar=2280/60/100 Torr and thermal neutron flux threshold 5×1012 n/cm2 sec, while the reactor experiments made in the research reactor ``YAYOI'' of the University of Tokyo and ``JRR-4'' of JAERI also demonstrate that excitational efficiency is maximized in a similar gas mixture predicted by the calculation.

Comparison study on in-core neutron detector for online neutron flux mapping of research and power reactor

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Mohd Idris Taib; Izhar Abu Husin; Nurfarhana Ayuni

2010-01-01

This paper presents the comparison study on In-Core neutron detector using for online flux mapping of Research and Power reactor. Technical description of in-core neutron also taken into consideration to identify the different characterization of neutron detector and describe on Self Power neutron detector (SPND) for online neutron flux mapping. Able to provide information on the neutron flux distribution and understand how in-core neutron detector are being used in nuclear power plant including to enable to state the principles of neutron detector. (author)

Using track detectors in neutron dosimetry

International Nuclear Information System (INIS)

Spurny, F.; Turek, K.

1977-01-01

The usage of track detectors of charged particles provides a new possibility of neutron dosimetry. Presented is a comparison of the main dosimetric characteristics of three various types of track detectots of fast neutrons, i.e. glass in the contact with 232 Th; KODAK LR115 cellulose nitrate; MAKROFOL E polycarbonate. Results of studing energy dependences of detectors are presented. Results obtained using phantoms under radiation fields of various sources of complex gamma-neutron radiation are discussed [ru

Fabrication of boron-phosphide neutron detectors

International Nuclear Information System (INIS)

Fitzsimmons, M.; Pynn, R.

1997-01-01

Boron phosphide is a potentially viable candidate for high neutron flux neutron detectors. The authors have explored chemical vapor deposition methods to produce such detectors and have not been able to produce good boron phosphide coatings on silicon carbide substrates. However, semi-conducting quality films have been produced. Further testing is required

Self-powered neutron detector

International Nuclear Information System (INIS)

Goldstein, N.P.; Todt, W.H.

1976-01-01

A self-powered neutron detector is detailed wherein a thin conductive layer of low neutron cross section, high density material is disposed about an emitter core of material which spontaneously emits radiation on neutron capture. The high density material is absorptive of beta radiation emitted by decay of the emitter core activation product, but is substantially transmissive to the high average energy prompt electrons emitted by the emitter core material. (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.

Measurements of fast neutrons by bubble detectors

Energy Technology Data Exchange (ETDEWEB)

Castillo, F.; Martinez, H. [Laboratorio de Espectroscopia, Instituto de Ciencias Fisicas, Universidad Nacional Autonoma de Mexico, Apartado Postal 48-3, 62251, Cuernavaca Morelos (Mexico); Leal, B. [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Apartado Postal 70-543, 04510, Ciudad Universitaria, Mexico D. F. (Mexico); Rangel, J. [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Apartado Postal 70-543, 04510, Ciudad Universitaria, Mexico D. F (Mexico); Reyes, P. G. [Facultad de Ciencias, Universidad Autonoma del Estado de Mexico, Instituto Literario 100, Col. Centro, 50000, Toluca Estado de Mexico (Mexico)

2013-07-03

Neutron bubble detectors have been studied using Am-Be and D-D neuron sources, which give limited energy information. The Bubble Detector Spectrometer (BDS) have six different energy thresholds ranging from 10 KeV to 10 Mev. The number of bubbles obtained in each measurement is related to the dose (standardized response R) equivalent neutrons through sensitivity (b / {mu}Sv) and also with the neutron flux (neutrons per unit area) through a relationship that provided by the manufacturer. Bubble detectors were used with six different answers (0.11 b/ {mu}Sv, 0093 b/{mu}Sv, 0.14 b/{mu}Sv, 0.17 b/{mu}Sv, 0051 b/{mu}Sv). To test the response of the detectors (BDS) radiate a set of six of them with different energy threshold, with a source of Am-Be, placing them at a distance of one meter from it for a few minutes. Also, exposed to dense plasma focus Fuego Nuevo II (FN-II FPD) of ICN-UNAM, apparatus which produces fusion plasma, generating neutrons by nuclear reactions of neutrons whose energy emitting is 2.45 MeV. In this case the detectors were placed at a distance of 50 cm from the pinch at 90 Degree-Sign this was done for a certain number of shots. In both cases, the standard response is reported (Dose in {mu}Sv) for each of the six detectors representing an energy range, this response is given by the expression R{sub i}= B{sub i} / S{sub i} where B{sub i} is the number of bubbles formed in each and the detector sensitivity (S{sub i}) is given for each detector in (b / {mu}Sv). Also, reported for both cases, the detected neutron flux (n cm{sup -2}), by a given ratio and the response involves both standardized R, as the average cross section sigma. The results obtained have been compared with the spectrum of Am-Be source. From these measurements it can be concluded that with a combination of bubble detectors, with different responses is possible to measure the equivalent dose in a range of 10 to 100 {mu}Sv fields mixed neutron and gamma, and pulsed generated fusion

Neutron radiation damage studies on silicon detectors

International Nuclear Information System (INIS)

Li, Zheng; Chen, W.; Kraner, H.W.

1990-10-01

Effects of neutron radiation on electrical properties of Si detectors have been studied. At high neutron fluence (Φ n ≥ 10 12 n/cm 2 ), C-V characteristics of detectors with high resistivities (ρ ≥ 1 kΩ-cm) become frequency dependent. A two-trap level model describing this frequency dependent effect is proposed. Room temperature anneal of neutron damaged (at LN 2 temperature) detectors shows three anneal stages, while only two anneal stages were observed in elevated temperature anneal. 19 refs., 14 figs

A new detector for the measurement of neutron flux in nuclear reactors; Nouvelle methode de mesure des flux de neutrons dans les reacteurs atomiques

Energy Technology Data Exchange (ETDEWEB)

Koch, L; Labeyrie, J; Tarassenko, S [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high {gamma} ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their {beta} and {gamma} disintegration. Thermal neutron fluxes as low as few neutrons per cm{sup 2} per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm{sup 2}) for fluxes higher than 10{sup 8} n/cm{sup 2}/s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author)Fren. [French] On decrit un appareil permettant la mesure instantanee des flux de neutrons thermiques accompagnes de flux intenses de rayons {gamma} et situes dans des enceintes pouvant etre portees a des temperatures superieures a 500 deg. C. On utilise la radioactivite des atomes resultant de la fission des noyaux lourds; ces atomes sont entraines par un courant gazeux vers un detecteur de radioactivite qui enregistre leurs desintegrations {beta} et {gamma}. On peut mesurer des flux partir de quelques neutrons thermiques par cm{sup 2} et par seconde. L'appareil permet de suivre la puissance d'un reacteur atomique, de tracer des cartes de densite de neutrons avec une tres bonne definition (1 mm{sup 2}) dans le cas de flux superieurs a 10{sup 8} cm{sup 2}/s. Le temps de reponse du systeme a une variation du flux de neutrons est limite, poes flux importants, par le temps de transit du gaz entre l'emetteur de produits de fission et le detecteur: soit quelques dizaines de

Thermal flow in detectors of CNA-II with spontaneous fissions source of 238U

International Nuclear Information System (INIS)

Mascitti, J. A

2012-01-01

The thermal flux in the position of ex-core and in-core CNA-II Nuclear Power Plant (CNA-II) detectors is estimated considering neutron from the 238 U spontaneous fissions as the source, for the reactor cold state (isothermal state with both coolant and moderator at a temperature of 60 o C, a pressure of 35 ata and 15.46 ppm of natural Boron), and 24% inserted control rods (slightly sub-critical). Results are obtained for two different situations: with and without photo-neutrons due to the (γ,n) reaction in D 2 O. It is concluded that the thermal flux is under the detection limit of the boron trifluoride 104-SR or 282-IB detectors (≅10 -1 cm-2.s -1 ). These detectors are located in opposite positions in the inner concrete shielding, having the lowest detection limit among all ex-core detectors. A significant difference is verified in neutron fluxes between both cases, which suggest that photo-neutrons in large heavy water reactors such as CNA-II should not be ignored. The total neutron flux attenuation factor between the inner and outer region of the reactor pressure vessel was estimated to be 7.0 x 10 -7 . It should be mentioned that none of the results here presented has been affected by any correction factor. Each value has a percentage relative error representing the statistical uncertainty due to the probabilistic Monte Carlo method used to obtain it (author)

Neutron detector with monitoring elements

International Nuclear Information System (INIS)

Haller, P.

1976-01-01

To check the reliable reading of a neutron detector the signal of which results from (n,e) processes and which is used for neutron flux supervision in the reactor core of pressurized-water reactors, a circuit is given which makes it possible to record the isolation resistivity of the cable connected to the input of the current amplifier and of the neutron detector, this resistivity determining, among others, the output signal. For supervision, the input offset voltage of the current amplifier is modulated by a low-frequency ac voltage and a filter is assigned to the output of an op amplifier, this filter feeding a limiting value recorder. (ORU) [de

Transient response of self-powered neutron detectors

International Nuclear Information System (INIS)

Boeck, H.; Gebureck, P.; Stegemann, D.

The behaviour of self-powered neutron detectors with Co, Er, Hf and Pt emitters was investigated during reactor square wave and pulse operation. The detector's response was compared with the current of an excore ionization chamber. Characteristical deviations from linearity were observed with all detectors at fast reactor periods. The exact cause of these deviations is not yet fully understood but several possibilities for the nonlinear behaviour of self-powered neutron detectors are outlined. (author)

Semi-insulating GaAs detectors of fast neutrons

International Nuclear Information System (INIS)

Sagatova, A.; Sedlackova, K.; Necas, V.; Zatko, B.; Dubecky, F.; Bohacek, P.

2012-01-01

The present work deals with the technology of HDPE neutron conversion layer application on the surface of semi-insulating (SI) GaAs detectors via developed polypropylene (PP) based glue. The influence of glue deposition on the electric properties of the detectors was studied as well as the ability of the detectors to register the fast neutrons from "2"3"9Pu-Be neutron source. (authors)

Sample dependent response of a LaCl{sub 3}:Ce detector in prompt gamma neutron activation analysis of bulk hydrocarbon samples

Energy Technology Data Exchange (ETDEWEB)

Naqvi, A.A., E-mail: aanaqvi@kfupm.edu.sa [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); Al-Matouq, Faris A.; Khiari, F.Z. [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); Isab, A.A. [Department of Chemistry, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); Khateeb-ur-Rehman,; Raashid, M. [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia)

2013-08-11

The response of a LaCl{sub 3}:Ce detector has been found to depend upon the hydrogen content of bulk samples in prompt gamma analysis using 14 MeV neutron inelastic scattering. The moderation of 14 MeV neutrons from hydrogen in the bulk sample produces thermal neutrons around the sample which ultimately excite chlorine capture gamma rays in the LaCl{sub 3}:Ce detector material. Interference of 6.11 MeV chlorine gamma rays from the detector itself with 6.13 MeV oxygen gamma rays from the bulk samples makes the intensity of the 6.13 MeV oxygen gamma ray peak relatively insensitive to variations in oxygen concentration. The strong dependence of the 1.95 MeV doublet chlorine gamma ray yield on hydrogen content of the bulk samples confirms fast neutron moderation from hydrogen in the bulk samples as a major source of production of thermal neutrons and chlorine gamma rays in the LaCl{sub 3}:Ce detector material. Despite their poor oxygen detection capabilities, these detectors have nonetheless excellent detection capabilities for hydrogen and carbon in benzene, butyl alcohol, propanol, propanic acid, and formic acid bulk samples using 14 MeV neutron inelastic scattering.

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)

Detection efficiency of the neutron detector BELEN-48 measured at the PTB Braunschweig

Energy Technology Data Exchange (ETDEWEB)

Marta, Michele [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); II. Physikalisches Institut, Justus-Liebig Universitaet Giessen (Germany); Agramunt, Jorge; Tain, Jose Luis [IFIC-CSIC University of Valencia, Valencia (Spain); Caballero-Folch, Roger; Cortes, Guillem; Riego, Albert [INTE-DFEN, Universitat Politecnica de Catalunya, Barcelona (Spain); Dillmann, Iris [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); II. Physikalisches Institut, Justus-Liebig Universitaet Giessen (Germany); TRIUMF, Vancouver (Canada); Erhard, Martin; Giesen, Ulrich; Nolte, Ralf; Roettger, Stefan [Physikalisch-Technische Bundesanstalt (PTB), Braunschweig (Germany); Fraile, Luis M. [Universidad Complutense de Madrid (Spain)

2014-07-01

The BEta-deLayEd Neutron detector BELEN-48 is a highly efficient detector of β-delayed neutrons, for nuclear structure, nuclear astrophysics and reactor studies. It consists of 48 {sup 3}He proportional counters arranged in a polyethylene matrix in a way that the detection efficiency remains constant for neutron energies from thermal up to a few MeV. In order to validate MCNPX simulations, the detection efficiency has been calibrated with well-known (p,n) and (α,n) reactions on {sup 7}Li, {sup 13}C and {sup 51}V producing neutrons with energies between 0.1 and 5 MeV. The experiment has been performed at the neutron metrology facility of PTB, which allowed the measurement of yields and angular distributions with a calibrated monitor. The new results indicate anisotropies, which are not reported in literature and have been taken into account to obtain the experimental efficiencies for BELEN.

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)

Hot nuclei studies with a 4 π-Neutron detector

International Nuclear Information System (INIS)

Galin, J.; Crema, E.; Doubre, H.; Guerreau, D.; Jiang, D.X.; Morjean, M.; Piasecki, E.; Pouthas, J.; Saint-Laurent, F.; Sokolov, A.; Wang, X.D.; Charvet, J.L.; Frehaut, J.; Lott, B.; Magnago, C.; Patin, Y.; Gatty, B.; Jacquet, D.; Lott, B.

1989-01-01

When studying the behavior of hot nuclei, the challenge is twofold: how are they formed in nucleus-nucleus collisions and how do they decay Systematic studies have been undertaken at GANIL by bombarding heavy targets (Au, Th) with intermediate energy projectiles (27, 35, 44 and 77 MeV/u Ar and 32 MeV/u Kr). For such neutron rich systems the thermalized energy can be roughly evaluated simply by measuring the multiplicity of evaporated neutrons using a 4 π Gd loaded, liquid scintillator detector. The influence of the bombarding energy and projectile mass on the energy dissipation has been investigated for different exit channels and special emphasis has been put in the study of the most dissipative collisions. The thermal energy, derived by summing the energy removed by both evaporated neutrons and light charged particles is shown to saturate at E * ≅ 650 MeV in the Ar induced reactions between 27 and 77 MeV/u, in good consistency with the predictions of semi-classical Landau-Vlasov calculations

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

Development of a neutron detector with high detection efficiency and high spatial resolution and its applications to reactor physics experiments

International Nuclear Information System (INIS)

Tojo, Takao

1979-09-01

For detection of thermal neutrons in multiplying systems, a scintillator mixture of ZnS(Ag), 6 LiF and polyethylene was prepared, and its characteristics were shown. A sintillation detector using the mixture and a long acrylic-resin light guide was developed for measuring thermal neutrons in an U-H 2 O subcritical assembly(JAERISA). The detector was applied in the following reactor physics measurements with JAERISA: (1) cadmium ratio, (2) infinite multiplication factor, (3) material buckling, and (4) prompt neutron lifetime by pulsed neutron method. These experiments revealed that neutrons in the assembly are successfully detected by the detector owing to its outstanding characteristics of gamma-ray insensitivity, high detection efficiency and high spatial resolution. In the process of activity measurement of a foil activation detector with a GM counter, it was shown that accurate counting loss correction are difficult by usual method, because of the appreciable resolving time dependence on counting rates. In accurate correction, a new method was introduced for precise measurement of the resolving time; the dependence was made clear. A new correction method was developed, which enables direct reading of the corrected counting rates, even at high counting rates. (author)

A large-area, position-sensitive neutron detector with neutron/γ-ray discrimination capabilities

International Nuclear Information System (INIS)

Zecher, P.D.; Galonsky, A.; Kruse, J.J.; Gaff, S.J.; Ottarson, J.; Wang, J.; Seres, Z.; Ieki, K.; Iwata, Y.; Schelin, H.

1997-01-01

To further study neutron-rich halo nuclei, we have constructed a neutron detector array. The array consists of two separate banks of detectors, each of area 2 x 2 m 2 and containing 250 l of liquid scintillator. Each bank is position-sensitive to better than 10 cm. For neutron time-of-flight measurements, the time resolution of the detector has been demonstrated to be about 1 ns. By using the scintillator NE-213, we are able to distinguish between neutron and γ-ray signals above 1 MeV electron equivalent energy. Although the detector array was constructed for a particular experiment it has also been used in a number of other experiments. (orig.)

Neutron detector for fusion reaction-rate measurements

International Nuclear Information System (INIS)

Lerche, R.A.; Phillion, D.W.; Tietbohl, G.L.

1993-01-01

We have developed a fast, sensitive neutron detector for recording the fusion reaction-rate history of inertial-confinement fusion (ICF) experiments. The detector is based on the fast rise-time of a commercial plastic scintillator (BC-422) and has a response 7 neutrons

Design of neutron detectors utilising luminescent glass

International Nuclear Information System (INIS)

Spowart, A.R.

1983-01-01

Impetus for the development of new neutron detector designs has derived from the worldwide commissioning of neutron spallation sources. The design concepts, and principal methods of utilisation of these major installations, have been recently reviewed. Their principal feature of interest is their broadband neutron emission allowing neutron investigations of all types of structure in materials from biological molecules to steels. Conventional neutron detectors are gas-filled devices, based on BF/sub 3/ or /sup 3/He gas. Their major advantage is their intrinsically low background count. Their principal disadvantage is their slow response time (10-100 μs), high cost and relative lack of flexibility in design to cope with large areas or complex geometry detection. They are, however, long established and the research facilities around the world have a heavy investment in the interpretative hardware for gas detectors

Evaluation of B10Plus+* proportional detectors for neutron coincidence counting

Energy Technology Data Exchange (ETDEWEB)

Beddingfield, David H.; Yoon, Seokryung [International Atomic Energy Agency, Vienna International Centre, PO Box 100, 1400 Vienna, (Austria)

2015-07-01

GE-Reuter-Stokes (GERS) has developed a new line of neutron proportional counters, the B10Plus+* proportional counter. The detector design is intended to serve as a cost-effective alternative to traditional {sup 3}He proportional counters in a variety of applications. The detector is a hybrid design 10B-lined tube optimized with the addition of a small quantity of 3He gas to improve the detector performance and efficiency. As a demonstration of the B10Plus+* detector, GERS has constructed a Uranium Neutron Collar (UNCL) system consisting of B-10Plus+* proportional counters. GERS has designed and built a demonstration UNCL system intended to match the performance of a Type-I UNCL design in Pressurized Water Reactor (PWR) geometry operating in thermal mode. GERS offered their system on loan to the International Atomic Energy Agency (IAEA) Safeguards Division of Technical and Scientific Services for an assessment of the detector technology and the demonstration system. We have characterized the demonstration UNCL system and compared its performance with a traditional Type-I UNCL design in regular use by the IAEA. This paper summarizes our findings and observations during the characterization and testing activity. (authors)

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

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

The response of the BTI bubble detectors in mixed gamma-neutron workplace fields

International Nuclear Information System (INIS)

Vanhavere, F.; Coeck, M.; Lievens, B.; Reginatto, M.

2005-01-01

Full text: Bubble detectors have become a mature technology and are used as neutron dosemeters in a wide range of applications. At the SCK-CEN and Belgonucleaire they are used as official personal neutron dosemeter for the personnel. Two types are commercially available from Bubble Technology Industries: the BD-PND, which has a neutron energy threshold of around 100 keV, and the BDT, which is mainly sensitive to thermal neutrons. At Belgonucleaire only the BD-PND is worn, and the results are corrected with a site specific factor. At the SCK-CEN both the BD-PND and BDT are worn and a combination of both results is applied for the dose records. In the EC project EVIDOS (Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields), a whole range of neutron dosemeters were irradiated in workplace fields in nuclear installations in Europe, including both types of bubble detectors. The bubble detectors were exposed on a phantom with different angles towards the reference directions in the workplace fields. We will report the bubble detectors' results in the simulated workplace fields at Cadarache (CANEL and Sigma), in the workplaces at Kruemmel (boiling water reactor, transport cask), at Mol (Venus research reactor SCK-CEN, MOX-fuel facility Belgonucleaire) and Ringhals (pressurized water reactor, transport cask). The responses of the bubble detectors and the combination of both will be compared to the reference values determined with Bonner Spheres and a novel directional spectrometer. The dosemeter readings were checked for consistency by folding the dosemeter response functions with the corresponding workplace fluence spectra in the same workplace. (author)

Neutron detectors for nuclear reactor control

International Nuclear Information System (INIS)

Duchene, Jean; Verdant, Robert.

1974-01-01

In view of the importance of in-core measurements the distinction is made between detectors used outside and inside the core. In the former case proportional counters, fission chambers and boron chambers are reviewed in turn. The only in-core detectors considered are those giving a direct measurement, i.e. supplying an electric signal representative of the neutron fluence rate while in the measurement position at the point given. Two kinds of detectors are used for direct measurements: miniature fission chambers and collectors, known also as neutron-electron converters [fr

Neutron generation time of the reactor 'crocus' by an interval distribution method for counts collected by two detectors

International Nuclear Information System (INIS)

Haldy, P.-A.; Chikouche, M.

1975-01-01

The distribution is considered of time intervals between a count in one neutron detector and the consequent event registered in a second one. A 'four interval' probability generating function was derived by means of which the expression for the distribution of the time intervals, lasting from triggering detection in the first detector to subsequent count in the second, one could be obtained. The experimental work was conducted in the zero thermal power reactor Crocus, using a neutron source provided by spontaneous fission, a BF 3 counter for the first detector and an He 3 detector for the second instrument. (U.K.)

Activation method for measuring the reaction rates and studying the neutron spectra parameters, based on using the unified composition detectors

International Nuclear Information System (INIS)

Demidov, A.M.; Dikarev, V.S.; Efimov, B.V.; Ionov, V.S.; Marin, S.V.

2005-01-01

The method proposed for estimation of parameters thermal and epithermal parts of energy distribution of neutrons is described. The method based on application of activation measuring with use of unified composition detectors (UCD) and samples of fuel. The method is applicable for definition of neutron spectrum parameters and velocities of division in fuel of nuclear installations. Theoretical bases and the description of a method, expedients of manufacturing and calibration for the detectors, the experimental data, carried out in RRC KI are given and processing of experimental data, and also. The parametric model of a spectrum constructed on the basis of Westcott's formalism is described. The parameter of stiffness is entered and its role for temperature of neutron gas, spectral coefficients of isotopes of detectors, the transition area thermal and epithermal parts of neutron spectra is observationally appreciated. It is offered to confirm the found results by calculations with use of MCU Monte Carlo code [ru

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)

Designing a new type of neutron detector for neutron and gamma-ray discrimination via GEANT4

International Nuclear Information System (INIS)

Shan, Qing; Chu, Shengnan; Ling, Yongsheng; Cai, Pingkun; Jia, Wenbao

2016-01-01

Design of a new type of neutron detector, consisting of a fast neutron converter, plastic scintillator, and Cherenkov detector, to discriminate 14-MeV fast neutrons and gamma rays in a pulsed n–γ mixed field and monitor their neutron fluxes is reported in this study. Both neutrons and gamma rays can produce fluorescence in the scintillator when they are incident on the detector. However, only the secondary charged particles of the gamma rays can produce Cherenkov light in the Cherenkov detector. The neutron and gamma-ray fluxes can be calculated by measuring the fluorescence and Cherenkov light. The GEANT4 Monte Carlo simulation toolkit is used to simulate the whole process occurring in the detector, whose optimum parameters are known. Analysis of the simulation results leads to a calculation method of neutron flux. This method is verified by calculating the neutron fluxes using pulsed n–γ mixed fields with different n/γ ratios, and the results show that the relative errors of all calculations are <5%. - Highlights: • A neutron detector is developed to discriminate 14-MeV fast neutrons and gamma rays. • The GEANT4 is used to optimize the parameters of the detector. • A calculation method of neutron flux is established through the simulation. • Several n/γ mixture fields are simulated to validate of the calculation method.

Detectors that don’t fear neutrons

CERN Multimedia

Antonella Del Rosso

2013-01-01

High-intensity pulsed neutron fields are produced at particle accelerators such as CERN’s PS and LHC. The efficient detection of this stray pulsed radiation is technically difficult and standard detectors show strong limitations when measuring such fields. A new test performed at the HiRadMat facility has recently shed light on the performance of various neutron detectors exposed to extreme conditions.   In order to limit the required human intervention to the beginning and the end of the test, detectors were mounted on a dedicated wheel that CERN’s HiRadMat team built for the HRMT-15 experiment. High-intensity pulsed neutron fields are among the toughest conditions a detector can be asked to face. Particle accelerators produce such stray radiation when primary beams are dumped or lost because of, for example, an orbit instability that can occur during ordinary operation. Accurately measuring the radiation levels is the first requirement in order for experts to be able to...

New thermal neutron calibration channel at LNMRI/IRD

International Nuclear Information System (INIS)

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

2015-01-01

A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four 241 Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

A multilayer surface detector for ultracold neutrons

Energy Technology Data Exchange (ETDEWEB)

Wang, Zhehui, E-mail: zwang@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Hoffbauer, M.A.; Morris, C.L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Callahan, N.B.; Adamek, E.R. [Indiana University, Bloomington, IN 47405 (United States); Bacon, J.D. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Blatnik, M. [Cleveland State University, Cleveland, OH 44115 (United States); Brandt, A.E. [North Carolina State University, Raleigh, NC 27695 (United States); Broussard, L.J.; Clayton, S.M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Cude-Woods, C. [North Carolina State University, Raleigh, NC 27695 (United States); Currie, S. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Dees, E.B. [North Carolina State University, Raleigh, NC 27695 (United States); Ding, X. [Virginia Polytechnic Institute and State University, Blacksburg, VA 24061 (United States); Gao, J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Gray, F.E. [Regis University, Denver, CO 80221 (United States); Hickerson, K.P. [University of California Los Angeles, Los Angeles, CA 90095 (United States); Holley, A.T. [Tennessee Technological University, Cookeville, TN 38505 (United States); Ito, T.M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Liu, C.-Y. [Indiana University, Bloomington, IN 47405 (United States); and others

2015-10-21

A multilayer surface detector for ultracold neutrons (UCNs) is described. The top {sup 10}B layer is exposed to vacuum and directly captures UCNs. The ZnS:Ag layer beneath the {sup 10}B layer is a few microns thick, which is sufficient to detect the charged particles from the {sup 10}B(n,α){sup 7}Li neutron-capture reaction, while thin enough that ample light due to α and {sup 7}Li escapes for detection by photomultiplier tubes. A 100-nm thick {sup 10}B layer gives high UCN detection efficiency, as determined by the mean UCN kinetic energy, detector materials, and other parameters. Low background, including negligible sensitivity to ambient neutrons, has also been verified through pulse-shape analysis and comparison with other existing {sup 3}He and {sup 10}B detectors. This type of detector has been configured in different ways for UCN flux monitoring, development of UCN guides and neutron lifetime research.

Evaluation of power history during power burst experiments in TRACY by combination of gamma-ray and thermal neutron detectors

International Nuclear Information System (INIS)

Yanagisawa, Hiroshi; Ohno, Akio

2002-01-01

A combination method using γ-ray and thermal neutron detectors was newly applied to the accurate evaluation of power histories during reactivity-initiated power burst experiments in the Transient Experiment Critical Facility (TRACY). During an initial power burst, the power history was determined using a fast response γ-ray ionization chamber, which was used because of its ability to exactly trace the power history within a short duration of the initial burst. After the initial burst, a micro fission chamber containing highly enriched uranium was used for the determination of the power history because the γ-ray ionization chamber could not be applied due to the contribution of delayed γ-rays from fission products. By the present method, the power histories were evaluated for the experiments in the range of 1.50 to 2.93$ of the reactivity insertion. It was found that the peak power and integrated power as determined by the previous method using only the micro fission chamber were underestimated to be 40% and 30% in maximum, respectively, in comparison with the results from the present evaluation. The numerical simulation performed by using the Monte Carlo method indicated that the underestimation could be comprehended by considering the time delay of thermal neutron detection of the fission chamber, which arose from the flight-time of neutrons from the TRACY core to the fission chamber. (author)

Polycrystalline semiconductor probes for monitoring the density distribution of an intense thermal neutron flux in nuclear reactors

International Nuclear Information System (INIS)

Graul, J.; Mueller, R.G.; Wagner, E.

1975-05-01

The applicability of semiconductor detectors for high thermal neutron flux densities is theoretically estimated and experimentally examined. For good thermal stability and low radiation capture rate silicon carbide is used as semiconductor material, produced in polycristalline layers to achieve high radiation resistance. The relations between crystallinity, photoelectric sensitivity and radiation resistance are shown. The radiation resistance of polycrystalline SiC-probes is approximately 100 times greater than that of conventional single crystal radiation detectors. For thermal neutron measurement they can be used in the flux range of approx. 10 10 13 (cm -2 sec -1 ) with operation times of 1.6 a >= tsub(b,max) >= 30 d, resp. (orig.) [de

Measurements for the energy calibration of the TANSY neutron detectors

International Nuclear Information System (INIS)

Drozdowicz, K.; Hoek, M.; Aronsson, D.

1990-05-01

The report describes measurements performed for the energy calibration of the TANSY neutron detectors (two arrays of 16 detectors each one). The calibration procedure determines four calibration parameters for each detector. Results of the calibration measurements are given and test measurements are presented. A relation of the neutron detector calibration parameters to producer's data for the photomulipliers is analysed. Also the tests necessary during normal operation of the TANSY neutron spectrometer are elaborated (passive and active tests). A method how to quickly get the calibration parameters for a spare detector in an array of the neutron detectors is included

Neutron counting and gamma spectroscopy with PVT detectors

International Nuclear Information System (INIS)

Mitchell, Dean James; Brusseau, Charles A.

2011-01-01

Radiation portals normally incorporate a dedicated neutron counter and a gamma-ray detector with at least some spectroscopic capability. This paper describes the design and presents characterization data for a detection system called PVT-NG, which uses large polyvinyl toluene (PVT) detectors to monitor both types of radiation. The detector material is surrounded by polyvinyl chloride (PVC), which emits high-energy gamma rays following neutron capture reactions. Assessments based on high-energy gamma rays are well suited for the detection of neutron sources, particularly in border security applications, because few isotopes in the normal stream of commerce have significant gamma ray yields above 3 MeV. Therefore, an increased count rate for high-energy gamma rays is a strong indicator for the presence of a neutron source. The sensitivity of the PVT-NG sensor to bare 252 Cf is 1.9 counts per second per nanogram (cps/ng) and the sensitivity for 252 Cf surrounded by 2.5 cm of polyethylene is 2.3 cps/ng. The PVT-NG sensor is a proof-of-principal sensor that was not fully optimized. The neutron detector sensitivity could be improved, for instance, by using additional moderator. The PVT-NG detectors and associated electronics are designed to provide improved resolution, gain stability, and performance at high-count rates relative to PVT detectors in typical radiation portals. As well as addressing the needs for neutron detection, these characteristics are also desirable for analysis of the gamma-ray spectra. Accurate isotope identification results were obtained despite the common impression that the absence of photopeaks makes data collected by PVT detectors unsuitable for spectroscopic analysis. The PVT detectors in the PVT-NG unit are used for both gamma-ray and neutron detection, so the sensitive volume exceeds the volume of the detection elements in portals that use dedicated components to detect each type of radiation.

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.))

Scintillating fibre tracking neutron detector

International Nuclear Information System (INIS)

Karlsson, Joakim.

1995-04-01

A detector for measurements of collimated fluxes of neutrons in the energy range 2-20 MeV is proposed. It utilizes (n.p) elastic scattering in scintillating optical fibres placed in successive orthogonal layers perpendicular to the neutron flux. A test module has been designed, constructed and tested with respect to separation of neutron and gamma events. The pulse height measurements show the feasibility to discriminate between neutron, gamma and background events. Application to measurements of fusion neutrons is considered. 18 refs, 22 figs, 4 tabs

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

Neutron multiplicity measurements with ³He alternative: Straw neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Mukhopadhyay, Sanjoy [Arnold Avenue Andrews AFB, Joint Base Andrews, MD (United States); Wolff, Ronald [Arnold Avenue Andrews AFB, Joint Base Andrews, MD (United States); Detwiler, Ryan [Arnold Avenue Andrews AFB, Joint Base Andrews, MD (United States); Maurer, Richard [Arnold Avenue Andrews AFB, Joint Base Andrews, MD (United States); Mitchell, Stephen [National Security Technologies, LLC, Las Vegas, NV (United States); Guss, Paul [Remote Sensing Lab. - Nellis, Las Vegas, NV (United States); Lacy, Jeffrey L. [Proportional Technologies, Inc., Houston, TX (United States); Sun, Liang [Proportional Technologies, Inc., Houston, TX (United States); Athanasiades, Athanasios [Proportional Technologies, Inc., Houston, TX (United States)

2015-01-27

Counting neutrons emitted by special nuclear material (SNM) and differentiating them from the background neutrons of various origins is the most effective passive means of detecting SNM. Unfortunately, neutron detection, counting, and partitioning in a maritime environment are complex due to the presence of high-multiplicity spallation neutrons (commonly known as ‘‘ship effect ’’) and to the complicated nature of the neutron scattering in that environment. A prototype neutron detector was built using ¹⁰B as the converter in a special form factor called ‘‘straws’’ that would address the above problems by looking into the details of multiplicity distributions of neutrons originating from a fissioning source. This paper describes the straw neutron multiplicity counter (NMC) and assesses the performance with those of a commercially available fission meter. The prototype straw neutron detector provides a large-area, efficient, lightweight, more granular (than fission meter) neutron-responsive detection surface (to facilitate imaging) to enhance the ease of application of fission meters. Presented here are the results of preliminary investigations, modeling, and engineering considerations leading to the construction of this prototype. This design is capable of multiplicity and Feynman variance measurements. This prototype may lead to a near-term solution to the crisis that has arisen from the global scarcity of ³He by offering a viable alternative to fission meters. This paper describes the work performed during a 2-year site-directed research and development (SDRD) project that incorporated straw detectors for neutron multiplicity counting. The NMC is a two-panel detector system. We used ¹⁰B (in the form of enriched boron carbide: ¹⁰B₄C) for neutron detection instead of ³He. In the first year, the project worked with a panel of straw neutron detectors, investigated its characteristics, and

6LiF sandwich type detectors for low dose individual monitoring in mixed neutron-photon fields

International Nuclear Information System (INIS)

Olko, P.; Budzanowski, M.; Bilski, P.; Burgkhardt, B.; Piesch, E.

1994-01-01

ICRP Publication 60 recommends the reduction of the annual dose limit for occupational exposure from 50 to 20 mSv and a doubling of the quality factor for medium energy neutrons. If occupational doses are evaluated every month (which is obligatory e.g. in Germany and in Poland), the individual neutron dosemeter will have to measure neutron doses in the range of 100 μSv. No commercially available, automatic individual dosimetry monitoring system exists that fulfils this requirement. Some of the parameters which influence the evaluation of the neutron dose from readings of TL dosemeters have been studied in order to decrease the variance of the measured neutron signal. In mixed neutron-photon fields, clear separation of the neutron component from the total reading depends also on the uncertainty of the gamma dose measurements. While the thermal albedo neutrons are absorbed mostly at the surface of the 6 LiF detector, the reduction of the detector thickness results in a decrease of its photon sensitivity, while its neutron sensitivity is almost principally maintained. As a consequence, the uncertainty of gamma dose contributes with lower weight to the variance of the evaluated neutron signal. First tests of an optimised 200 μm thick sandwich detector and 0.9 mm thick standard LiF chips were made at low neutron and photon dose ranges using different readers, in order to determine the uncertainty versus dose for different neutron-photon combinations. The conditions under which the new sandwich type detectors may improve albedo neutron dosimetry are demonstrated. (Author)

New neutron detector based on micromegas technology for ADS projects

International Nuclear Information System (INIS)

Andriamonje, Samuel; Andriamonje, Gregory; Aune, Stephan; Ban, Gilles; Breaud, Stephane; Blandin, Christophe; Ferrer, Esther; Geslot, Benoit; Giganon, Arnaud; Giomataris, Ioannis; Jammes, Christian; Kadi, Yacine; Laborie, Philippe; Lecolley, Jean Francois; Pancin, Julien; Riallot, Marc; Rosa, Roberto; Sarchiapone, Lucia; Steckmeyer, Jean Claude; Tillier, Joel

2006-01-01

A new neutron detector based on Micromegas technology has been developed for the measurement of the simulated neutron spectrum in the ADS project. After the presentation of simulated neutron spectra obtained in the interaction of 140 MeV protons with the spallation target inside the TRIGA core, a full description of the new detector configuration is given. The advantage of this detector compared to conventional neutron flux detectors and the results obtained with the first prototype at the CELINA 14 MeV neutron source facility at CEA-Cadarache are presented. The future developments of operational Piccolo-Micromegas for fast neutron reactors are also described

New neutron detector based on micromegas technology for ADS projects

Energy Technology Data Exchange (ETDEWEB)

Andriamonje, Samuel [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France)]. E-mail: sandriamonje@cea.fr; Andriamonje, Gregory [IXL-Universite Bordeaux 1-BAT. A31-351 cours de la Liberation-F-33405 Talence Cedex (France); Aune, Stephan [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Ban, Gilles [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Breaud, Stephane [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Blandin, Christophe [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Ferrer, Esther [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Geslot, Benoit [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Giganon, Arnaud [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Giomataris, Ioannis [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Jammes, Christian [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Kadi, Yacine [CERN CH 1211 Geneva (Switzerland); Laborie, Philippe [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Lecolley, Jean Francois [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Pancin, Julien [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Riallot, Marc [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Rosa, Roberto [ENEA-Casaccia, Via Anguillarese, 00060 Rome (Italy); Sarchiapone, Lucia [CERN CH 1211 Geneva (Switzerland); Steckmeyer, Jean Claude [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Tillier, Joel [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France)

2006-06-23

A new neutron detector based on Micromegas technology has been developed for the measurement of the simulated neutron spectrum in the ADS project. After the presentation of simulated neutron spectra obtained in the interaction of 140 MeV protons with the spallation target inside the TRIGA core, a full description of the new detector configuration is given. The advantage of this detector compared to conventional neutron flux detectors and the results obtained with the first prototype at the CELINA 14 MeV neutron source facility at CEA-Cadarache are presented. The future developments of operational Piccolo-Micromegas for fast neutron reactors are also described.

Fast neutron activating detectors for pulsed flow measurements

International Nuclear Information System (INIS)

Dyatlov, V.D.; Kunaev, G.T.; Popytaev, A.N.; Cheremukhov, B.V.

1979-01-01

The requirements to the activation detectors of the pulsed flows of the fast neutrons are considered; the criteria of optimum measurement time, geometrical moderator sizes and radioactive detector element properties have been obtained. On their analysis parameter selection has been carried out. The neutron detector to register the short pulses has been designed and calibrated. The ways of further increase of sensitivity and efficiency of such detectors are discussed

Kalman filtering of self-powered neutron detectors

International Nuclear Information System (INIS)

Kantrowitz, M.L.

1992-01-01

Pressurized water reactors employ a wide variety of in-core detectors to determine the neutronic behavior within the core. Among the detectors used are rhodium and vanadium self-powered detectors (SPDs), which are very accurate, but respond slowly to changes in neutron flux. This paper describes a new dynamic compensation algorithm, based on Kalman filtering, which converts delayed-responding rhodium and vanadium SPDs into prompt-responding detectors by reconstructing the dynamic flux signal sensed by the detectors from the prompt and delayed components. This conversion offers the possibility of utilizing current fixed in-core detector systems based on these delayed-responding detectors for core control and/or core protection functions without the need for fixed in-core detectors which are prompt-responding. As a result, the capabilities of current fixed in-core detector systems could be expanded significantly without a major hardware investment

Orion, a high efficiency 4π neutron detector

International Nuclear Information System (INIS)

Crema, E.; Piasecki, E.; Wang, X.M.; Doubre, H.; Galin, J.; Guerreau, D.; Pouthas, J.; Saint-Laurent, F.

1990-01-01

In intermediate energy heavy ion collisions the multiplicity of emitted neutrons is strongly connected to energy dissipation and to impact parameter. We present the 4π detector ORION, a high efficiency liquid scintillator detector which permits to get information on the multiplicity of neutrons measured event-wise and on the spatial distribution of these neutrons [fr

nGEM fast neutron detectors for beam diagnostics

International Nuclear Information System (INIS)

Croci, G.; Claps, G.; Cavenago, M.; Dalla Palma, M.; Grosso, G.; Murtas, F.; Pasqualotto, R.; Perelli Cippo, E.; Pietropaolo, A.; Rebai, M.; Tardocchi, M.; Tollin, M.; Gorini, G.

2013-01-01

Fast neutron detectors with a sub-millimetric space resolution are required in order to qualify neutron beams in applications related to magnetically-controlled nuclear fusion plasmas and to spallation sources. A nGEM detector has been developed for the CNESM diagnostic system of the SPIDER NBI prototype for ITER and as beam monitor for fast neutrons lines at spallation sources. The nGEM is a triple GEM gaseous detector equipped with polypropylene and polyethylene layers used to convert fast neutrons into recoil protons through the elastic scattering process. This paper describes the results obtained by testing a nGEM detector at the ISIS spallation source on the VESUVIO beam line. Beam profiles (σ x =14.35 mm, σ y =15.75 mm), nGEM counting efficiency (around 10 -4 for 3 MeV n <15 MeV), detector stability (≈4.5%) and the effect of filtering the beam with different type of materials were successfully measured. The x beam profile was compared to the one measured by a single crystal diamond detector. Finally, the efficiency of the detector was simulated exploiting the GEANT4 tool

Improving neutron dosimetry using bubble detector technology

International Nuclear Information System (INIS)

Buckner, M.A.

1993-02-01

Providing accurate neutron dosimetry for a variety of neutron energy spectra is a formidable task for any dosimetry system. Unless something is known about the neutron spectrum prior to processing the dosimeter, the calculated dose may vary greatly from that actually encountered; that is until now. The entrance of bubble detector technology into the field of neutron dosimetry has eliminated the necessity of having an a priori knowledge of the neutron energy spectra. Recently, a new approach in measuring personnel neutron dose equivalent was developed at Oak Ridge National Laboratory. By using bubble detectors in combination with current thermoluminescent dosimeters (TLDs) as a Combination Personnel Neutron Dosimeter (CPND), not only is it possible to provide accurate dose equivalent results, but a simple four-interval neutron energy spectrum is obtained as well. The components of the CPND are a Harshaw albedo TLD and two bubble detectors with theoretical energy thresholds of 100 key and 1500 keV. Presented are (1) a synoptic history surrounding emergence of bubble detector technology, (2) a brief overview of the current theory on mechanisms of interaction, (3) the data and analysis process involved in refining the response functions, (4) performance evaluation of the original CPND and a reevaluation of the same data under the modified method, (5) the procedure used to determine the reference values of component fluence and dose equivalent for field assessment, (6) analysis of the after-modification results, (7) a critique of some currently held assumptions, offering some alternative explanations, and (8) thoughts concerning potential applications and directions for future research

Computed neutron response of spherical moderator-detector systems for radiation protection monitoring

International Nuclear Information System (INIS)

Dhairyawan, M.P.

1979-01-01

Neutrons of energies below 500 keV are important from the point of view of radiation protection of personnel working around reactors. However, as no neutron sources are available at lower energies, no measured values of neutron energy response are available between thermal and 0.5 MeV (but for Sb-Be source at 24 keV). The response functions in this range are, therefore, arrived at theoretically. After giving a comprehensive review of the work done in the field of response of moderated neutron detectors, a Monte Carlo method developed for this purpose is described and used to calculate energy response functions of the two spherical moderator-detector systems, namely, one using a central BF 3 counter and the other using 6 LiI(Eu) scintillator of 0.490 dia crystal. The polythene sphere diameter ranged from 2'' to 12''. The results obtained follow the trend predicted by other calculations and experiments, but are a definite improvement over them, because the most recent data on cross sections and angular distribution are used and the opacity of the detector i.e. the presence and size of the detector within the moderator is taken into account in the present calculations. The reasons for the discrepancies in the present results and those obtained earlier by other methods are discussed. The response of the Leake counter arrived at by the present method agrees very well with experimental calibration. (M.G.B.)

Two-dimensional microstrip detector for neutrons

Energy Technology Data Exchange (ETDEWEB)

Oed, A [Institut Max von Laue - Paul Langevin (ILL), 38 - Grenoble (France)

1997-04-01

Because of their robust design, gas microstrip detectors, which were developed at ILL, can be assembled relatively quickly, provided the prefabricated components are available. At the beginning of 1996, orders were received for the construction of three two-dimensional neutron detectors. These detectors have been completed. The detectors are outlined below. (author). 2 refs.

Spent Nuclear Fuel Cask and Storage Monitoring with {sup 4}He Scintillation Fast Neutron Detectors

Energy Technology Data Exchange (ETDEWEB)

Chung, Hee jun; Kelley, Ryan P; Jordan, Kelly A [Univ. of Florida, Florida (United States); Lee, Wanno [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chung, Yong Hyun [Yonsei Univ., Wonju (Korea, Republic of)

2014-10-15

With this increasing quantity of spent nuclear fuel being stored at nuclear plants across S. Korea, the demand exists for building a long-term disposal facility. However, the Korean government first requires a detailed plan for the monitoring and certification of spent fuel. Several techniques have been developed and applied for the purpose of spent fuel monitoring, including the digital Cerenkov viewing device (DCVD), spent fuel attribute tester (SFAT), and FORK detector. Conventional gamma measurement methods, however, suffer from a lack of nuclear data and interfering background radiation. To date, the primary method of neutron detection for spent fuel monitoring has been through the use of thermal neutron detectors such as {sup 3}He and BF{sub 3} proportional counters. Unfolding the neutron spectrum becomes extremely complicated. In an attempt to overcome these difficulties, a new fast neutron measurement system is currently being developed at the University of Florida. This system is based on the {sup 4}He scintillation detector invented by Arktis Radiation Detectors Ltd. These detectors are a relatively new technological development and take advantage of the high {sup 4}He cross-section for elastic scattering at fast neutron energies, particularly the resonance around 1 MeV. This novel {sup 4}He scintillation neutron detector is characterized by its low electron density, leading to excellent gamma rejection. This detector also has a fast response time on the order of nanoseconds and most importantly, preserves some neutron energy information since no moderator is required. Additionally, these detectors rely on naturally abundant {sup 4}He as the fill gas. This study proposes a new technique using the neutron spectroscopy features of {sup 4}He scintillation detectors to maintain accountability of spent fuel in storage. This research will support spent fuel safeguards and the detection of fissile material, in order to minimize the risk of nuclear proliferation

Evaluation of the Neutron Detector Response for Cosmic Ray Energy Spectrum by Monte Carlo Transport Simulation

International Nuclear Information System (INIS)

Pazianotto, Mauricio T.; Carlson, Brett V.; Federico, Claudio A.; Gonzalez, Odair L.

2011-01-01

Neutrons generated by the interaction of cosmic rays with the atmosphere make an important contribution to the dose accumulated in electronic circuits and aircraft crew members at flight altitude. High-energy neutrons are produced in spallation reactions and intranuclear cascade processes by primary cosmic-ray particle interactions with atoms in the atmosphere. These neutrons can produce secondary neutrons and also undergo a moderation process due to atmosphere interactions, resulting in a wider energy spectrum, ranging from thermal energies (0.025 eV) to energies of several hundreds of MeV. The Long-Counter (LC) detector is a widely used neutron detector designed to measure the directional flux of neutrons with about constant response over a wide energy range (thermal to 20 MeV). ). Its calibration process and the determination of its energy response for the wide-energy of cosmic ray induced neutron spectrum is a very difficult process due to the lack of installations with these capabilities. The goal of this study is to assess the behavior of the response of a Long Counter using the Monte Carlo (MC) computational code MCNPX (Monte Carlo N-Particle eXtended). The dependence of the Long Counter response on the angle of incidence, as well as on the neutron energy, will be carefully investigated, compared with the experimental data previously obtained with 241 Am-Be and 252 Cf neutron sources and extended to the neutron spectrum produced by cosmic rays. (Author)

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.

Neutron detector for detecting rare events of spontaneous fission

International Nuclear Information System (INIS)

Ter-Akop'yan, G.M.; Popeko, A.G.; Sokol, E.A.; Chelnokov, L.P.; Smirnov, V.I.; Gorshkov, V.A.

1981-01-01

The neutron detector for registering rare events of spontaneous fission by detecting multiple neutron emission is described. The detector represents a block of plexiglas of 550 mm diameter and 700 mm height in the centre of which there is a through 160 mm diameter channel for the sample under investigation. The detector comprises 56 3 He filled counters (up to 7 atm pressure) with 1% CO 2 addition. The counters have a 500 mm length and a 32 mm diameter. The sampling of fission events is realized by an electron system which allows determining the number of detected neutrons, numbers of operated counters, signal amplitude and time for fission event detecting. A block diagram of a neutron detector electron system is presented and its operation principle is considered. For protection against cosmic radiation the detector is surronded by a system of plastic scintillators and placed behind the concrete shield of 6 m thickness. The results of measurements of background radiation are given. It has been found that the background radiation of single neutron constitutes about 150 counts per hour, the detecting efficiency of single neutron equals 0.483 +- 0.005, for a 10l detector sensitive volume. By means of the detector described the parameters of multiplicity distribution of prompt neutrons for 256 Fm spontaneous fission are measured. The average multiplicity equals 3.59+-0.06 the dispersion being 2.30+-0.65

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.

Detector Response to Neutrons Slowed Down in Media Containing Cadmium

Energy Technology Data Exchange (ETDEWEB)

Broda, E.

1943-07-01

This report was written by E. Broda, H. Hereward and L. Kowarski at the Cavendish Laboratory (Cambridge) in September 1943 and is about the detector response to neutrons slowed down in media containing cadmium. The following measurement description and the corresponding results can be found in this report: B, Mn, In, I, Dy and Ir detectors were activated, with and without a Cd shield, near the source in a vessel containing 7 litres of water or solutions of CdSO{sub 4} ranging between 0.1 and 2.8 mols per litre. Numerical data on observed activities are discussed in two different ways and the following conclusions can be drawn: The capture cross-section of dysprosium decreases quicker than 1/v and this discrepancy becomes noticeable well within the limits of the C-group. This imposes obvious limitations on the use of Dy as a detector of thermal neutrons. Cadmium differences of manganese seem to be a reliable 1/v detector for the whole C-group. Indium and iridium show definite signs of an increase of vσ in the upper regions of the C-group. Deviations shown by iodine are due to the imperfections of the technique rather than to a definite departure from the 1/v law. (nowak)

Characterization of hybrid self-powered neutron detector under neutron irradiation

CERN Document Server

Nakamichi, M; Yamamura, C; Nakazawa, M; Kawamura, H

2000-01-01

To evaluate the irradiation behaviour of a blanket mock-up on in-pile functional test, it is necessary to measure the neutron flux change in the in-pile mock-up by a neutron detector, such as the self-powered neutron detector (SPND). With its small-sized emitter, which has high sensitivity and fast response time, SPND is an indispensable tool in order to measure the local neutron flux change. In the case of an in-pile functional test, it is necessary that response time is less than 1s and ratio of SPND output current is more than 0.3 of output current of SPND with Rh emitter. Therefore, a hybrid SPND with high sensitivity and fast response time was developed. This hybrid SPND used a hybrid emitter, i.e. Co cladded Pt-13%Rh.

Thermal stability of dyed tracks and electrochemical etching sensitivity of some polymeric detectors

International Nuclear Information System (INIS)

Monnin, M.; Gourcy, J.; Somogyi, G.; Dajko, D.

1980-01-01

Recent results on the mechanism of the formation of tracks obtained by the dyed tracks technique are given and the thermal annealing of the detectors is used to demonstrate their ability to retain tracks under more severe conditions than by the etching technique. Electrochemical etching of polycarbonate and polyethylene terephthalate detectors is investigated both from the background and sensitivity points of view. The polyethylene terephthalate detector is shown to be well suited for low neutron flux measurements. (author)

He-4 fast neutron detectors in nuclear security applications

International Nuclear Information System (INIS)

Murer, D. E.

2014-01-01

This work presents studies of "4He fast neutron detectors for nuclear security applications. Such devices are high pressure gas scintillation detectors, sensitive to neutrons in the energy range of fission sources. First, an introduction to the scope of the intended application is given. This is followed by a description of all components relevant to the operation of the detector. The next chapter presents studies of various characteristics of the neutron detector, among them properties of its scintillation response, differences between neutron and gamma interactions and effects of the light collection process. The results of the detector characterization are used to develop neutron gamma discrimination methods. These methods are put to the test using measurements with a high gamma flux, and the results are compared to performance requirements of Radiation Portal Monitors. Background neutron measurements are presented next. Measured neutron rates are compared to values published in scientific literature. The fluctuation of the background count rate was studied, and the contribution of muons evaluated. Two applications of the detectors in the field of nuclear security are discussed in the last two chapters. The first one is a novel method to measure the plutonium mass in a container filled with Mixed Oxide Fuel. The last chapter presents the development of a Radiation Portal Monitor which, in addition to neutron and gamma counting, exploits time correlation to detect threats such as plutonium and "6"0Co. (author)

He-4 fast neutron detectors in nuclear security applications

Energy Technology Data Exchange (ETDEWEB)

Murer, D. E.

2014-07-01

This work presents studies of {sup 4}He fast neutron detectors for nuclear security applications. Such devices are high pressure gas scintillation detectors, sensitive to neutrons in the energy range of fission sources. First, an introduction to the scope of the intended application is given. This is followed by a description of all components relevant to the operation of the detector. The next chapter presents studies of various characteristics of the neutron detector, among them properties of its scintillation response, differences between neutron and gamma interactions and effects of the light collection process. The results of the detector characterization are used to develop neutron gamma discrimination methods. These methods are put to the test using measurements with a high gamma flux, and the results are compared to performance requirements of Radiation Portal Monitors. Background neutron measurements are presented next. Measured neutron rates are compared to values published in scientific literature. The fluctuation of the background count rate was studied, and the contribution of muons evaluated. Two applications of the detectors in the field of nuclear security are discussed in the last two chapters. The first one is a novel method to measure the plutonium mass in a container filled with Mixed Oxide Fuel. The last chapter presents the development of a Radiation Portal Monitor which, in addition to neutron and gamma counting, exploits time correlation to detect threats such as plutonium and {sup 60}Co. (author)

High-Sensitivity Fast Neutron Detector KNK-2-8M

Science.gov (United States)

Koshelev, A. S.; Dovbysh, L. Ye.; Ovchinnikov, M. A.; Pikulina, G. N.; Drozdov, Yu. M.; Chuklyaev, S. V.; Pepyolyshev, Yu. N.

2017-12-01

The design of the fast neutron detector KNK-2-8M is outlined. The results of he detector study in the pulse counting mode with pulses from 238U nuclei fission in the radiator of the neutron-sensitive section and in the current mode with separation of functional section currents are presented. The possibilities of determination of the effective number of 238U nuclei in the radiator of the neutron-sensitive section are considered. The diagnostic capabilities of the detector in the counting mode are demonstrated, as exemplified by the analysis of reference data on characteristics of neutron fields in the BR-1 reactor hall. The diagnostic capabilities of the detector in the current mode are demonstrated, as exemplified by the results of measurements of 238U fission intensity in the power startup of the BR-K1 reactor in the fission pulse generation mode with delayed neutrons and the detector placed in the reactor cavity in conditions of large-scale variation of the reactor radiation fields.

Selective Filtration of Gadolinium Trichloride for Use in Neutron Detection in Large Water Cherenkov Detectors

International Nuclear Information System (INIS)

Vagins, Mark R.

2013-01-01

Super-??Kamiokande Water Cherenkov detectors have been used for many years as inexpensive, effective detectors for neutrino interactions and nucleon decay searches. While many important measurements have been made with these detectors a major drawback has been their inability to detect the absorption of thermal neutrons. We believe an inexpensive, effective technique could be developed to overcome this situation via the addition to water of a solute with a large neutron cross section and energetic gamma daughters which would make neutrons detectable. Gadolinium seems an excellent candidate especially since in recent years it has become very inexpensive, now less than $8 per kilogram in the form of commercially-available gadolinium trichloride, GdCl 3 . This non-toxic, non-reactive substance is highly soluble in water. Neutron capture on gadolinium yields a gamma cascade which would be easily seen in detectors like Super-Kamiokande. We have been investigating the use of GdCl 3 as a possible upgrade for the Super-Kamiokande detector with a view toward improving its performance as a detector for atmospheric neutrinos, supernova neutrinos, wrong-sign solar neutrinos, reactor neutrinos, proton decay, and also as a target for the coming T2K long-baseline neutrino experiment. This focused study of selective water filtration and GdCl 3 extraction techniques, conducted at UC Irvine, followed up on highly promising benchtop-scale and kiloton-scale work previously carried out with the assistance of 2003 and 2005 Advanced Detector Research Program grants

Self-Powered Neutron and Gamma Detectors for In-Core Measurements

International Nuclear Information System (INIS)

Strindehag, O.

1971-11-01

The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

Self-Powered Neutron and Gamma Detectors for In-Core Measurements

Energy Technology Data Exchange (ETDEWEB)

Strindehag, O

1971-11-15

The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

New thermal neutron calibration channel at LNMRI/IRD

Energy Technology Data Exchange (ETDEWEB)

Astuto, A.; Lopes, R.T., E-mail: achillesbr@gmail.com [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ/LNMRI), Rio de Janeiro, RJ (Brazil). Lab. Nacional de Metrologia das Radiacoes Ionizantes

2015-07-01

A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four {sup 241}Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

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.

Neutron spectrometry by diamond detector for nuclear radiation

International Nuclear Information System (INIS)

Kozlov, S.F.; Konorova, E.A.; Barinov, A.L.; Jarkov, V.P.

1975-01-01

Experiments on fast neutron spectrometry using the nuclear radiation diamond detector inside a horizontal channel of a water-cooled and water-moderated reactor are described. It is shown that the diamond detector enables neutron spectra to be measured within the energy range of 0.3 to 10 MeV against reactor gamma-radiation background and has radiation resistance higher than that of conventional semiconductor detectors. (U.S.)

Characterization of hybrid self-powered neutron detector under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Nakamichi, M. E-mail: masaru@oarai.jaeri.go.jp; Nagao, Y.; Yamamura, C.; Nakazawa, M.; Kawamura, H

2000-11-01

To evaluate the irradiation behaviour of a blanket mock-up on in-pile functional test, it is necessary to measure the neutron flux change in the in-pile mock-up by a neutron detector, such as the self-powered neutron detector (SPND). With its small-sized emitter, which has high sensitivity and fast response time, SPND is an indispensable tool in order to measure the local neutron flux change. In the case of an in-pile functional test, it is necessary that response time is less than 1s and ratio of SPND output current is more than 0.3 of output current of SPND with Rh emitter. Therefore, a hybrid SPND with high sensitivity and fast response time was developed. This hybrid SPND used a hybrid emitter, i.e. Co cladded Pt-13%R000.

Assessment of a silicon detector for pulsed neutron scattering experiments

International Nuclear Information System (INIS)

Tardocchi, M.; Arnaboldi, C.; Gorini, G.; Imberti, S.; Pessina, G.; Previtali, E.; Andreani, C.; Pietropaolo, A.; Senesi, R.

2004-01-01

Resonance detectors (RD) are being developed for neutron spectroscopy in the epithermal energy region at spallation neutron sources. Different choices of converter foils and gamma spectrometers are being compared as part of an optimization and selection process within the TECHNI project. This paper reports on the design of a silicon detector system and some preliminary tests on the VESUVIO spectrometer. The detector has a good efficiency in the X-ray energy range, where two intense photon peaks (at 12 and 48 keV) are expected to be emitted following neutron capture in a uranium converter foil. The detector energy resolution has been improved by nitrogen vapor cooling of the silicon chip and by careful design of the preamplifier electronics. Neutron time of flight spectra have been measured on VESUVIO when the converter foil is placed in the neutron beam. In that case, the detector response is dominated by a continuum due to Compton detection of gammas of higher energy. These results provide a basis for a critical assessment of the applicability of silicon detectors for RD measurements of epithermal neutrons

Neutron energy response measurement of scintillation detectors

International Nuclear Information System (INIS)

Yang Hongqiong; Peng Taiping; Yang Jianlun; Tang Zhengyuan; Yang Gaozhao; Li Linbo; Hu Mengchun; Wang Zhentong; Zhang Jianhua; Li Zhongbao; Wang Lizong

2004-01-01

Neutron sensitivities of detectors composed of plastic scintillator ST401, ST1422, ST1423 and phyotomultiplier tube in primary energy range of fission neutron are calibrated by direct current. The energy response curve of the detectors is obtained in this experiment. The experimental result has been compared with the theoretical calculation and they are in agreement within measuring uncertainty. (authors)

Technological advances in cosmogenic neutron detectors for measuring soil water content

Science.gov (United States)

Zreda, M. G.; Schrön, M.; Köhli, M.

2017-12-01

The cosmic-ray neutron probe is used for measuring area-average soil water content at the hectometer scale. Early work showed a simple exponential decrease with distance of the instrument's sensitivity and a footprint 300 m in radius. Recent research suggested a much higher sensitivity to local neutrons and reduced footprint. We show results confirming the high sensitivity to local neutrons, describe two ways to reduce local and increase far-field effects, and propose ways of measuring neutrons at different spatial scales. Measurements with moderated detectors across a 10-m-wide creek and a 2-m-wide water tank show a decrease by 30% and 20%, respectively, of neutron intensity over water compared to that over land nearby. These results mean that the detector is sensitive to meter-scale heterogeneities of water content. This sensitivity can be reduced by rising the detector or by shielding it from local neutrons. The effect of local water distributions on the measured neutron intensity decreases with height. In the water tank experiment it disappeared almost completely at the height of 2 m, leading to the conjecture that the height roughly equal to the horizontal scale of heterogeneity would eliminate the sensitivity. This may or may not be practical. Shielding the detector below by a hydrogenous material removes a substantial fraction of the local neutrons. The shielded detector has a reduced count rate, reduced sensitivity to local neutrons and increased sensitivity to neutrons farther afield, and a larger footprint. Such a detector could be preferable to the current cosmogenic-neutron probe under heterogeneous soil water conditions. The shielding experiments also inspired the development of a local-area neutron detector. It has hydrogenous neutron shields on all sides except the bottom, substantially blocking the neutrons coming from afar, while allowing the neutrons coming directly from below. Its footprint is equal to its physical dimension when the detector is

The investigation of fast neutron Threshold Activation Detectors (TAD)

International Nuclear Information System (INIS)

Gozani, T; King, M J; Stevenson, J

2012-01-01

The detection of fast neutrons is usually done by liquid hydrogenous organic scintillators, where the separation between the ever present gamma rays and neutrons is achieved by the pulse shape discrimination (PSD). In many practical situation the detection of fast neutrons has to be carried out while the intense source (be it neutrons, gamma rays or x-rays) that creates these neutrons, for example by the fission process, is present. This source, or ''flash'', usually blinds the neutron detectors and temporarily incapacitates them. By the time the detectors recover the prompt neutron signature does not exist. Thus to overcome the blinding background, one needs to search for processes whereby the desired signature, such as fission neutrons could in some way be measured long after the fission occurred and when the neutron detector is fully recovered from the overload. A new approach was proposed and demonstrated a good sensitivity for the detection of fast neutrons in adverse overload situations where normally it could not be done. A temporal separation of the fission event from the prompt neutrons detection is achieved via the activation process. The main idea, called Threshold Activation Detection (or detector)-TAD, is to find appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation, and then measure the radioactively decaying activation products (typically beta and γ-rays) well after the source pulse has ended. The activation material should possess certain properties: a suitable half-life; an energy threshold below which the numerous source neutrons will not activate it (e.g. about 3 MeV); easily detectable activation products and has a usable cross section for the selected reaction. Ideally the substance would be part of the scintillator. There are several good candidates for TAD. The first one we have selected is based on fluorine. One of the major advantages of this element is the fact that it is a major

The investigation of fast neutron Threshold Activation Detectors (TAD)

Science.gov (United States)

Gozani, T.; King, M. J.; Stevenson, J.

2012-02-01

The detection of fast neutrons is usually done by liquid hydrogenous organic scintillators, where the separation between the ever present gamma rays and neutrons is achieved by the pulse shape discrimination (PSD). In many practical situation the detection of fast neutrons has to be carried out while the intense source (be it neutrons, gamma rays or x-rays) that creates these neutrons, for example by the fission process, is present. This source, or ``flash'', usually blinds the neutron detectors and temporarily incapacitates them. By the time the detectors recover the prompt neutron signature does not exist. Thus to overcome the blinding background, one needs to search for processes whereby the desired signature, such as fission neutrons could in some way be measured long after the fission occurred and when the neutron detector is fully recovered from the overload. A new approach was proposed and demonstrated a good sensitivity for the detection of fast neutrons in adverse overload situations where normally it could not be done. A temporal separation of the fission event from the prompt neutrons detection is achieved via the activation process. The main idea, called Threshold Activation Detection (or detector)-TAD, is to find appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation, and then measure the radioactively decaying activation products (typically beta and γ-rays) well after the source pulse has ended. The activation material should possess certain properties: a suitable half-life; an energy threshold below which the numerous source neutrons will not activate it (e.g. about 3 MeV); easily detectable activation products and has a usable cross section for the selected reaction. Ideally the substance would be part of the scintillator. There are several good candidates for TAD. The first one we have selected is based on fluorine. One of the major advantages of this element is the fact that it is a major

Measurement of the diffusion length of thermal neutrons inside graphite

International Nuclear Information System (INIS)

Ertaud, A.; Beauge, R.; Fauquez, H.; De Laboulay, H.; Mercier, C.; Vautrey, L.

1948-11-01

The diffusion length of thermal neutrons inside a given industrial graphite is determined by measuring the neutron density inside a parallelepipedal piling up of graphite bricks (2.10 x 2.10 x 2.442 m). A 3.8 curies (Ra α → Be) source is placed inside the parallelepipedal block of graphite and thin manganese detectors are used. Corrections are added to the unweighted measurements to take into account the effects of the damping of supra-thermal neutrons in the measurement area. These corrections are experimentally deduced from the differential measurements made with a cadmium screen interposed between the source and the first plane of measurement. An error analysis completes the report. The diffusion length obtained is: L = 45.7 cm ± 0.3. The average density of the graphite used is 1.76 and the average apparent density of the piling up is 1.71. (J.S.)

Secondary standard neutron detector for measuring total reaction cross sections

International Nuclear Information System (INIS)

Sekharan, K.K.; Laumer, H.; Gabbard, F.

1975-01-01

A neutron detector has been constructed and calibrated for the accurate measurement of total neutron-production cross sections. The detector consists of a polyethylene sphere of 24'' diameter in which 8- 10 BF 3 counters have been installed radially. The relative efficiency of this detector has been determined for average neutron energies, from 30 keV to 1.5 MeV by counting neutrons from 7 Li(p,n) 7 Be. By adjusting the radial positions of the BF 3 counters in the polyethylene sphere the efficiency for neutron detection was made nearly constant for this energy range. Measurement of absolute efficiency for the same neutron energy range has been done by counting the neutrons from 51 V(p,n) 51 Cr and 57 Fe(p,n) 57 Co reactions and determining the absolute number of residual nuclei produced during the measurement of neutron yield. Details of absolute efficiency measurements and the use of the detector for measurement of total neutron yields from neutron producing reactions such as 23 Na(p,n) 23 Mg are given

Self-powered in-core neutron detector assembly with uniform perturbation characteristics

International Nuclear Information System (INIS)

Todt, W.H.; Playfoot, K.C.

1979-01-01

Disclosed is a self-powered in-core neutron detector assembly in which a plurality of longitudinally extending self-powered detectors have neutron responsive active portions spaced along a longitudinal path. A low neutron absorptive extension extends from the active portions of the spaced active portions of the detectors in symmetrical longitudinal relationship with the spaced active detector portions of each succeeding detector. The detector extension terminates with the detector assembly to provide a uniform perturbation characteristic over the entire assembly length

Calibration of thermal neutron detection compound BN-1 and CR-39 in the exposure room of Triga Reactor in Ljubljana

International Nuclear Information System (INIS)

Kristof, E.; Ilic, R.; Skvarc, J.; Dijanosic, R.

1994-01-01

Description of determination of thermal neutron fluences in the range from 1.E+02 to 1.E+12 cm -2 for calibration of the neutron sensitive compound consisting of the neutron converter BN-1 and charged particle detector CR-39 is given. The method employs two proportional BF3 detectors supplemented by a Ge(Li) gamma spectometer utilizing gold foils. The results of the measurements are also presented. (author)

MCNP-REN a Monte Carlo tool for neutron detector design

CERN Document Server

Abhold, M E

2002-01-01

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel w...

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