Study of the environmental neutron spectrum at Zacatecas city

International Nuclear Information System (INIS)

Vega C, H.R.

2003-01-01

The environmental neutron spectrum has been measured at Zacatecas City in Mexico. Neutron spectrum was unfolded from count rates obtained with a multisphere neutron spectrometer with a Li I(Eu) scintillator. With the spectrum information the ambient dose equivalent and the isotropic effective dose were calculated. A model based upon the geomagnetic latitude and the altitude above sea level, that allows to estimate the neutron fluence rate is proposed, the model results are compared with total neutron fluences measured at several locations worldwide. Environmental neutron spectrum shows peaks at 1 and 100 MeV as well as a relevant amount of low energy neutrons. The neutron fluence rate was 65 ± 3 cm -2 -h -1 , producing 13.7 ± 0.6 n Sv-h -1 due to ambient dose equivalent rate and an isotropic effective dose rate of 14.1 ± 0.6 n Sv-h -1 . Neutron fluence rates predicted with the model are in agreement with those reported in the literature. (Author)

Study of the environmental neutron spectrum at Zacatecas city

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R. [Universidad Autonoma de Zacatecas, Cuerpo Academico de Radiobiologia, A.P. 336, 98000 Zacatecas (Mexico)

2003-07-01

The environmental neutron spectrum has been measured at Zacatecas City in Mexico. Neutron spectrum was unfolded from count rates obtained with a multisphere neutron spectrometer with a Li I(Eu) scintillator. With the spectrum information the ambient dose equivalent and the isotropic effective dose were calculated. A model based upon the geomagnetic latitude and the altitude above sea level, that allows to estimate the neutron fluence rate is proposed, the model results are compared with total neutron fluences measured at several locations worldwide. Environmental neutron spectrum shows peaks at 1 and 100 MeV as well as a relevant amount of low energy neutrons. The neutron fluence rate was 65 {+-} 3 cm{sup -2}-h{sup -1}, producing 13.7 {+-} 0.6 n Sv-h{sup -1} due to ambient dose equivalent rate and an isotropic effective dose rate of 14.1 {+-} 0.6 n Sv-h{sup -1}. Neutron fluence rates predicted with the model are in agreement with those reported in the literature. (Author)

Uncertainties related to numerical methods for neutron spectra unfolding

International Nuclear Information System (INIS)

Glodic, S.; Ninkovic, M.; Adarougi, N.A.

1987-10-01

One of the often used techniques for neutron detection in radiation protection utilities is the Bonner multisphere spectrometer. Besides its advantages and universal applicability for evaluating integral parameters of neutron fields in health physics practices, the outstanding problems of the method are data analysis and the accuracy of the results. This paper briefly discusses some numerical problems related to neutron spectra unfolding, such as uncertainty of the response matrix as a source of error, and the possibility of real time data reduction using spectrometers. (author)

Evaluation of response matrix of a multisphere neutron spectrometer ...

Indian Academy of Sciences (India)

Abstract. Neutron energy responses of water sphere spectrometers (WSS) to 30 MeV have been calculated by means of Monte Carlo calculations, using the computer code MCNP4C with ENDF/. B-VI.0 neutron cross-section. The calculations have been performed for 3He detector (typical SP9) placed inside 2, 3, 5, 8, ...

Evaluation of room-scattered neutrons at the JNC Tokai neutron reference field

Energy Technology Data Exchange (ETDEWEB)

Yoshida, Tadayoshi; Tsujimura, Norio [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan). Tokai Works; Oyanagi, Katsumi [Japan Radiation Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

2002-09-01

Neutron reference fields for calibrating neutron-measuring devices in JNC Tokai Works are produced by using radionuclide neutron sources, {sup 241}Am-Be and {sup 252}Cf sources. The reference field for calibration includes scattered neutrons from the material surrounding sources, wall, floor and ceiling of the irradiation room. It is, therefore, necessary to evaluate the scattered neutrons contribution and their energy spectra at reference points. Spectral measurements were performed with a set of Bonner multi-sphere spectrometers and the reference fields were characterized in terms of spectral composition and the fractions of room-scattered neutrons. In addition, two techniques stated in ISO 10647, the shadow-cone method and the polynomial fit method, for correcting the contributions from the room-scattered neutrons to the readings of neutron survey instruments were compared. It was found that the two methods gave an equivalent result within a deviation of 3.3% at a source-to-detector distance from 50cm to 500cm. (author)

Evaluation of room-scattered neutrons at the JNC Tokai neutron reference field

International Nuclear Information System (INIS)

Yoshida, Tadayoshi; Tsujimura, Norio

2002-01-01

Neutron reference fields for calibrating neutron-measuring devices in JNC Tokai Works are produced by using radionuclide neutron sources, 241 Am-Be and 252 Cf sources. The reference field for calibration includes scattered neutrons from the material surrounding sources, wall, floor and ceiling of the irradiation room. It is, therefore, necessary to evaluate the scattered neutrons contribution and their energy spectra at reference points. Spectral measurements were performed with a set of Bonner multi-sphere spectrometers and the reference fields were characterized in terms of spectral composition and the fractions of room-scattered neutrons. In addition, two techniques stated in ISO 10647, the shadow-cone method and the polynomial fit method, for correcting the contributions from the room-scattered neutrons to the readings of neutron survey instruments were compared. It was found that the two methods gave an equivalent result within a deviation of 3.3% at a source-to-detector distance from 50cm to 500cm. (author)

Measurements of neutron spectrum from uranium converter

International Nuclear Information System (INIS)

Ninkovic, M.; Sotic, O.; Marinkovic, S.

1978-01-01

The procedure for determination of energetic distribution of neutrons by the multisphere technique is given. The theoretical basis and features of the method are explained. The spectral distribution of neutrons emerging from the neutron converter constructed at the bare reactor assembly RB, has been determined applying the existing computer programme and literature data for the energetic dependence functions of spheres of various diameters. The obtained spectral distribution has a specific maximum in the domain of fast neutrons, justifying thus the reacton for the construction of the converter. The neutron spectrum data obtained and given in this report are very important for the use of the converter in neutron dosimetry and radiation protection, as well as in the radiobiology, shielding, reactor physics etc. (author)

Correlation of Neutron Data Taken at Commercial Nuclear Sites

Energy Technology Data Exchange (ETDEWEB)

Rathbun, L. A.; Endres, G. W.R.

1983-10-01

In this report, data from neutron measurement and dosimetry studies performed by Pacific Northwest Laboratory, the Environmental Measurements Laboratory, and Rensselaer Polytechnic Institute are examined and compared. The purpose of this data correlation effort is to determine whether useful relationships exist between the actual neutron dose equivalent in a typical commercial nuclear power reactor and various measurement parameters, such as ratios of the response of 9-in. to 3-in. spheres, neutron/gamma ratios, albedo dosimeter response and neutron spectrometer readings. In most neutron radiation fields found in the reactors visited, the response of albedo dosimeters can be brought into reasonable agreement with dose equivalents measured with multispheres, tissue equivalent proportional counters (TEPCs) or remmeters. Because the responses of the remmeters, like the responses of albedo dosimeters, are energy dependent, it is preferable to correct the responses of the albedo dosimeters to agree with dose equivalents measured with either TEPCs or multispheres. If one of these laboratory systems has been used to measure neutron dose equivalents at a specific pressurized water reactor, a calculated average albedo dosimeter correction factor can be used for most locations at that reactor. However, if the measured 9-in. to 3-in. remmeter ratio is greater than 0.20, it is advisable to use a plot of 9-in. to 3-in. remmeter ratios versus albedo dosimeter correction factors to obtain an albedo dosimeter correction factor. Because 9-in. to 3-in. remmeter ratios at boiling water reactors are typically greater than 0.20, the latter approach applies to this type of reactor.

Evaluation of neutron dosimetry techniques for well-logging operations

International Nuclear Information System (INIS)

Cummings, F.M.; Haggard, D.L.; Endres, G.W.R.

1985-07-01

Neutron dose and energy spectral measurements from 241 AmBe and a 14 MeV neutron generator were performed at a well-logging laboratory. The measurement technique included the tissue equivalent proportional counter, multisphere, two types of remmeters and five types of personnel neutron dosimeters. Several source configurations were used to attempt to relate data to field situations. The results of the measurements indicated that the thermoluminescent albedo dosimeter was the most appropriate personnel neutron dosimeter, and that the most appropriate calibration source would be the source normally employed in the field with the calibration source being used in the unmoderated configuration. 7 refs., 35 figs., 14 tabs

NEMUS--the PTB Neutron Multisphere Spectrometer Bonner spheres and more

CERN Document Server

Wiegel, B

2002-01-01

The original Bonner sphere spectrometer as it is used and characterized by PTB consists of 12 polyethylene spheres with diameters from 7.62 cm (3'') to 45.72 cm (18'') and a sup 3 He-filled spherical proportional counter used as a central thermal-neutron-sensitive detector and as a bare or cadmium-shielded bare detector. In this paper, a set of four new spheres made of polyethylene with copper or lead inlets is introduced. All spheres are less than 18 kg in mass and their responses to high energy neutrons increase with energy as a result of the increasing (n,xn) cross-sections of copper and lead. The fluence response matrix was calculated up to 10 GeV using an extended neutron cross-section library (LA150) and the MCNP(X) Monte Carlo code. Calibration measurements with neutron energies up to 60 MeV were used to compare the calculated response functions to measured values. For measurements outside the laboratory, a miniaturized, battery-powered electronic set-up was developed. This system with the additional, ...

FRUIT: An operational tool for multisphere neutron spectrometry in workplaces

International Nuclear Information System (INIS)

Bedogni, Roberto; Domingo, Carles; Esposito, Adolfo; Fernandez, Francisco

2007-01-01

FRUIT (Frascati Unfolding Interactive Tool) is an unfolding code for Bonner sphere spectrometers (BSS) developed, under the Labview environment, at the INFN-Frascati National Laboratory. It models a generic neutron spectrum as the superposition of up to four components (thermal, epithermal, fast and high energy), fully defined by up to seven positive parameters. Different physical models are available to unfold the sphere counts, covering the majority of the neutron spectra encountered in workplaces. The iterative algorithm uses Monte Carlo methods to vary the parameters and derive the final spectrum as limit of a succession of spectra fulfilling the established convergence criteria. Uncertainties on the final results are evaluated taking into consideration the different sources of uncertainty affecting the input data. Relevant features of FRUIT are (1) a high level of interactivity, allowing the user to follow the convergence process, (2) the possibility to modify the convergence tolerances during the run, allowing a rapid achievement of meaningful solutions and (3) the reduced dependence of the results from the initial hypothesis. This provides a useful instrument for spectrometric measurements in workplaces, where detailed a priori information is usually unavailable. This paper describes the characteristics of the code and presents the results of performance tests over a significant variety of reference and workplace neutron spectra ranging from thermal up to hundreds MeV neutrons

A diffusion-theoretical method to calculate the neutron flux distribution in multisphere configurations

International Nuclear Information System (INIS)

Schuerrer, F.

1980-01-01

For characterizing heterogene configurations of pebble-bed reactors the fine structure of the flux distribution as well as the determination of the macroscopic neutronphysical quantities are of interest. When calculating system parameters of Wigner-Seitz-cells the usual codes for neutron spectra calculation always neglect the modulation of the neutron flux by the influence of neighbouring spheres. To judge the error arising from that procedure it is necessary to determinate the flux distribution in the surrounding of a spherical fuel element. In the present paper an approximation method to calculate the flux distribution in the two-sphere model is developed. This method is based on the exactly solvable problem of the flux determination of a point source of neutrons in an infinite medium, which contains a spherical perturbation zone eccentric to the point source. An iteration method allows by superposing secondary fields and alternately satisfying the conditions of continuity on the surface of each of the two fuel elements to advance to continually improving approximations. (orig.) 891 RW/orig. 892 CKA [de

Altitude variation of cosmic-ray neutrons

International Nuclear Information System (INIS)

Nakamura, T.; Uwamino, Y.; Ohkubo, T.; Hara, A.

1987-01-01

The altitude variation of the cosmic-ray neutron energy spectrum and the dose equivalent rate was measured at an average geomagnetic latitude of 24 degrees N by using the high-efficiency multi-sphere neutron spectrometer and neutron dose-equivalent counter developed by the authors. The data were obtained from a 2-h flight over Japan on 27 February 1985. The neutron energy spectra measured at sea level and at altitudes of 4880 m and at 11,280 m were compared with the calculated spectra of O'Brien and with other experimental spectra, and they are in moderately good agreement with them. The dose equivalent rate increases according to a quadratic curve up to about 6000 m and then increases linearly between 6000 m and 11,280 m. The dependence of dose equivalent rates at sea level and at an altitude of 12,500 m on geomagnetic latitude also is given by referring to other experimental results

Neutron spectra produced by moderating an isotopic neutron source

International Nuclear Information System (INIS)

Carrillo Nunnez, Aureliano; Vega Carrillo, Hector Rene

2001-01-01

A Monte Carlo study has been carried out to determine the neutron spectra produced by an isotopic neutron source inserted in moderating media. Most devices used for radiation protection have a response strongly dependent on neutron energy. ISO recommends several neutron sources and monoenergetic neutron radiations, but actual working situations have broad spectral neutron distributions extending from thermal to MeV energies, for instance, near nuclear power plants, medical applications accelerators and cosmic neutrons. To improve the evaluation of the dosimetric quantities, is recommended to calibrate the radiation protection devices in neutron spectra which are nearly like those met in practice. In order to complete the range of neutron calibrating sources, it seems useful to develop several wide spectral distributions representative of typical spectra down to thermal energies. The aim of this investigation was to use an isotopic neutron source in different moderating media to reproduce some of the neutron fields found in practice. MCNP code has been used during calculations, in these a 239PuBe neutron source was inserted in H2O, D2O and polyethylene moderators. Moderators were modeled as spheres and cylinders of different sizes. In the case of cylindrical geometry the anisotropy of resulting neutron spectra was calculated from 0 to 2 . From neutron spectra dosimetric features were calculated. MCNP calculations were validated by measuring the neutron spectra of a 239PuBe neutron source inserted in a H2O cylindrical moderator. The measurements were carried out with a multisphere neutron spectrometer with a 6LiI(Eu) scintillator. From the measurements the neutron spectrum was unfolded using the BUNKIUT code and the UTA4 response matrix. Some of the moderators with the source produce a neutron spectrum close to spectra found in actual applications, then can be used during the calibration of radiation protection devices

Calibration and intercomparison of neutron moderation spectrometers

International Nuclear Information System (INIS)

Rimpler, A.; Hermanska, J.; Prouza, Z.

1989-01-01

Results have been reported of comparative measurements of neutron fields from bare PuBe and Cf sources using multisphere (Bonner) spectrometers. The experiments were carried out by the Institute of Biophysics and Nuclear Medicine at Charles University in Prague and the National Board for Atomic Safety and Radiation Protection in Berlin. Both sides agreed upon uniform measuring conditions and calibration factors thus rendering possible the comparability of the dosimetric parameters which have been determined and verified, respectively, to an accuracy of ± 10%. 20 refs., 10 tabs., 2 figs. (author)

Neutron dose and energy spectra measurements at Savannah River Plant

International Nuclear Information System (INIS)

Brackenbush, L.W.; Soldat, K.L.; Haggard, D.L.; Faust, L.G.; Tomeraasen, P.L.

1987-08-01

Because some workers have a high potential for significant neutron exposure, the Savannah River Plant (SRP) contracted with Pacific Northwest Laboratory (PNL) to verify the accuracy of neutron dosimetry at the plant. Energy spectrum and neutron dose measurements were made at the SRP calibrations laboratory and at several other locations. The energy spectra measurements were made using multisphere or Bonner sphere spectrometers, 3 He spectrometers, and NE-213 liquid scintillator spectrometers. Neutron dose equivalent determinations were made using these instruments and others specifically designed to determine dose equivalent, such as the tissue equivalent proportional counter (TEPC). Survey instruments, such as the Eberline PNR-4, and the thermoluminescent dosimeter (TLD)-albedo and track etch dosimeters (TEDs) were also used. The TEPC, subjectively judged to provide the most accurate estimation of true dose equivalent, was used as the reference for comparison with other devices. 29 refs., 43 figs., 13 tabs

Use of the helium-3 proportional counter for neutron spectrometry; Utilisation du compteur proportionnel a helium 3 pour la spectrometrie des neutrons

Energy Technology Data Exchange (ETDEWEB)

Vialettes, H; Le Thanh, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

Up to now, two methods have been mainly used for neutron spectrometry near nuclear installations: - photographic emulsion spectrometry - the so-called, 'multisphere' technique spectrometry. The first method, which is fairly difficult to apply, has a threshold energy of about 500 keV; this is a big disadvantage for an apparatus which has to be used for spectrometry around nuclear installations where the neutron radiation is very much degraded energetically. The second method does not suffer from this disadvantage but the results which it yields are only approximate. In order to extend the energy range of the neutron spectra studied with sufficient accuracy the use of a helium-3 proportional counter has been considered. This report presents the principles of operation of the helium-3 spectrometer, and the calculation methods which make it possible to take into account the two main effects tending to deform the spectra obtained: - energy absorption by the walls of the counter, - energy loss of the incident neutrons due to elastic collisions with helium-3 nuclei. As an example of the application, the shape of the neutron spectrum emitted by a polonium-lithium source is given; the results obtained are in excellent agreement with theoretical predictions. (authors) [French] Jusqu'ici deux methodes ont ete utilisees principalement pour la spectrometrie des neutrons autour des installations nucleaires: - la spectrometrie par emulsions photographiques - la spectrometrie par la technique dite des multispheres. La premiere methode, d'un emploi assez delicat presente un seuil en energie d'environ 500 keV qui est un obstacle serieux a la spectrometrie autour des installations nucleaires ou le rayonnement neutronique est tres degrade en energie. La deuxieme methode ne presente pas cet inconvenient mais les resultats qu'elle permet d'obtenir ne sont qu'approches. Pour etendre la gamme d'energie des spectres de neutrons etudies avec une precision suffisante, l'utilisation du

Site-specific calibration of the Hanford personnel neutron dosimeter

International Nuclear Information System (INIS)

Endres, A.W.; Brackenbush, L.W.; Baumgartner, W.V.; Rathbone, B.A.

1994-10-01

A new personnel dosimetry system, employing a standard Hanford thermoluminescent dosimeter (TLD) and a combination dosimeter with both CR-39 nuclear track and TLD-albedo elements, is being implemented at Hanford. Measurements were made in workplace environments in order to verify the accuracy of the system and establish site-specific factors to account for the differences in dosimeter response between the workplace and calibration laboratory. Neutron measurements were performed using sources at Hanford's Plutonium Finishing Plant under high-scatter conditions to calibrate the new neutron dosimeter design to site-specific neutron spectra. The dosimeter was also calibrated using bare and moderated 252 Cf sources under low-scatter conditions available in the Hanford Calibration Laboratory. Dose equivalent rates in the workplace were calculated from spectrometer measurements using tissue equivalent proportional counter (TEPC) and multisphere spectrometers. The accuracy of the spectrometers was verified by measurements on neutron sources with calibrations directly traceable to the National Institute of Standards and Technology (NIST)

Response matrix of a multisphere neutron spectrometer with an 3 He proportional counter

International Nuclear Information System (INIS)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A.

2005-01-01

The response matrix of a Bonner sphere spectrometer was calculated by use of the MCNP code. As thermal neutron counter, the spectrometer has a 3.2 cm-diameter 3 He-filled proportional counter which is located at the center of a set of polyethylene spheres. The response was calculated for 0, 3, 5, 6, 8, 10, 12, and 16 inches-diameter polyethylene spheres for neutrons whose energy goes from 10 -9 to 20 MeV. The response matrix was compared with a set of responses measured with several monoenergetic neutron sources. In this comparison the calculated matrix agrees with the experimental results. The matrix was also compared with the response matrix calculated for the PTB C spectrometer. Even though that calculation was carried out using a detailed model to describe the proportional counter; both matrices do agree, but small differences are observed in the bare case because of the difference in the model used during calculations. Other differences are in some spheres for 14.8 and 20 MeV neutrons, probably due to the differences in the cross sections used during both calculations. (Author) 28 refs., 1 tab., 6 figs

Study on neutron dosimetry in JNC Tokai Works

Energy Technology Data Exchange (ETDEWEB)

Tsujimura, Norio [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan). Tokai Works

2003-03-01

The author developed the neutron reference calibration fields using a {sup 252}Cf standard source surrounded with PMMA (polymethylmethacrylates) moderators at the Japan Nuclear Cycle Development Institute (JNC), Tokai Works. The moderators are concentric, annular cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the {sup 252}Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO{sub 2}-UO{sub 2} mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments. (author)

Characteristics of thermal neutron calibration fields using a graphite pile

International Nuclear Information System (INIS)

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

2005-03-01

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

Development of the neutron reference calibration field using a 252Cf standard source surrounded with PMMA moderators

International Nuclear Information System (INIS)

Yoshida, T.; Kanai, K.; Tsujimura, N.

2002-01-01

The authors developed the neutron reference calibration fields using a 252 Cf standard source surrounded with PMMA moderators at the Japan Nuclear Cycle Development (JNC), Tokai Works. The moderators are co-axial, hollow cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the 252 Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO 2 -UO 2 mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments

Spectrometry using the PTB neutron multisphere spectrometer (NEMUS) at flight altitudes and at ground level

CERN Document Server

Wiegel, B; Matzke, M; Schrewe, U J; Wittstock, J

2002-01-01

Bonner sphere measurements are presented for flights at altitudes of up to 12 km and geomagnetic latitudes between 26 deg.N and 86 deg.N and compared with results obtained by several survey meters. As an example of the natural neutron background near sea level, results from a recent longterm measurement campaign performed at the PTB site using an extended spectrometer are presented. The dependence of neutron fluence and ambient dose equivalent on the atmospheric pressure is demonstrated.

Measurement of the energy spectrum of cosmic-ray induced neutrons aboard an ER-2 high-altitude airplane

CERN Document Server

Goldhagen, P E; Kniss, T; Reginatto, M; Singleterry, R C; Van Steveninck, W; Wilson, J W

2002-01-01

Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from galactic cosmic radiation. Crews of future high-speed commercial aircraft flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the atmospheric ionizing radiation (AIR) project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on five flights of a NASA ER-2 high-altitude aircraft. The primary AIR instrument was a highly sensitive extended-energy multisphere neutron spectrometer with lead and steel shells placed within the moderators of two of its 14 detectors to enhance response at high energies. Detector responses were calculated for neutrons and charged hadrons at energies up to 100 GeV using MCNPX. Neutron spectra were unfolded from the measured count rates using the new MAXED code. We have measured the cosmic-ray neutron spectrum (t...

Use of the helium-3 proportional counter for neutron spectrometry

International Nuclear Information System (INIS)

Vialettes, H.; Le Thanh, P.

1967-01-01

Up to now, two methods have been mainly used for neutron spectrometry near nuclear installations: - photographic emulsion spectrometry - the so-called, 'multisphere' technique spectrometry. The first method, which is fairly difficult to apply, has a threshold energy of about 500 keV; this is a big disadvantage for an apparatus which has to be used for spectrometry around nuclear installations where the neutron radiation is very much degraded energetically. The second method does not suffer from this disadvantage but the results which it yields are only approximate. In order to extend the energy range of the neutron spectra studied with sufficient accuracy the use of a helium-3 proportional counter has been considered. This report presents the principles of operation of the helium-3 spectrometer, and the calculation methods which make it possible to take into account the two main effects tending to deform the spectra obtained: - energy absorption by the walls of the counter, - energy loss of the incident neutrons due to elastic collisions with helium-3 nuclei. As an example of the application, the shape of the neutron spectrum emitted by a polonium-lithium source is given; the results obtained are in excellent agreement with theoretical predictions. (authors) [fr

Development of the neutron reference calibration field using a {sup 252}Cf standard source surrounded with PMMA moderators

Energy Technology Data Exchange (ETDEWEB)

Yoshida, T.; Kanai, K.; Tsujimura, N. [Japan Nuclear Cycle Development Institute, Ibaraki-ken (Japan)

2002-07-01

The authors developed the neutron reference calibration fields using a {sup 252} Cf standard source surrounded with PMMA moderators at the Japan Nuclear Cycle Development (JNC), Tokai Works. The moderators are co-axial, hollow cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the {sup 252} Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO{sub 2}-UO{sub 2} mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments.

Neutron spectrometry with Bonner spheres for area monitoring in particle accelerators

International Nuclear Information System (INIS)

Bedogni, R.

2011-01-01

Selecting the instruments to determine the operational quantities in the neutron fields produced by particle accelerators involves a combination of aspects, which is peculiar to these environments: the energy distribution of the neutron field, the continuous or pulsed time structure of the beam, the presence of other radiations to which the neutron instruments could have significant response and the large variability in the dose rate, which can be observed when moving from areas near the beam line to free-access areas. The use of spectrometric techniques in support of traditional instruments is highly recommended to improve the accuracy of dosimetric evaluations. The multi-sphere or Bonner Sphere Spectrometer (BSS) is certainly the most used device, due to characteristics such as the wide energy range, large variety of active and passive detectors suited for different workplaces, good photon discrimination and the simple signal management. Disadvantages are the poor energy resolution, weight and need to sequentially irradiate the spheres, leading to usually long measurement sessions. Moreover, complex unfolding analyses are needed to obtain the neutron spectra. This work is an overview of the BSS for area monitoring in particle accelerators. (authors)

A systematic approach to personnel neutron monitoring

International Nuclear Information System (INIS)

Griffith, R.V.; Hankins, D.E.

1980-01-01

In selection, calibration and interpretation of personnel neutron dosimeters used in radiation protection, adequate attention is often not given to matching the characteristics of the dosimeter with the quality of the neutron field. A particular concern is the use of albedo detectors which have little energy response similarity to the neutron dose equivalent conversion curve. At the Lawrence Livermore Laboratory we have developed a system for dosimeter calibration and neutron field characterization using Bonner spheres and remmeters. Rapid surveys of the work area with detectors in 3-in and 9-in polyethylene spheres establish a qualitative estimate of spectral variation found in the facility. We also use this data to determine the appropriate albedo dosimeter calibration factors. At several locations representing the spectral range, multisphere spectra measure-ments are made and the spectrum weighted dose equivalent rates calculated. These rates are compared with survey instrument results to establish correction factors for the relative over- or under-response expected from these instruments, particularly in highly moderated neutron fields where remmeters overrespond. We also use the spectral information to determine the appropriateness of dosimeters considered for future use. This technique has been applied at power reactors to provide information valuable to selection of proper personnel dosimeters. We find that the spectral range is sufficiently narrow that albedo detectors can be used with confidence. On the other hand, most of the dose occurs at energies below the effective threshold NTA film. (author)

The energy spectrum of cosmic-ray induced neutrons measured on an airplane over a wide range of altitude and latitude

International Nuclear Information System (INIS)

Goldhagen, P.; Clem, J. M.; Wilson, J. W.

2004-01-01

Crews of high-altitude aircraft are exposed to radiation from galactic cosmic rays (GCRs). To help determine such exposures, the Atmospheric Ionizing Radiation Project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on a NASA ER-2 high-altitude airplane. The primary instrument was a sensitive extended-energy multisphere neutron spectrometer. Its detector responses were calculated for energies up to 100 GeV using the radiation transport code MCNPX 2.5.d with improved nuclear models and including the effects of the airplane structure. New calculations of GCR-induced particle spectra in the atmosphere were used to correct for spectrometer counts produced by protons, pions and light nuclear ions. Neutron spectra were unfolded from the corrected measured count rates using the deconvolution code MAXED 3.1. The results for the measured cosmic-ray neutron spectrum (thermal to >10 GeV), total neutron fluence rate, and neutron dose equivalent and effective dose rates, and their dependence on altitude and geomagnetic cut-off agree well with results from recent calculations of GCR-induced neutron spectra. (authors)

Accelerator based neutron source for neutron capture therapy

International Nuclear Information System (INIS)

Salimov, R.; Bayanov, B.; Belchenko, Yu.; Belov, V.; Davydenko, V.; Donin, A.; Dranichnikov, A.; Ivanov, A.; Kandaurov, I; Kraynov, G.; Krivenko, A.; Kudryavtsev, A.; Kursanov, N.; Savkin, V.; Shirokov, V.; Sorokin, I.; Taskaev, S.; Tiunov, M.

2004-01-01

Full text: The Budker Institute of Nuclear Physics (Novosibirsk) and the Institute of Physics and Power Engineering (Obninsk) have proposed an accelerator based neutron source for neutron capture and fast neutron therapy for hospital. Innovative approach is based upon vacuum insulation tandem accelerator (VITA) and near threshold 7 Li(p,n) 7 Be neutron generation. Pilot accelerator based neutron source for neutron capture therapy is under construction now at the Budker Institute of Nuclear Physics, Novosibirsk, Russia. In the present report, the pilot facility design is presented and discussed. Design features of facility components are discussed. Results of experiments and simulations are presented. Complete experimental tests are planned by the end of the year 2005

Determination of the neutron and photon dose equivalent at work places in nuclear facilities of Sweden. An SSI - EURADOS comparison exercise. Part 2: Evaluation

Energy Technology Data Exchange (ETDEWEB)

Bartlett, D. [National Radiological Protection Board, Chilton (United Kingdom); Drake, P. [Vattenfall AB, Vaeroebacka (Sweden); Lindborg, L. [Swedish Radiation Protection Inst., Stockholm (Sweden); Klein, H. [Physikalisch-Technische Bundesanstalt, Braunschweig (Germany); Schmitz, Th. [Forschungszentrum Juelich GmbH, Juelich (Germany); Tichy, M

1999-06-01

Various mixed neutron-photon fields at workplaces in the containment of pressurised water reactors and in the vicinity of transport containers with spent fuel elements were investigated with spectrometers and dosimeters. The spectral neutron fluences evaluated from measurements with multisphere systems were recommended to be used for the calculation of dosimetric reference values for comparison with the readings of the dosemeters applied simultaneously. It turned out that most of the moderator based area dosemeters overestimated, while the TEPC systems generally underestimated the ambient dose equivalent (DE) values of the rather soft neutron fields encountered at these workplaces. The discrepancies can, however, be explained on the basis of energy dependent responses of the instruments used. The ambient DE values obtained with recently developed area dosemeters based on superheated drop detectors and with track etch based personal dosemeters on phantoms, however, were in satisfying agreement with the reference data. Sets of personal dosemeters simultaneously irradiated on a phantom allowed to roughly estimate the directional dependence of the neutron fluence. Hence, personal and limiting dose equivalent quantities could also be calculated. The personal and ambient DE values were always conservative estimates of the limiting quantities. Unexpectedly, discrepancies were observed for photon DE data measured with GM counters and TEPC systems. The up to 50 % higher readings of the GM counters may be explained by a considerable contribution of high energy photons to the total photon dose equivalent, but photon spectrometry is necessary for final clarification.

Determination of the neutron and photon dose equivalent at work places in nuclear facilities of Sweden. An SSI - EURADOS comparison exercise. Part 2: Evaluation

International Nuclear Information System (INIS)

Bartlett, D.; Drake, P.; Lindborg, L.; Klein, H.; Schmitz, Th.; Tichy, M.

1999-06-01

Various mixed neutron-photon fields at workplaces in the containment of pressurised water reactors and in the vicinity of transport containers with spent fuel elements were investigated with spectrometers and dosimeters. The spectral neutron fluences evaluated from measurements with multisphere systems were recommended to be used for the calculation of dosimetric reference values for comparison with the readings of the dosemeters applied simultaneously. It turned out that most of the moderator based area dosemeters overestimated, while the TEPC systems generally underestimated the ambient dose equivalent (DE) values of the rather soft neutron fields encountered at these workplaces. The discrepancies can, however, be explained on the basis of energy dependent responses of the instruments used. The ambient DE values obtained with recently developed area dosemeters based on superheated drop detectors and with track etch based personal dosemeters on phantoms, however, were in satisfying agreement with the reference data. Sets of personal dosemeters simultaneously irradiated on a phantom allowed to roughly estimate the directional dependence of the neutron fluence. Hence, personal and limiting dose equivalent quantities could also be calculated. The personal and ambient DE values were always conservative estimates of the limiting quantities. Unexpectedly, discrepancies were observed for photon DE data measured with GM counters and TEPC systems. The up to 50 % higher readings of the GM counters may be explained by a considerable contribution of high energy photons to the total photon dose equivalent, but photon spectrometry is necessary for final clarification

Using multi spherical spectrometry for determination of dosimetric characteristics of mixed neutron and gamma radiation fields of fission sources

International Nuclear Information System (INIS)

Fyulep, M.; Nikodemova, D.; Grabovtsova, A.; Galan, P.; Trousil, J.

1977-01-01

Possibilities of the application of multispherical spectrometry in personnel dosimetry of neutrons (n) and gamma radiation (γ) are considered. Studies were made to elucidate a possibility of using albedo dosemeters to increase the sensitivity of personnel dosemeters. Determined were the dose due to the (n,γ) reaction in a human body, absorbed dose and dose equivalent. The effect of (n,γ) dose on the reading of personnel gamma dosemeter was considered. It is shown that the above effect on the dosemeter readings for frontal irradiation by a broad neutron beam in everyday personnel dosimetry near 252 Cf sources may be neglected. Only in the case of strongly slowed-down fission spectrum the effect of the (n,γ) reaction is considerable. The application of albedo dosemeter is expedient to take into account the corrections of personnel dosemeter readings [ru

Neutron Dose Measurement Using a Cubic Moderator

International Nuclear Information System (INIS)

Sheinfeld, M.; Mazor, T.; Cohen, Y.; Kadmon, Y.; Orion, I.

2014-01-01

The Bonner Sphere Spectrometer (BSS), introduced In July 1960 by a research group from Rice University, Texas, is a major approach to neutron spectrum estimation. The BSS, also known as multi-sphere spectrometer, consists of a set of a different diameters polyethylene spheres, carrying a small LiI(Eu) scintillator in their center. What makes this spectrometry method such widely used, is its almost isotropic response, covering an extraordinary wide range of energies, from thermal up to even hundreds of MeVs. One of the most interesting and useful consequences of the above study is the 12'' sphere characteristics, as it turned out that the response curve of its energy dependence, have a similar shape compared with the neutron's dose equivalent as a function of energy. This inexplicable and happy circumstance makes it virtually the only monitoring device capable providing realistic neutron dose estimates over such a wide energy range. However, since the detection mechanism is not strictly related to radiation dose, one can expect substantial errors when applied to widely different source conditions. Although the original design of the BSS included a small 4mmx4mmO 6LiI(Eu) scintillator, other thermal neutron detectors has been used over the years: track detectors, activation foils, BF3 filled proportional counters, etc. In this study we chose a Boron loaded scintillator, EJ-254, as the thermal neutron detector. The neutron capture reaction on the boron has a Q value of 2.78 MeV of which 2.34 MeV is shared by the alpha and lithium particles. The high manufacturing costs, the encasement issue, the installation efficiency and the fabrication complexity, led us to the idea of replacing the sphere with a cubic moderator. This article describes the considerations, as well as the Monte-Carlo simulations done in order to examine the applicability of this idea

Development and Analysis of Volume Multi-Sphere Method Model Generation using Electric Field Fitting

Science.gov (United States)

Ingram, G. J.

Electrostatic modeling of spacecraft has wide-reaching applications such as detumbling space debris in the Geosynchronous Earth Orbit regime before docking, servicing and tugging space debris to graveyard orbits, and Lorentz augmented orbits. The viability of electrostatic actuation control applications relies on faster-than-realtime characterization of the electrostatic interaction. The Volume Multi-Sphere Method (VMSM) seeks the optimal placement and radii of a small number of equipotential spheres to accurately model the electrostatic force and torque on a conducting space object. Current VMSM models tuned using force and torque comparisons with commercially available finite element software are subject to the modeled probe size and numerical errors of the software. This work first investigates fitting of VMSM models to Surface-MSM (SMSM) generated electrical field data, removing modeling dependence on probe geometry while significantly increasing performance and speed. A proposed electric field matching cost function is compared to a force and torque cost function, the inclusion of a self-capacitance constraint is explored and 4 degree-of-freedom VMSM models generated using electric field matching are investigated. The resulting E-field based VMSM development framework is illustrated on a box-shaped hub with a single solar panel, and convergence properties of select models are qualitatively analyzed. Despite the complex non-symmetric spacecraft geometry, elegantly simple 2-sphere VMSM solutions provide force and torque fits within a few percent.

Research of accelerator-based neutron source for boron neutron capture therapy

International Nuclear Information System (INIS)

Li Changkai; Ma Yingjie; Tang Xiaobin; Xie Qin; Geng Changran; Chen Da

2013-01-01

Background: 7 Li (p, n) reaction of high neutron yield and low threshold energy has become one of the most important neutron generating reactions for Accelerator-based Boron Neutron Capture Therapy (BNCT). Purpose Focuses on neutron yield and spectrum characteristics of this kind of neutron generating reaction which serves as an accelerator-based neutron source and moderates the high energy neutron beams to meet BNCT requirements. Methods: The yield and energy spectrum of neutrons generated by accelerator-based 7 Li(p, n) reaction with incident proton energy from 1.9 MeV to 3.0 MeV are researched using the Monte Carlo code-MCNPX2.5.0. And the energy and angular distribution of differential neutron yield by 2.5-MeV incident proton are also given in this part. In the following part, the character of epithermal neutron beam generated by 2.5-MeV incident protons is moderated by a new-designed moderator. Results: Energy spectra of neutrons generated by accelerator-based 7 Li(p, n) reaction with incident proton energy from 1.9 MeV to 3.0 MeV are got through the simulation and calculation. The best moderator thickness is got through comparison. Conclusions: Neutron beam produced by accelerator-based 7 Li(p, n) reaction, with the bombarding beam of 10 mA and the energy of 2.5 MeV, can meet the requirement of BNCT well after being moderated. (authors)

Neutron spectrum measurement inside containment vessel at Kori nuclear power plant unit 1

Energy Technology Data Exchange (ETDEWEB)

Han, J. M.; Kim, T. W.; Kim, K. D.; Youn, C. H. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

2003-10-01

There would be a case for the radiation worker have to work inside of the containment vessel to inspect or repair reactor facilities. In this case, the information about distribution of neutron field is needed to estimate neutron exposure dose of worker. Neutron spectra were measured by BMS(Bonner Multisphere Spectrometer) at 4 points of 6 ft and 20 ft, 2 points of 44 ft, 5 points of 70 ft in containment vessel of Kori unit 1. From the calculation, the following results were obtained. Neutron fluxes of 6 ft were between 2.623 x 10{sup 2} and 2.746 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 9.209 x 10{sup -6} and 3.377 x 10{sup -2} MeV, equivalent doses of neutron were between 0.025 and 2.675 mSv/hr. Neutron fluxes of 20 ft were between 1.771 x 10{sup 1} and 1.682 x 10{sup 3} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 6.084 x 10{sup -6} and 2.988 x 10{sup -1} MeV, equivalent doses of neutron were between 0.004 and 0.228 mSv/hr. Neutron fluxes of 44 ft were between 3.367 x 10{sup 2} and 3.483 x 10{sup 2} neutron / cm{sup 2}{center_dot}sec, average neutron energies were between 3.962 x 10{sup -2} and 7.360 x 10{sup -2} MeV, equivalent doses of neutron were between 0.069 and 0.089 mSv/hr. Neutron fluxes of 70 ft were between 4.553 x 10{sup 3} and 1.407 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 3.668 x 10{sup -4} and 6.764 x 10{sup -2} MeV, equivalent doses of neutron were between 0.449 and 2.660 mSv/hr.

Accelerator based continuous neutron source.

CERN Document Server

Shapiro, S M; Ruggiero, A G

2003-01-01

Until the last decade, most neutron experiments have been performed at steady-state, reactor-based sources. Recently, however, pulsed spallation sources have been shown to be very useful in a wide range of neutron studies. A major review of neutron sources in the US was conducted by a committee chaired by Nobel laureate Prof. W. Kohn: ''Neutron Sources for America's Future-BESAC Panel on Neutron Sources 1/93''. This distinguished panel concluded that steady state and pulsed sources are complementary and that the nation has need for both to maintain a balanced neutron research program. The report recommended that both a new reactor and a spallation source be built. This complementarity is recognized worldwide. The conclusion of this report is that a new continuous neutron source is needed for the second decade of the 20 year plan to replace aging US research reactors and close the US neutron gap. it is based on spallation production of neutrons using a high power continuous superconducting linac to generate pr...

Use of accelerator based neutron sources

International Nuclear Information System (INIS)

2000-05-01

With the objective of discussing new requirements related to the use of accelerator based neutron generators an Advisory Group meeting was held in October 1998 in Vienna. This meeting was devoted to the specific field of the utilization of accelerator based neutron generators. This TECDOC reports on the technical discussions and presentations that took place at this meeting and reflects the current status of neutron generators. The 14 MeV neutron generators manufactured originally for neutron activation analysis are utilised also for nuclear structure and reaction studies, nuclear data acquisition, radiation effects and damage studies, fusion related studies, neutron radiography

Characterization of neutron spectra using sources of {sup 241}AmBe, {sup 238}PuBe e {sup 252}Cf moderated in water; Caracterização de espectros neutrônicos com fontes de {sup 241}AmBe, {sup 238}PuBe e {sup 252}Cf moderados em água

Energy Technology Data Exchange (ETDEWEB)

Gonçalves, A.S.; Silva, F.S.; Patrão, K.C.S.; Fonseca, E.S. da; Pereira, W.W., E-mail: angela.souzagon@gmail.com [Instituto de Radioproteção e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Laboratório de Metrologia de Nêutrons

2017-07-01

Recent studies have demonstrated the importance of characterizing the spectrum of neutron sources for various energies in order to make the understanding of neutron interaction closer to reality they work with. This work presents the determination of neutron energy flux from the source of {sup 241}AmBe (0.6 TBq), {sup 238}PuBe (1.8 TBq) and {sup 252}Cf (120 μg) free in the air and inserted into spheres of various diameters containing distilled water. The determination of the spectrum is based on the measurement and simulation by the Monte Carlo computational method, using the sources under study, with the Bonner multisphere spectrometer containing readings with the detector without sphere and covered with polyethylene balls of diameters: 5,08 cm (2 ″), 7.62 cm (3″), 12.70 cm (5 ″), 20.32 cm (8 ″), 25.40 cm (10 ″) and 30.48 cm (12 ″). It is sought to characterize a new moderate spectrum in water using the sources of {sup 241}AmBe, {sup 238}PuBe and {sup 252}Cf that may be useful for testing, calibration and irradiation of individual and area monitors for neutrons.

Development of beryllium-based neutron target system with three-layer structure for accelerator-based neutron source for boron neutron capture therapy.

Science.gov (United States)

Kumada, Hiroaki; Kurihara, Toshikazu; Yoshioka, Masakazu; Kobayashi, Hitoshi; Matsumoto, Hiroshi; Sugano, Tomei; Sakurai, Hideyuki; Sakae, Takeji; Matsumura, Akira

2015-12-01

The iBNCT project team with University of Tsukuba is developing an accelerator-based neutron source. Regarding neutron target material, our project has applied beryllium. To deal with large heat load and blistering of the target system, we developed a three-layer structure for the target system that includes a blistering mitigation material between the beryllium used as the neutron generator and the copper heat sink. The three materials were bonded through diffusion bonding using a hot isostatic pressing method. Based on several verifications, our project chose palladium as the intermediate layer. A prototype of the neutron target system was produced. We will verify that sufficient neutrons for BNCT treatment are generated by the device in the near future. Copyright © 2015 Elsevier Ltd. All rights reserved.

Progress in neutron beam development at the HFR Petten (feasibility study for a BNCT facility)

International Nuclear Information System (INIS)

Constantine, G.; Moss, R.L.; Watkins, P.R.D.; Perks, C.A.; Delafield, H.J.; Ross, D.; Voorbraak, W.P.; Paardekooper, A.; Freudenreich, W.E.; Stecher-Rasmussen, F.

1990-08-01

Boron Neutron Capture Therapy, using intermediate energy neutrons to achieve the deep penetration essential for treating brain tumours, can be implemented with a filtered reactor neutron beam. This is designed to minimize the mean energy of the neutrons to keep proton recoil damage to the scalp within normal tissue tolerance limits whilst delivering the required thermal neutron fluence to the tumour over a reasonably short period. This can only be realized in conjunction with a high power density reactor. At the Joint Research Centre Petten an optimized neutron filter is currently being built for installation into the HB11 beam tube of the High Flux Reactor HFR. Part of the development leading to this design has been an extensive study of broad spectrum, filtered beam performance on the HB7 beam tube facility. A wide range of calculations was performed using the Monte Carlo code, MCPN, supported by validation experiments in which several filter configuration incorporating aluminium, sulphur, liquid argon, titanium and cadmium were installed for low power measurements of the neutron fluence rate, neutron spectra and beam gamma-ray contamination. The measurements were carried out within a successful European collaboration. Evaluations were made of the reactor core edge and unfiltered beam spectra, for comparison with MCNP calculations. Multi-foil activation methods and also gamma dose determination in the filtered beam using thermo-luminescent detectors were performed by the ECN. The Harwell/ Birmingham University collaborators undertook the neutron spectrum measurements in the filtered beam. proton recoil spectrometry was used above 30 keV, combined with a multi-sphere and BF 3 chamber response modification technique. Subsequent spectrum adjustment was carried out with the SENSAK code. The agreement between the calculated and measured spectra has given confidence in the reactor and filter modelling methods used to design the HB11 therapy facility. (author). 12 refs

Observation of Neutron Skyshine from an Accelerator Based Neutron Source

Energy Technology Data Exchange (ETDEWEB)

Franklyn, C. B. [Radiation Science Department, Necsa, PO Box 582, Pretoria 0001 (South Africa)

2011-12-13

A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >10{sup 11} n{center_dot}s{sup -1}. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.

Cyclotron-based neutron source for BNCT

Energy Technology Data Exchange (ETDEWEB)

Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Ogasawara, T.; Fujita, K. [Sumitomo Heavy Industries, Ltd (Japan); Tanaka, H.; Sakurai, Y.; Maruhashi, A. [Kyoto University Research Reactor Institute (Japan)

2013-04-19

Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation and treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutron collimator is in the range from 100 mm to 250 mm. The peak neutron flux in the water phantom is 1.8 Multiplication-Sign 109 neutrons/cm{sup 2}/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with {sup 10}B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.

Accelerator-based pulsed cold neutron source

International Nuclear Information System (INIS)

Inoue, Kazuhiko; Iwasa, Hirokatsu; Kiyanagi, Yoshiaki

1979-01-01

An accelerator-based pulsed cold neutron source was constructed. The accelerator is a 35 MeV electron linear accelerator with 1 kW average beam power. The cold neutron beam intensity at a specimen is equivalent to that of a research reactor of 10 14 n/cm 2 .s thermal flux in the case of the quasi-elastic neutron scattering measurements. In spite of some limitations to the universal uses, it has been demonstrated by this facility that the modest capacity accelerator-based pulsed cold neutron source is a highly efficient cold neutron source with low capital investment. Design philosophy, construction details, performance and some operational experiences are described. (author)

Development of fast neutron radiography system based on portable neutron generator

Energy Technology Data Exchange (ETDEWEB)

Yi, Chia Jia, E-mail: gei-i-kani@hotmail.com; Nilsuwankosit, Sunchai, E-mail: sunchai.n@chula.ac.th [Department of Nuclear Engineering, Faculty of Engineering, Chulalongkorn University, Phayathai Rd., Patumwan, Bangkok, THAILAND 10330 (Thailand)

2016-01-22

Due to the high installation cost, the safety concern and the immobility of the research reactors, the neutron radiography system based on portable neutron generator is proposed. Since the neutrons generated from a portable neutron generator are mostly the fast neutrons, the system is emphasized on using the fast neutrons for the purpose of conducting the radiography. In order to suppress the influence of X-ray produced by the neutron generator, a combination of a shielding material sandwiched between two identical imaging plates is used. A binary XOR operation is then applied for combining the information from the imaging plates. The raw images obtained confirm that the X-ray really has a large effect and that XOR operation can help enhance the effect of the neutrons.

Study of neutron spectra using sources of {sup 241}AmBE and {sup 238}PuBe moderated in water; Estudo de espectros neutrônicos com fontes de {sup 241}AmBE e {sup 238}PuBe moderados em água

Energy Technology Data Exchange (ETDEWEB)

Gonçalves, Angela S.; Silva, Fellipe S.; Patrão, Karla C.S.; Fonseca, Evaldo S. da; Pereira, Walsan W., E-mail: angela.souzagon@gmail.com [Instituto de Radioprotecao e Dosimetria, (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Laboratorio de Metrologia de Neutrons; Fundação Técnico-Educacional Souza Marques (FTESM), Rio de Janeiro, RJ (Brazil)

2017-11-01

Recent works demonstrate the increasing importance of characterizing the spectrum of neutron sources for various energies. The main objective of this study is to make the understanding of the interaction of neutrons as close as possible to the reality in which the workers act, thus allowing to act directly in the area of radioprotection. In this way, neutron fluence determination of the {sup 241}AmBe source of 0.6 TBq (16 Ci) and {sup 238} PuBe 1.8 TBq (50 Ci) free in the air and inserted in aluminium spheres of 16 cm and 20.5 cm filled with distilled water. The measurements were carried out in the low scattering laboratory of the Laboratory of Neutron Metrology, in order to obtain a more realistic spectrum. Spectrum determination is based on measurement using the Bonner multisphere spectrometer containing readings with the ball-free detector and covered with polyethylene spheres having diameters of: 5,08 cm (2″), 7,62 cm (3″), 12,70 cm (5″), 20,32 cm (8″), 25,40 cm (10″) e 30,48 cm (12″). The aim is to characterize a new moderate spectrum in water using the sources of {sup 238}PuBe and {sup 241}AmBe that may represent realistic fields in the radioprotection area useful for testing, calibration and irradiation of individual and area monitors for neutrons.

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.

Pramana – Journal of Physics | Indian Academy of Sciences

Indian Academy of Sciences (India)

Home; Journals; Pramana – Journal of Physics; Volume 77; Issue 4. Issue front cover thumbnail. Volume 77, Issue 4. October 2011, pages 599-764. pp 599-609 Research Articles. Evaluation of response matrix of a multisphere neutron spectrometer with water moderator · Rahim Khabaz Hashem Miri Hakimabad.

Base neutron noise in PWRs

International Nuclear Information System (INIS)

Kosaly, G.; Albrecht, R.W.; Dailey, D.J.; Fry, D.N.

1981-01-01

Considerable activity has been devoted in recent years to the use of neutron noise for investigation of problems in pressurized-water reactors (PWRs). The investigators have found that neutron noise provides an effective way to monitor reactor internal vibrations such as vertical and lateral core motion; core support barrel and thermal shield shell modes, bending modes of fuel assemblies, and control rod vibrations. However, noise analysts have also concluded that diagnosis of a problem is easier if baseline data for normal plant operation is available. Therefore, the authors have obtained ex-core neutron noise signatures from eight PWRs to determine the similarity of signatures between plants and to build a base of data to determine the sources of neutron noise and thus the potential diagnostic information contained in the data. It is concluded that: (1) ex-core neutron noise contains information about the vibration of components in the pressure vessel; (2) baseline signature acquisition can aid understanding of plant specific vibration frequencies and provide a bases for diagnosis of future problems if they occur; and (3) abnormal core support barrel vibration can most likely be detected over and above the plant-to-plant signature variation observed thus far

Fast neutron dosimeter with wide base silicon diode

International Nuclear Information System (INIS)

Ma Lu

1986-01-01

This paper briefly introduces a wide base silicon diode fast neutron dosimeter with wide measuring range and good energy response to fast neutron. It is suitable to be used to detect fast neutrons in the mixed field of γ-ray, thermal neutrons and fast neutrons

Advances in neutron based bulk explosive detection

Science.gov (United States)

Gozani, Tsahi; Strellis, Dan

2007-08-01

Neutron based explosive inspection systems can detect a wide variety of national security threats. The inspection is founded on the detection of characteristic gamma rays emitted as the result of neutron interactions with materials. Generally these are gamma rays resulting from thermal neutron capture and inelastic scattering reactions in most materials and fast and thermal neutron fission in fissile (e.g.235U and 239Pu) and fertile (e.g.238U) materials. Cars or trucks laden with explosives, drugs, chemical agents and hazardous materials can be detected. Cargo material classification via its main elements and nuclear materials detection can also be accomplished with such neutron based platforms, when appropriate neutron sources, gamma ray spectroscopy, neutron detectors and suitable decision algorithms are employed. Neutron based techniques can be used in a variety of scenarios and operational modes. They can be used as stand alones for complete scan of objects such as vehicles, or for spot-checks to clear (or validate) alarms indicated by another inspection system such as X-ray radiography. The technologies developed over the last two decades are now being implemented with good results. Further advances have been made over the last few years that increase the sensitivity, applicability and robustness of these systems. The advances range from the synchronous inspection of two sides of vehicles, increasing throughput and sensitivity and reducing imparted dose to the inspected object and its occupants (if any), to taking advantage of the neutron kinetic behavior of cargo to remove systematic errors, reducing background effects and improving fast neutron signals.

Accelerator-based epithermal neutron sources for boron neutron capture therapy of brain tumors.

Science.gov (United States)

Blue, Thomas E; Yanch, Jacquelyn C

2003-01-01

This paper reviews the development of low-energy light ion accelerator-based neutron sources (ABNSs) for the treatment of brain tumors through an intact scalp and skull using boron neutron capture therapy (BNCT). A major advantage of an ABNS for BNCT over reactor-based neutron sources is the potential for siting within a hospital. Consequently, light-ion accelerators that are injectors to larger machines in high-energy physics facilities are not considered. An ABNS for BNCT is composed of: (1) the accelerator hardware for producing a high current charged particle beam, (2) an appropriate neutron-producing target and target heat removal system (HRS), and (3) a moderator/reflector assembly to render the flux energy spectrum of neutrons produced in the target suitable for patient irradiation. As a consequence of the efforts of researchers throughout the world, progress has been made on the design, manufacture, and testing of these three major components. Although an ABNS facility has not yet been built that has optimally assembled these three components, the feasibility of clinically useful ABNSs has been clearly established. Both electrostatic and radio frequency linear accelerators of reasonable cost (approximately 1.5 M dollars) appear to be capable of producing charged particle beams, with combinations of accelerated particle energy (a few MeV) and beam currents (approximately 10 mA) that are suitable for a hospital-based ABNS for BNCT. The specific accelerator performance requirements depend upon the charged particle reaction by which neutrons are produced in the target and the clinical requirements for neutron field quality and intensity. The accelerator performance requirements are more demanding for beryllium than for lithium as a target. However, beryllium targets are more easily cooled. The accelerator performance requirements are also more demanding for greater neutron field quality and intensity. Target HRSs that are based on submerged-jet impingement and

High sensitivity MOSFET-based neutron dosimetry

International Nuclear Information System (INIS)

Fragopoulou, M.; Konstantakos, V.; Zamani, M.; Siskos, S.; Laopoulos, T.; Sarrabayrouse, G.

2010-01-01

A new dosemeter based on a metal-oxide-semiconductor field effect transistor sensitive to both neutrons and gamma radiation was manufactured at LAAS-CNRS Laboratory, Toulouse, France. In order to be used for neutron dosimetry, a thin film of lithium fluoride was deposited on the surface of the gate of the device. The characteristics of the dosemeter, such as the dependence of its response to neutron dose and dose rate, were investigated. The studied dosemeter was very sensitive to gamma rays compared to other dosemeters proposed in the literature. Its response in thermal neutrons was found to be much higher than in fast neutrons and gamma rays.

Scintillator Based Coded-Aperture Imaging for Neutron Detection

International Nuclear Information System (INIS)

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

2013-06-01

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

Neutron matter, neutron pairing, and neutron drops based on chiral effective field theory interactions

Energy Technology Data Exchange (ETDEWEB)

Krueger, Thomas

2016-10-19

calculate the pairing gaps in neutron matter and provide uncertainty estimates. The formation of heavy elements in the early universe proceeds through the rapid neutron-capture process. This process requires precise knowledge of the properties of very neutron-rich nuclei, which are unstable and at present not accessible in experiments. Thus, one can explore their properties only with theoretical calculations. Currently the only approach to the properties of all nuclei are energy-density functionals (EDFs). All EDFs used today are based on phenomenological models and fits to stable nuclei, which makes their predictive power for unknown (neutron-rich) nuclei unclear. Deriving an ab initio EDF directly from the nuclear forces is an important goal of nuclear theory. A promising approach is the optimised effective potential (OEP) method. We take a step into that direction and calculate neutron drops within the OEP formalism. In addition to the exact-exchange approximation we study for the first time the effect of second-order contributions and compare to quantum Monte Carlo and other results.

Calculations of accelerator-based neutron sources characteristics

International Nuclear Information System (INIS)

Tertytchnyi, R.G.; Shorin, V.S.

2000-01-01

Accelerator-based quasi-monoenergetic neutron sources (T(p,n), D(d;n), T(d;n) and Li (p,n)-reactions) are widely used in experiments on measuring the interaction cross-sections of fast neutrons with nuclei. The present work represents the code for calculation of the yields and spectra of neutrons generated in (p, n)- and ( d; n)-reactions on some targets of light nuclei (D, T; 7 Li). The peculiarities of the stopping processes of charged particles (with incident energy up to 15 MeV) in multilayer and multicomponent targets are taken into account. The code version is made in terms of the 'SOURCE,' a subroutine for the well-known MCNP code. Some calculation results for the most popular accelerator- based neutron sources are given. (authors)

Characterization of the neutron spectra at the final of the installations labyrinth with medical accelerators; Caracterizacion del espectro de neutrones al final del laberinto de instalaciones con aceleradores medicos

Energy Technology Data Exchange (ETDEWEB)

Carelli, J.; Cruzate, J.A.; Gregori, B.; Papadopulos, S.; Discacciatti, A. [Autoridad Reguladora Nuclear, Av. del Libertador 8250, Buenos Aires (Argentina)]. e-mail: jcarelli@cae.arn.gov.ar

2006-07-01

A linear electron accelerator for medical use is an equipment dedicated to the production of collimated beams of electrons and/or photons. In an accelerator of a bigger potential or equal to 6 MV, are produced neutrons starting from the reaction (gamma, n) due to the interaction of the photons with the materials that compose the headset and the target. In this work the theoretical and experimental studies carried out to characterize the neutron spectra to the exit of the labyrinth of three bunkers of different geometry with accelerators of 15 MV, with the purpose of evaluating the effective dose of the occupationally exposure personnel are presented. It was carried out the simulation of the neutron transport with the MCNPX code and the ENDF/B - VI library. With the objective of analyzing the variables that affect the spectral distribution the bunkers of two existent facilities in Argentina were modeled. It was considered a isotropic punctual source located in the supposed position of the target. The spectra of {sup 252} Cf and of Watt of 1.8 MeV of half energy were simulated. The election of the sources was based on published works that suppose initial neutron sources with half energy between 1.8 and 2.3 MeV for accelerators of 15 at 25 MV. Its were considered headsets of different dimensions, with and without phantom of water disperser in the patient's position and several field dimensions in the isocenter. The spectral distribution doesn't present significant differences in the different modeling situations. Its were carried out measurements, with the multisphere spectrometric system based on twelve polyethylene spheres and a spherical detector of {sup 3} He, to the exit of each one of the bunkers. It was carried out the convolution of the spectrum using the MXD{sub F}C33 code (of the UMG33 set), considering as initial spectrum that of the fission type (inverse of the energy). The obtained spectra and the environmental equivalent dose rate in each case

Larmor-precession based neutron scattering instrumentation

International Nuclear Information System (INIS)

Ioffe, Alexander

2009-01-01

The Larmor precession of the neutron spin in a magnetic field allows the attachment of a Larmor clock to every neutron. Such Larmor labelling opens the possibility for the development of unusual neutron scattering techniques, where the energy (momentum) resolution does not require the initial and final states to be well selected. This principally allows for achievement of very high energy (momentum) resolution that is not feasible at all with conventional neutron scattering techniques, because the required neutron beam monochromatization (collimation) will result in intolerable intensity losses. Such decoupling of resolution and collimation allows, for example, for a significant increase in the luminosity of small-angle scattering or high-resolution diffractometers; the fact that opens new perspectives for their implementation at middle flux neutron sources. Different kinds of Larmor clock-based instrumentation, particularly two alternative NSE techniques using rotating and time-gradient magnetic field arrangements, which can be considered as inexpensive and affordable alternatives to present day NSE techniques, will be discussed and results of simulations and first experiments will be presented. (author)

Synchrotron based spallation neutron source concepts

International Nuclear Information System (INIS)

Cho, Y.

1998-01-01

During the past 20 years, rapid-cycling synchrotrons (RCS) have been used very productively to generate short-pulse thermal neutron beams for neutron scattering research by materials science communities in Japan (KENS), the UK (ISIS) and the US (IPNS). The most powerful source in existence, ISIS in the UK, delivers a 160-kW proton beam to a neutron-generating target. Several recently proposed facilities require proton beams in the MW range to produce intense short-pulse neutron beams. In some proposals, a linear accelerator provides the beam power and an accumulator ring compresses the pulse length to the required ∼ 1 micros. In others, RCS technology provides the bulk of the beam power and compresses the pulse length. Some synchrotron-based proposals achieve the desired beam power by combining two or more synchrotrons of the same energy, and others propose a combination of lower and higher energy synchrotrons. This paper presents the rationale for using RCS technology, and a discussion of the advantages and disadvantages of synchrotron-based spallation sources

An accelerator based steady state neutron source

International Nuclear Information System (INIS)

Burke, R.J.; Johnson, D.L.

1985-01-01

Using high current, c.w. linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the Accelerator Based Neutron Research Facility (ABNR) would initially achieve the 10 16 n/cm 2 .s thermal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of $300-450M is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source in most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc. With the development of multi-beam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs

Test of sup 3 He-based neutron polarizers at NIST

CERN Document Server

Jones, G L; Thompson, A K; Chowdhuri, Z; Dewey, M S; Snow, W M; Wietfeldt, F E

2000-01-01

Neutron spin filters based on polarized sup 3 He are useful over a wide neutron energy range and have a large angular acceptance among other advantages. Two optical pumping methods, spin-exchange and metastability-exchange, can produce the volume of highly polarized sup 3 He gas required for such neutron spin filters. We report a test of polarizers based on each of these two methods on a new cold, monochromatic neutron beam line at the NIST Center for Neutron Research.

A neutron spectrum unfolding code based on iterative procedures

International Nuclear Information System (INIS)

Ortiz R, J. M.; Vega C, H. R.

2012-10-01

In this work, the version 3.0 of the neutron spectrum unfolding code called Neutron Spectrometry and Dosimetry from Universidad Autonoma de Zacatecas (NSDUAZ), is presented. This code was designed in a graphical interface under the LabVIEW programming environment and it is based on the iterative SPUNIT iterative algorithm, using as entrance data, only the rate counts obtained with 7 Bonner spheres based on a 6 Lil(Eu) neutron detector. The main features of the code are: it is intuitive and friendly to the user; it has a programming routine which automatically selects the initial guess spectrum by using a set of neutron spectra compiled by the International Atomic Energy Agency. Besides the neutron spectrum, this code calculates the total flux, the mean energy, H(10), h(10), 15 dosimetric quantities for radiation protection porpoises and 7 survey meter responses, in four energy grids, based on the International Atomic Energy Agency compilation. This code generates a full report in html format with all relevant information. In this work, the neutron spectrum of a 241 AmBe neutron source on air, located at 150 cm from detector, is unfolded. (Author)

Accelerator-based neutron source and its future

International Nuclear Information System (INIS)

Kiyanagi, Yoshiaki

2008-01-01

Neutrons are useful tool for the material science and also for the industrial applications. Now, high intensity neutron sources based on MW class big accelerators are under commissioning in Japan, Japan Spallation Neutron Source (JSNS) at J-PARC and in the US, SNS. Such high power neutron sources required the moderators that can be used under high radiation field and also give high neutronic performance. We have been performing experimental and Monte Carlo simulation studies to develop the cold neutron moderator systems for the high power sources since it is becoming important for materials and life science. Hydrogen is the unique candidate at the present stage due to its high resistibility to the radiation. It was indicated the para hydrogen moderator gave a good neutronic performance by experimental results. On the other hand, in the future, low power neutron sources are recognized to be useful to perform sprouting experiments and to promote the neutron science. The moderator systems need a concept different from the high power source. Therefore, we studied neutronic performances of the mesitylene and the methane moderators to get high intensity in a definite area on the moderator surface. Single groove moderators were studied and optimal geometry and the intensity gain were obtained. The mesitylene moderator gave a rather good performance compared to the methane moderator. (author)

Neutron diffraction measurements at the INES diffractometer using a neutron radiative capture based counting technique

Energy Technology Data Exchange (ETDEWEB)

Festa, G. [Centro NAST, Universita degli Studi di Roma Tor Vergata, Roma (Italy); Pietropaolo, A., E-mail: antonino.pietropaolo@roma2.infn.it [Centro NAST, Universita degli Studi di Roma Tor Vergata, Roma (Italy); Grazzi, F.; Barzagli, E. [CNR-ISC Firenze (Italy); Scherillo, A. [CNR-ISC Firenze (Italy); ISIS facility Rutherford Appleton Laboratory (United Kingdom); Schooneveld, E.M. [ISIS facility Rutherford Appleton Laboratory (United Kingdom)

2011-10-21

The global shortage of {sup 3}He gas is an issue to be addressed in neutron detection. In the context of the research and development activity related to the replacement of {sup 3}He for neutron counting systems, neutron diffraction measurements performed on the INES beam line at the ISIS pulsed spallation neutron source are presented. For these measurements two different neutron counting devices have been used: a 20 bar pressure squashed {sup 3}He tube and a Yttrium-Aluminum-Perovskite scintillation detector. The scintillation detector was coupled to a cadmium sheet that registers the prompt radiative capture gamma rays generated by the (n,{gamma}) nuclear reactions occurring in cadmium. The assessment of the scintillator based counting system was done by performing a Rietveld refinement analysis on the diffraction pattern from an ancient Japanese blade and comparing the results with those obtained by a {sup 3}He tube placed at the same angular position. The results obtained demonstrate the considerable potential of the proposed counting approach based on the radiative capture gamma rays at spallation neutron sources.

Development of a neutron imager based on superconducting detectors

Energy Technology Data Exchange (ETDEWEB)

Miyajima, Shigeyuki, E-mail: miyajima@nict.go.jp [Department of Physics and Engineering, Osaka Prefecture University (Japan); Institute for Nanofabrication Research, Osaka Prefecture University (Japan); Yamaguchi, Hiroyuki; Nakayama, Hirotaka; Shishido, Hiroaki [Department of Physics and Engineering, Osaka Prefecture University (Japan); Institute for Nanofabrication Research, Osaka Prefecture University (Japan); Fujimaki, Akira [Department of Quantum Engineering, Nagoya University (Japan); Hidaka, Mutsuo [National Institute of Advanced Industrial Science and Technology (Japan); Harada, Masahide; Oikawa, Kenichi; Oku, Takayuki; Arai, Masatoshi [J-PARC Center, Japan Atomic Energy Agency (Japan); Ishida, Takekazu [Department of Physics and Engineering, Osaka Prefecture University (Japan); Institute for Nanofabrication Research, Osaka Prefecture University (Japan)

2016-11-15

Highlights: • A neutron detector based on superconducting meander line is demonstrated. • Fast response time of a few tens ns is obtained. • Spatial resolution is 1 μm and can be improved to sub-μm scale. • The proposed neutron detector can operate under the γ-ray fields. - Abstract: We succeeded in demonstrating a neutron detector based on a Nb superconducting meander line with a {sup 10}B conversion layer for a neutron imager based on superconductor devices. We use a current-biased kinetic inductance detector (CB-KID), which is composed of a meander line, for detection of a neutron with high spatial resolution and fast response time. The thickness of Nb meander lines is 40 nm and the line width is narrower than 3 mu m. The area of 8 mm × 8 mm is covered by CB-KIDs, which are assembled at the center of the Si chip of the size 22 mm × 22 mm. The Nb CB-KIDs with a {sup 10}B conversion layer output the voltage by irradiating pulsed neutrons. We have investigated γ/n discrimination of a Nb-based CB-KID with {sup 10}B conversion layer using a Cd plate, which indicates that a CB-KID can operate as a neutron detector under the strong γ-ray fields.

Development of a neutron imager based on superconducting detectors

International Nuclear Information System (INIS)

Miyajima, Shigeyuki; Yamaguchi, Hiroyuki; Nakayama, Hirotaka; Shishido, Hiroaki; Fujimaki, Akira; Hidaka, Mutsuo; Harada, Masahide; Oikawa, Kenichi; Oku, Takayuki; Arai, Masatoshi; Ishida, Takekazu

2016-01-01

Highlights: • A neutron detector based on superconducting meander line is demonstrated. • Fast response time of a few tens ns is obtained. • Spatial resolution is 1 μm and can be improved to sub-μm scale. • The proposed neutron detector can operate under the γ-ray fields. - Abstract: We succeeded in demonstrating a neutron detector based on a Nb superconducting meander line with a "1"0B conversion layer for a neutron imager based on superconductor devices. We use a current-biased kinetic inductance detector (CB-KID), which is composed of a meander line, for detection of a neutron with high spatial resolution and fast response time. The thickness of Nb meander lines is 40 nm and the line width is narrower than 3 mu m. The area of 8 mm × 8 mm is covered by CB-KIDs, which are assembled at the center of the Si chip of the size 22 mm × 22 mm. The Nb CB-KIDs with a "1"0B conversion layer output the voltage by irradiating pulsed neutrons. We have investigated γ/n discrimination of a Nb-based CB-KID with "1"0B conversion layer using a Cd plate, which indicates that a CB-KID can operate as a neutron detector under the strong γ-ray fields.

Evaluation of moderator assemblies for use in an accelerator-based neutron source for boron neutron capture therapy

International Nuclear Information System (INIS)

Woollard, J.E.; Blue, T.E.; Gupta, N.; Gahbauer, R.A.

1998-01-01

The neutron fields produced by several moderator assemblies were evaluated using both in-phantom and in-air neutron field assessment parameters. The parameters were used to determine the best moderator assembly, from among those evaluated, for use in the accelerator-based neutron source for boron neutron capture therapy. For a 10-mA proton beam current and the specified treatment parameters, a moderator assembly consisting of a BeO moderator and a Li 2 CO 3 reflector was found to be the best moderator assembly whether the comparison was based on in-phantom or in-air neutron field assessment parameters. However, the parameters were discordant regarding the moderator thickness. The in-phantom neutron field assessment parameters predict 20 cm of BeO as the best moderator thickness, whereas the in-air neutron field assessment parameters predict 25 cm of BeO as the best moderator thickness

Measurement channel of neutron flow based on software

International Nuclear Information System (INIS)

Rivero G, T.; Benitez R, J. S.

2008-01-01

The measurement of the thermal power in nuclear reactors is based mainly on the measurement of the neutron flow. The presence of these in the reactor core is associated to neutrons released by the fission reaction of the uranium-235. Once moderate, these neutrons are precursors of new fissions. This process it is known like chain reaction. Thus, the power to which works a nuclear reactor, he is proportional to the number of produced fissions and as these depend on released neutrons, also the power is proportional to the number of present neutrons. The measurement of the thermal power in a reactor is realized with called instruments nuclear channels. To low power (level source), these channels measure the individual counts of detected neutrons, whereas to a medium and high power, they measure the electrical current or fluctuation of the same one that generate the fission neutrons in ionization chambers especially designed to detect neutrons. For the case of TRIGA reactors, the measurement channels of neutron flow use discreet digital electronic technology makes some decades already. Recently new technological tools have arisen that allow developing new versions of nuclear channels of simple form and compacts. The present work consists of the development of a nuclear channel for TRIGA reactors based on the use of the correlated signal of a fission chamber for ample interval. This new measurement channel uses a data acquisition card of high speed and the data processing by software that to the being installed in a computer is created a virtual instrument, with what spreads in real time, in graphic and understandable form for the operator, the power indication to which it operates the nuclear reactor. This system when being based on software, offers a major versatility to realize changes in the signal processing and power monitoring algorithms. The experimental tests of neutronic power measurement show a reliable performance through seven decades of power, with a

Thin film CdTe based neutron detectors with high thermal neutron efficiency and gamma rejection for security applications

Energy Technology Data Exchange (ETDEWEB)

Smith, L.; Murphy, J.W. [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States); Kim, J. [Korean Research Institute of Standards and Science, Daejeon 305-600 (Korea, Republic of); Rozhdestvenskyy, S.; Mejia, I. [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States); Park, H. [Korean Research Institute of Standards and Science, Daejeon 305-600 (Korea, Republic of); Allee, D.R. [Flexible Display Center, Arizona State University, Phoenix, AZ 85284 (United States); Quevedo-Lopez, M. [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States); Gnade, B., E-mail: beg031000@utdallas.edu [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States)

2016-12-01

Solid-state neutron detectors offer an alternative to {sup 3}He based detectors, but suffer from limited neutron efficiencies that make their use in security applications impractical. Solid-state neutron detectors based on single crystal silicon also have relatively high gamma-ray efficiencies that lead to false positives. Thin film polycrystalline CdTe based detectors require less complex processing with significantly lower gamma-ray efficiencies. Advanced geometries can also be implemented to achieve high thermal neutron efficiencies competitive with silicon based technology. This study evaluates these strategies by simulation and experimentation and demonstrates an approach to achieve >10% intrinsic efficiency with <10{sup −6} gamma-ray efficiency.

Thermal neutron detectors based on complex oxide crystals

CERN Document Server

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

2002-01-01

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

Accelerator Based Neutron Beams for Neutron Capture Therapy

International Nuclear Information System (INIS)

Yanch, Jacquelyn C.

2003-01-01

The DOE-funded accelerator BNCT program at the Massachusetts Institute of Technology has resulted in the only operating accelerator-based epithermal neutron beam facility capable of generating significant dose rates in the world. With five separate beamlines and two different epithermal neutron beam assemblies installed, we are currently capable of treating patients with rheumatoid arthritis in less than 15 minutes (knee joints) or 4 minutes (finger joints) or irradiating patients with shallow brain tumors to a healthy tissue dose of 12.6 Gy in 3.6 hours. The accelerator, designed by Newton scientific Incorporated, is located in dedicated laboratory space that MIT renovated specifically for this project. The Laboratory for Accelerator Beam Applications consists of an accelerator room, a control room, a shielded radiation vault, and additional laboratory space nearby. In addition to the design, construction and characterization of the tandem electrostatic accelerator, this program also resulted in other significant accomplishments. Assemblies for generating epithermal neutron beams were designed, constructed and experimentally evaluated using mixed-field dosimetry techniques. Strategies for target construction and target cooling were implemented and tested. We demonstrated that the method of submerged jet impingement using water as the coolant is capable of handling power densities of up to 6 x 10(sup 7) W/m(sup 2) with heat transfer coefficients of 10(sup 6)W/m(sup 2)-K. Experiments with the liquid metal gallium demonstrated its superiority compared with water with little effect on the neutronic properties of the epithermal beam. Monoenergetic proton beams generated using the accelerator were used to evaluate proton RBE as a function of LET and demonstrated a maximum RBE at approximately 30-40 keV/um, a finding consistent with results published by other researchers. We also developed an experimental approach to biological intercomparison of epithermal beams and

Neutron-based portable drug probe

International Nuclear Information System (INIS)

Womble, P. C.; Vourvopoulos, G.; Ball Howard, J.; Paschal, J.

1999-01-01

Based on previous measurements, a probe prototype for contraband detection utilizing the neutron technique of Pulsed Fast-Thermal Neutron Analysis (PFTNA) is being constructed. The prototype weighs less than 45 kg and is composed of a probe (5 cm diameter), a power pack and a data acquisition and display system. The probe is designed to be inserted in confined spaces such as the boiler of a ship or a tanker truck filled with liquid. The probe provides information on a) the elemental content, and b) the density variations of the interrogated object. By measuring elemental content, the probe can differentiate between innocuous materials and drugs. Density variations can be found through fast neutron transmission. In all cases, hidden drugs are identified through the measurement of the elemental content of the object, and the comparison of expected and measured elemental ratios

Neutronics assessment of thorium-based fuel assembly in SCWR

International Nuclear Information System (INIS)

Liu, Shichang; Cai, Jiejin

2013-01-01

Highlights: • A novel thorium-based fuel assembly for SCWR has been introduced and investigated. • Neutronic properties of three thorium fuels have been studied, compared with UO 2 fuel. • The thorium-based fuel has advantages on fuel utilization and lower MAs generation. -- Abstract: Aiming to take advantage of neutron spectrum of SCWR, a novel thorium-based fuel assembly for SCWR is introduced in this paper. The neutronic characteristics of the introduced fuel assembly with three different thorium fuel types have been investigated using the “dragon” codes. The parameters in different working conditions, such as infinite multiplication factors, radial power peaking factor, temperature coefficient of reactivity and their relation with the operation period have been assessed by comparing with conventional uranium assembly. Moreover, the moderator-to-fuel ratio (MFR) was changed in order to investigate its influence on the neutronic characteristics of fuel assembly. Results show that the thorium-based fuel has advantages on both efficient fuel utilization and lower minor actinide generation, with some similar neutronic properties to the uranium fuel

A practical neutron shielding design based on data-base interpolation

International Nuclear Information System (INIS)

Jiang, S.H.; Sheu, R.J.

1993-01-01

Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry

Preliminary design of GDT-based 14 MeV neutron source

International Nuclear Information System (INIS)

Du Hongfei; Chen Dehong; Wang Hui; Wang Fuqiong; Jiang Jieqiong; Wu Yican; Chen Yiping

2012-01-01

To meet the need of D-T fusion neutron source for fusion material testing, design goals were presented in this paper according to the international requirements of neutron source for fusion material testing. A preliminary design scheme of GDT-based 14 MeV neutron source was proposed, and a physics model of the neutron source was built based on progress of GDT experiments. Two preliminary design schemes (i. e. FDS-GDT1, FDS-GDT2) were designed; among which FDS-GDT2 can be used for fusion material testing with neutron first wall loading of 2 MW/m 2 . (authors)

IEC-based neutron generator for security inspection system

International Nuclear Information System (INIS)

Miley, G.H.; Wu, L.; Kim, H.J.

2005-01-01

Use of a combined X-ray and neutron source for security inspections based on Inertial Electrostatic Confinement (IEC) fusion is discussed. Current inspection systems typically use X-ray techniques, but thermal neutron analysis (TNA) and fast neutron analysis (FNA), allow expanded detection of certain types of explosives. The integrated unit proposed here uses three separate IEC sources producing 14 and 2.45 MeV neutrons plus soft X-rays. This combination allows multiple detection methods with the composite signal analysis being done by a fuzzy logic system, significantly reducing false signals. (author)

Multi-sphere unit cell model to calculate the effective thermal conductivity in pebble bed reactors

International Nuclear Information System (INIS)

Van Antwerpen, W.; Rousseau, P.G.; Du Toit, C.G.

2010-01-01

A proper understanding of the mechanisms of heat transfer, fluid flow and pressure drop through a packed bed of spheres is of utmost importance in the design of a high temperature Pebble Bed Reactor (PBR). While the gas flows predominantly in the axial direction through the bed, the total effective thermal conductivity is a lumped parameter that characterises the total heat transfer in the radial direction through the packed bed. The study of the effective thermal conductivity is important because it forms an intricate part of the self-acting decay heat removal chain, which is directly related to the PBR safety case. The effective thermal conductivity is the summation of various heat transport phenomena. These are the enhanced thermal conductivity due to turbulent mixing as the fluid passes through the voids between pebbles, heat transfer due to the movement of the solid spheres and thermal conduction and thermal radiation between the spheres in a stagnant fluid environment. In this study, the conduction and radiation between the spheres are investigated. Firstly, existing correlations for the effective thermal conductivity are investigated, with particular attention given to its applicability in the near-wall region. Several phenomena in particular are examined namely: conduction through the spheres, conduction through the contact area between the spheres, conduction through the gas phase and radiation between solid surfaces. A new approach to simulate the effective thermal conductivity for randomly packed beds is then presented, namely the so-called Multi-sphere Unit Cell Model. The model is validated by comparing the results with that obtained in experiments. (authors)

Three new nondestructive evaluation tools based on high flux neutron sources

International Nuclear Information System (INIS)

Hubbard, C.R.; Raine, D.; Peascoe, R.; Wright, M.

1997-01-01

Nondestructive evaluation methods and systems based on specific attributes of neutron interactions with materials are being developed. The special attributes of neutrons are low attenuation in most engineering materials, strong interaction with low Z elements, and epithermal neutron absorption resonances. The three methods under development at ORNL include neutron based tomography and radiography; through thickness, nondestructive texture mapping; and internal, noninvasive temperature measurement. All three techniques require high flux sources such as the High Flux Isotope Reactor, a steady state source, or the Oak Ridge Electron Linear Accelerator, a pulsed neutron source. Neutrons are quite penetrating in most engineering materials and thus can be useful to detect internal flaws and features. Hydrogen atoms, such as in a hydrocarbon fuel, lubricant, or a metal hydride, are relatively opaque to neutron transmission and thus neutron based tomography/radiography is ideal to image their presence. Texture, the nonrandom orientation of crystalline grains within materials, can be mapped nondestructively using neutron diffraction methods. Epithermal neutron resonance absorption is being studied as a noncontacting temperature sensor. This paper highlights the underlying physics of the methods, progress in development, and the potential benefits for science and industry of the three facilities

Current status of accelerator-based boron neutron capture therapy

International Nuclear Information System (INIS)

Kreiner, A. J.; Bergueiro, J.; Di Paolo, H.; Castell, W.; Vento, V. Thatar; Cartelli, D.; Kesque, J.M.; Valda, A.A.; Ilardo, J.C.; Baldo, M.; Erhardt, J.; Debray, M.E.; Somacal, H.R.; Estrada, L.; Sandin, J.C. Suarez; Igarzabal, M.; Huck, H.; Padulo, J.; Minsky, D.M.

2011-01-01

The direct use of proton and heavy ion beams for radiotherapy is a well established cancer treatment modality, which is becoming increasingly widespread due to its clear advantages over conventional photon-based treatments. This strategy is suitable when the tumor is spatially well localized. Also the use of neutrons has a long tradition. Here Boron Neutron Capture Therapy (BNCT) stands out, though on a much smaller scale, being a second-generation promising alternative for tumors which are diffuse and infiltrating. On this sector, so far only nuclear reactors have been used as neutron sources. In this paper we describe the current situation worldwide as far as the use of accelerator-based neutron sources for BNCT is concerned (so-called Accelerator-Based (AB)-BNCT). In particular we discuss the present status of an ongoing project to develop a folded Tandem-ElectroStatic-Quadrupole (TESQ) accelerator at the Atomic Energy Commission of Argentina. The project goal is a machine capable of delivering 30 mA of 2.4 MeV protons to be used in conjunction with a neutron production target based on the 7 Li(p,n) 7 Be reaction. These are the specifications needed to produce sufficiently intense and clean epithermal neutron beams to perform BNCT for deep-seated tumors in less than an hour. (author)

A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki; An, Shigehiro.

1983-01-01

Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

Iterative nonlinear unfolding code: TWOGO

International Nuclear Information System (INIS)

Hajnal, F.

1981-03-01

a new iterative unfolding code, TWOGO, was developed to analyze Bonner sphere neutron measurements. The code includes two different unfolding schemes which alternate on successive iterations. The iterative process can be terminated either when the ratio of the coefficient of variations in terms of the measured and calculated responses is unity, or when the percentage difference between the measured and evaluated sphere responses is less than the average measurement error. The code was extensively tested with various known spectra and real multisphere neutron measurements which were performed inside the containments of pressurized water reactors

Insight on agglomerates of gold nanoparticles in glass based on surface plasmon resonance spectrum: study by multi-spheres T-matrix method

Science.gov (United States)

Avakyan, L. A.; Heinz, M.; Skidanenko, A. V.; Yablunovski, K. A.; Ihlemann, J.; Meinertz, J.; Patzig, C.; Dubiel, M.; Bugaev, L. A.

2018-01-01

The formation of a localized surface plasmon resonance (SPR) spectrum of randomly distributed gold nanoparticles in the surface layer of silicate float glass, generated and implanted by UV ArF-excimer laser irradiation of a thin gold layer sputter-coated on the glass surface, was studied by the T-matrix method, which enables particle agglomeration to be taken into account. The experimental technique used is promising for the production of submicron patterns of plasmonic nanoparticles (given by laser masks or gratings) without damage to the glass surface. Analysis of the applicability of the multi-spheres T-matrix (MSTM) method to the studied material was performed through calculations of SPR characteristics for differently arranged and structured gold nanoparticles (gold nanoparticles in solution, particles pairs, and core-shell silver-gold nanoparticles) for which either experimental data or results of the modeling by other methods are available. For the studied gold nanoparticles in glass, it was revealed that the theoretical description of their SPR spectrum requires consideration of the plasmon coupling between particles, which can be done effectively by MSTM calculations. The obtained statistical distributions over particle sizes and over interparticle distances demonstrated the saturation behavior with respect to the number of particles under consideration, which enabled us to determine the effective aggregate of particles, sufficient to form the SPR spectrum. The suggested technique for the fitting of an experimental SPR spectrum of gold nanoparticles in glass by varying the geometrical parameters of the particles aggregate in the recurring calculations of spectrum by MSTM method enabled us to determine statistical characteristics of the aggregate: the average distance between particles, average size, and size distribution of the particles. The fitting strategy of the SPR spectrum presented here can be applied to nanoparticles of any nature and in various

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

Commissioning of accelerator based boron neutron capture therapy system

International Nuclear Information System (INIS)

Nakamura, S.; Wakita, A.; Okamoto, H.; Igaki, H.; Itami, J.; Ito, M.; Abe, Y.; Imahori, Y.

2017-01-01

Boron neutron capture therapy (BNCT) is a treatment method using a nuclear reaction of 10 B(n, α) 7 Li. BNCT can be deposited the energy to a tumor since the 10 B which has a higher cross-section to a neutron is high is concentrated on the tumor. It is different from conventional radiation therapies that BNCT expects higher treatment effect to radiation resistant tumors since the generated alpha and lithium particles have higher radiological biological effectiveness. In general, BNCT has been performed in research nuclear reactor. Thus, BNCT is not widely applied in a clinical use. According to recent development of accelerator-based boron neutron capture therapy system, the system has an adequate flux of neutrons. Therefore, National Cancer Canter Hospital, Tokyo, Japan is planning to install accelerator based BNCT system. Protons with 2.5 MeV are irradiated to a lithium target system to generate neutrons. As a result, thermal load of the target is 50 kW since current of the protons is 20.0 mA. Additionally, when the accelerator-based BNCT system is installed in a hospital, the facility size is disadvantage in term of neutron measurements. Therefore, the commissioning of the BNCT system is being performed carefully. In this article, we report about the commissioning. (author)

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

Energy Technology Data Exchange (ETDEWEB)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; Folsom, Micah; Kouzes, Richard; Kukharev, Vladislav; Lintereur, Azaree; Robinson, Sean; Siciliano, Edward; Stave, Sean; Valdez, Patrick

2018-04-01

Past 3He shortages led to investigations into replacement options for neutron detectors in systems that previously used 3He-based technologies. The goal of this research was to investigate the feasibility of a full-scale lithium fluoride with silver activated zinc sulfide (LiF/ZnS) based neutron multiplicity counter. The LiF/ZnS based neutron multiplicity counter (LiNMC) was developed based on an iterative process between modeling and experimental measurements. Each active region of the LiNMC contains five sheets of LiF/ZnS sandwiched between six sheets of wavelength shifting plastic to form neutron detection stacks. The wavelength shifted scintillation light was collected by photomultiplier tubes located on each end of the stacks. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high density polyethylene blocks in the corners to reflect high energy neutrons and capture low energy neutrons. Preliminary calibration with a 252Cf neutron source showed that the LiNMC was able to achieve 36% neutron detection efficiency (ε) and an 11.7 μs neutron die-away time (τ) for a doubles Figure-of-merit (ε2/ τ) of 109. This is the highest doubles Figure-of-merit performance measured to-date for a 3He-free neutron multiplicity counter system. By the end of this project, the LiNMC’s basic components were integrated into a single laboratory scale system capable of proof-of-concept measurements.

Standard Practice for Conducting Irradiations at Accelerator-Based Neutron Sources

CERN Document Server

American Society for Testing and Materials. Philadelphia

1996-01-01

1.1 This practice covers procedures for irradiations at accelerator-based neutron sources. The discussion focuses on two types of sources, namely nearly monoenergetic 14-MeV neutrons from the deuterium-tritium T(d,n) interaction, and broad spectrum neutrons from stopping deuterium beams in thick beryllium or lithium targets. However, most of the recommendations also apply to other types of accelerator-based sources, including spallation neutron sources (1). Interest in spallation sources has increased recently due to their proposed use for transmutation of fission reactor waste (2). 1.2 Many of the experiments conducted using such neutron sources are intended to simulate irradiation in another neutron spectrum, for example, that from a DT fusion reaction. The word simulation is used here in a broad sense to imply an approximation of the relevant neutron irradiation environment. The degree of conformity can range from poor to nearly exact. In general, the intent of these simulations is to establish the fundam...

Prospect for application of compact accelerator-based neutron source to neutron engineering diffraction

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Yoshimasa, E-mail: yoshimasa.ikeda@riken.jp [Center for Advanced Photonics, RIKEN, Wako, Saitama 351-0198 (Japan); Taketani, Atsushi; Takamura, Masato; Sunaga, Hideyuki [Center for Advanced Photonics, RIKEN, Wako, Saitama 351-0198 (Japan); Kumagai, Masayoshi [Faculty of Engineering, Tokyo City University, Setagaya, Tokyo 158-8857 (Japan); Oba, Yojiro [Research Reactor Institute, Kyoto University, Kumatori, Osaka 590-0494 (Japan); Otake, Yoshie [Center for Advanced Photonics, RIKEN, Wako, Saitama 351-0198 (Japan); Suzuki, Hiroshi [Materials Sciences Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

2016-10-11

A compact accelerator-based neutron source has been lately discussed on engineering applications such as transmission imaging and small angle scattering as well as reflectometry. However, nobody considers using it for neutron diffraction experiment because of its low neutron flux. In this study, therefore, the neutron diffraction experiments are carried out using Riken Accelerator-driven Compact Neutron Source (RANS), to clarify the capability of the compact neutron source for neutron engineering diffraction. The diffraction pattern from a ferritic steel was successfully measured by suitable arrangement of the optical system to reduce the background noise, and it was confirmed that the recognizable diffraction pattern can be measured by a large sampling volume with 10 mm in cubic for an acceptable measurement time, i.e. 10 min. The minimum resolution of the 110 reflection for RANS is approximately 2.5% at 8 μs of the proton pulse width, which is insufficient to perform the strain measurement by neutron diffraction. The moderation time width at the wavelength corresponding to the 110 reflection is estimated to be approximately 30 μs, which is the most dominant factor to determine the resolution. Therefore, refinements of the moderator system to decrease the moderation time by decreasing a thickness of the moderator or by applying the decoupler system or application of the angular dispersive neutron diffraction technique are important to improve the resolution of the diffraction experiment using the compact neutron source. In contrast, the texture evolution due to plastic deformation was successfully observed by measuring a change in the diffraction peak intensity by RANS. Furthermore, the volume fraction of the austenitic phase in the dual phase mock specimen was also successfully evaluated by fitting the diffraction pattern using a Rietveld code. Consequently, RANS has been proved to be capable for neutron engineering diffraction aiming for the easy access

The stationary neutron radiography system: a TRIGA-based production neutron radiography facility

International Nuclear Information System (INIS)

Chesworth, Robert H.; Hagmann, Dean B.

1988-01-01

General Atomics (GA) is under contract to construct a Stationary Neutron Radiography System (SNRS) - on a turnkey basis - at McClellan Air Force Base in Sacramento, California. The SNRS is a custom designed neutron radiography system which will utilize a 1000 KW TRIGA reactor as the neutron source. The partially below-ground reactor will be equipped with four inclined beam tubes originating near the top of the reactor graphite reflector and installed tangential to the reactor core to provide a strong current of thermal neutrons with minimum gamma ray contamination. The inclined beam tubes will terminate in four large bays and will interface with rugged component positioning systems designed to handle intact aircraft wings, other honeycomb aircraft structures, and pyrotechnics. The SNRS will be equipped with real-time, near real-time, and film radiographic imaging systems to provide a broad spectrum of capability for detection of entrained moisture or corrosion in large aircraft panels. GA is prime contractor to the Air Force for the SNRS and is specifically responsible for the TRIGA reactor system and a portion of the neutron beam system design. Science Applications International Corporation and the Lionakis-Beaumont Design Group are principal subcontractors to GA on the project. (author)

Neutron imaging system based on a video camera

International Nuclear Information System (INIS)

Dinca, M.

2004-01-01

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

Experimental characterization of semiconductor-based thermal neutron detectors

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-21

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

Spectrum shaping of accelerator-based neutron beams for BNCT

CERN Document Server

Montagnini, B; Esposito, J; Giusti, V; Mattioda, F; Varone, R

2002-01-01

We describe Monte Carlo simulations of three facilities for the production of epithermal neutrons for Boron Neutron Capture Therapy (BNCT) and examine general aspects and problems of designing the spectrum-shaping assemblies to be used with these neutron sources. The first facility is based on an accelerator-driven low-power subcritical reactor, operating as a neutron amplifier. The other two facilities have no amplifier and rely entirely on their primary sources, a D-T fusion reaction device and a conventional 2.5 MeV proton accelerator with a Li target, respectively.

Soil-Carbon Measurement System Based on Inelastic Neutron Scattering

International Nuclear Information System (INIS)

Orion, I.; Wielopolski, L.

2002-01-01

Increase in the atmospheric CO 2 is associated with concurrent increase in the amount of carbon sequestered in the soil. For better understanding of the carbon cycle it is imperative to establish a better and extensive database of the carbon concentrations in various soil types, in order to develop improved models for changes in the global climate. Non-invasive soil carbon measurement is based on Inelastic Neutron Scattering (INS). This method has been used successfully to measure total body carbon in human beings. The system consists of a pulsed neutron generator that is based on D-T reaction, which produces 14 MeV neutrons, a neutron flux monitoring detector and a couple of large NaI(Tl), 6'' diameter by 6'' high, spectrometers [4]. The threshold energy for INS reaction in carbon is 4.8 MeV. Following INS of 14 MeV neutrons in carbon 4.44 MeV photons are emitted and counted during a gate pulse period of 10 μsec. The repetition rate of the neutron generator is 104 pulses per sec. The gamma spectra are acquired only during the neutron generator gate pulses. The INS method for soil carbon content measurements provides a non-destructive, non-invasive tool, which can be optimized in order to develop a system for in field measurements

Automatic read out system for superheated emulsion based neutron detector

International Nuclear Information System (INIS)

Meena, J.P.; Parihar, A.; Vaijapurkar, S.G.; Mohan, Anand

2010-01-01

Full text: Defence Laboratory, Jodhpur (DLJ) has developed superheated emulsion technology for neutron and gamma measurements. The laboratory has attempted to develop reader system to display neutron dose and dose rate based on acoustic technique. The paper presents a microcontroller based automatic reader system for neutron measurements using indigenously developed superheated emulsion detector. The system is designed for real time counting of bubbles formed in superheated emulsion detector. A piezoelectric transducer is used for sensing bubble acoustic. The front end of system is mainly consisting of specially designed signal conditioning unit consisted of piezoelectric transducer, an amplifier, a high-pass filter, a differentiator, a comparator and monostable multivibrator. The system is based on PIC 18F6520 microcontroller having large internal SRAM, 10-bit internal ADC, I 2 C interface, UART/USART modules. The paper also describes the design of following peripheral units interfaced to microcontroller temperature and battery monitoring, display, keypad and a serial communication. The reader system measures and displays neutron dose and dose rate, number of bubble and elapsed time. The developed system can be used for detecting very low neutron leakage in the accelerators, nuclear reactors and nuclear submarines. The important features of system are compact, light weight, cost effective and high neutron sensitivity. The prototype was tested and evaluated by exposing to 241 Am-Be neutron source and results have been reported

Monte Carlo simulation of grating-based neutron phase contrast imaging at CPHS

International Nuclear Information System (INIS)

Zhang Ran; Chen Zhiqiang; Huang Zhifeng; Xiao Yongshun; Wang Xuewu; Wie Jie; Loong, C.-K.

2011-01-01

Since the launching of the Compact Pulsed Hadron Source (CPHS) project of Tsinghua University in 2009, works have begun on the design and engineering of an imaging/radiography instrument for the neutron source provided by CPHS. The instrument will perform basic tasks such as transmission imaging and computerized tomography. Additionally, we include in the design the utilization of coded-aperture and grating-based phase contrast methodology, as well as the options of prompt gamma-ray analysis and neutron-energy selective imaging. Previously, we had implemented the hardware and data-analysis software for grating-based X-ray phase contrast imaging. Here, we investigate Geant4-based Monte Carlo simulations of neutron refraction phenomena and then model the grating-based neutron phase contrast imaging system according to the classic-optics-based method. The simulated experimental results of the retrieving phase shift gradient information by five-step phase-stepping approach indicate the feasibility of grating-based neutron phase contrast imaging as an option for the cold neutron imaging instrument at the CPHS.

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.

Proposed Brookhaven accelerator-based neutron generator

International Nuclear Information System (INIS)

Grand, P.; Batchelor, K.; Chasman, R.; Rheaume, R.

1976-01-01

The d-Li Neutron Source concept, which includes a high-current dueteron linac, is an outgrowth of attempts made to use the BNL, 200-MeV proton linac BLIP facility to do radiation damage studies. It included a 100 mA, 30-MeV deuteron linear accelerator and a fast-flowing liquid lithium jet as the target. The latest design is not very different, except that the current is now 200 mA and the linac energy has been raised to 35 MeV. Both parameters, were changed to optimize the effectiveness of the facility with respect to flux, experimental volume and match to 14 MeV neutron-radiation-damage effects. The proposed Brookhaven Accelerator-based Neutron Generator is described with particular emphasis on the linear accelerator. The proposed facility is a practical and efficient way of producing the intense, high energy neutron beams needed for CTR material studies. The accelerator and liquid-metal technologies are well proven, state-of-the-art technologies. The fact that no new technology is required guarantees the possibility of meeting construction schedules, and more importantly, guarantees a high level of operational reliability

A shielding design for an accelerator-based neutron source for boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Hawk, A.E.; Blue, T.E. E-mail: blue.1@osu.edu; Woollard, J.E

2004-11-01

Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a {sup 7}Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design.

Dosimetry in mixed (n,gamma) fields in relation to personnel monitoring

International Nuclear Information System (INIS)

Nikodemova, D.; Hrabovcova, A.; Kubu, M.

1987-01-01

Neutron spectra were determined using neutron detection with a set of moderating spheres and the SAND computer method. A 6 LiI crystal or a pair of 6 LiF- 7 LiF thermoluminescence detectors was placed in the centre of 7 moderating spheres of 5 to 30 cm in diameter. 252 Cf and 241 Am-Be source were used in the calibration of the multisphere spectrometer. Neutron scattering from the walls, floor and ceiling was taken into consideration. The contribution of scattered neutrons was also determined for the case of measuring the 241 Am-Be spectrum with the source placed behind a water wall 10 cm thick. It is shown how the neutron energy distribution and the values of dose quantities altered as a result of the presence of scattered neutrons. The result of the presence of scattered neutrons. The results are discussed with respect to personnel monitoring. (author)

Automatic readout system for superheated emulsion based neutron detector

International Nuclear Information System (INIS)

Meena, J.P.; Parihar, A.; Vaijapurkar, S.G.; Mohan, Anand

2011-01-01

The paper presents a microcontroller based automatic reader system for neutron measurement using indigenously developed superheated emulsion detector. The system is designed for real time counting of bubbles formed in superheated emulsion detector. A piezoelectric transducer is used for sensing bubble acoustic during the nucleation. The front end of system is mainly consisting of specially designed signal conditioning unit, piezoelectric transducer, an amplifier, a high-pass filter, a differentiator, a comparator and monostable multivibrator. The system is based on PlC 18F6520 microcontroller having large internal SRAM, 10-bit internal ADC, I 2 C interface, UART/USART modules. The paper also describes the design of following microcontroller peripheral units viz temperature monitoring, battery monitoring, LCD display, keypad and a serial communication. The reader system measures and displays neutron dose and dose rate, number of bubble and elapsed time. The developed system can be used for detecting very low neutron leakage in the accelerators, nuclear reactors and nuclear submarines. The important features of system are compact, light weight, cost effective and high neutron sensitivity. The prototype was tested and evaluated by exposing to 241 Am-Be neutron source and results have been reported. (author)

A real-time neutron-gamma discriminator based on the support vector machine method for the time-of-flight neutron spectrometer

Science.gov (United States)

Wei, ZHANG; Tongyu, WU; Bowen, ZHENG; Shiping, LI; Yipo, ZHANG; Zejie, YIN

2018-04-01

A new neutron-gamma discriminator based on the support vector machine (SVM) method is proposed to improve the performance of the time-of-flight neutron spectrometer. The neutron detector is an EJ-299-33 plastic scintillator with pulse-shape discrimination (PSD) property. The SVM algorithm is implemented in field programmable gate array (FPGA) to carry out the real-time sifting of neutrons in neutron-gamma mixed radiation fields. This study compares the ability of the pulse gradient analysis method and the SVM method. The results show that this SVM discriminator can provide a better discrimination accuracy of 99.1%. The accuracy and performance of the SVM discriminator based on FPGA have been evaluated in the experiments. It can get a figure of merit of 1.30.

Superintensive pulse slow neutron source SIN based on kaon factory

International Nuclear Information System (INIS)

Kolmichkov, N.V.; Laptev, V.D.; Matveev, V.A.

1991-01-01

Possibility of intensive pulse slow neutron source creation based on 45-GeV proton synchrotron of K-meson factory, planned to construction in INR AS USSR is considered. Calculated peak thermal neutrons flux density value, averaged on 'radiating' light-water moderator surface of 100 cm 2 is 6.6 x 10 17 neutrons/(cm 2 sec) for pulse duration of 35 microseconds. (author)

Gadolinium-Based GaN for Neutron Detection with Gamma Discrimination

Science.gov (United States)

2016-06-01

Gadolinium-Based GaN for Neutron Detection with Gamma Discrimination Distribution Statement A. Approved for public release; distribution is...Final Technical Report BRBAA08-Per5-Y-1-2-0030 Title: “Gadolinium-Based GaN for Neutron Detection with Gamma Discrimination ” Grant...Analysis  .............................................................................................  23   6.   Gamma-ray Discrimination

Simultaneous and integrated neutron-based techniques for material analysis of a metallic ancient flute

International Nuclear Information System (INIS)

Festa, G; Andreani, C; Pietropaolo, A; Grazzi, F; Scherillo, A; Barzagli, E; Sutton, L F; Bognetti, L; Bini, A; Schooneveld, E

2013-01-01

A metallic 19th century flute was studied by means of integrated and simultaneous neutron-based techniques: neutron diffraction, neutron radiative capture analysis and neutron radiography. This experiment follows benchmark measurements devoted to assessing the effectiveness of a multitask beamline concept for neutron-based investigation on materials. The aim of this study is to show the potential application of the approach using multiple and integrated neutron-based techniques for musical instruments. Such samples, in the broad scenario of cultural heritage, represent an exciting research field. They may represent an interesting link between different disciplines such as nuclear physics, metallurgy and acoustics. (paper)

An accelerator-based epithermal photoneutron source for boron neutron capture therapy

International Nuclear Information System (INIS)

Mitchell, H.E.

1996-04-01

Boron neutron capture therapy is an experimental binary cancer radiotherapy modality in which a boronated pharmaceutical that preferentially accumulates in malignant tissue is first administered, followed by exposing the tissue in the treatment volume to a thermal neutron field. Current usable beams are reactor-based but a viable alternative is the production of an epithermal neutron beam from an accelerator. Current literature cites various proposed accelerator-based designs, most of which are based on proton beams with beryllium or lithium targets. This dissertation examines the efficacy of a novel approach to BNCT treatments that incorporates an electron linear accelerator in the production of a photoneutron source. This source may help to resolve some of the present concerns associated with accelerator sources, including that of target cooling. The photoneutron production process is discussed as a possible alternate source of neutrons for eventual BNCT treatments for cancer. A conceptual design to produce epithermal photoneutrons by high photons (due to bremsstrahlung) impinging on deuterium targets is presented along with computational and experimental neutron production data. A clinically acceptable filtered epithermal neutron flux on the order of 10 7 neutrons per second per milliampere of electron current is shown to be obtainable. Additionally, the neutron beam is modified and characterized for BNCT applications by employing two unique moderating materials (an Al/AlF 3 composite and a stacked Al/Teflon design) at various incident electron energies

An accelerator-based epithermal photoneutron source for boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Mitchell, Hannah E. [Georgia Inst. of Technology, Atlanta, GA (United States)

1996-04-01

Boron neutron capture therapy is an experimental binary cancer radiotherapy modality in which a boronated pharmaceutical that preferentially accumulates in malignant tissue is first administered, followed by exposing the tissue in the treatment volume to a thermal neutron field. Current usable beams are reactor-based but a viable alternative is the production of an epithermal neutron beam from an accelerator. Current literature cites various proposed accelerator-based designs, most of which are based on proton beams with beryllium or lithium targets. This dissertation examines the efficacy of a novel approach to BNCT treatments that incorporates an electron linear accelerator in the production of a photoneutron source. This source may help to resolve some of the present concerns associated with accelerator sources, including that of target cooling. The photoneutron production process is discussed as a possible alternate source of neutrons for eventual BNCT treatments for cancer. A conceptual design to produce epithermal photoneutrons by high photons (due to bremsstrahlung) impinging on deuterium targets is presented along with computational and experimental neutron production data. A clinically acceptable filtered epithermal neutron flux on the order of 10⁷ neutrons per second per milliampere of electron current is shown to be obtainable. Additionally, the neutron beam is modified and characterized for BNCT applications by employing two unique moderating materials (an Al/AlF₃ composite and a stacked Al/Teflon design) at various incident electron energies.

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)

Improved Fission Neutron Data Base for Active Interrogation of Actinides

Energy Technology Data Exchange (ETDEWEB)

Pozzi, Sara; Czirr, J. Bart; Haight, Robert; Kovash, Michael; Tsvetkov, Pavel

2013-11-06

This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems both with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).

A FIFO based neutron arrival time collection technique for assay of plutonium

International Nuclear Information System (INIS)

Parthasarathy, R.; Saisubalakshmi, D.; Venkatasubramani, C.R.

2004-01-01

The system assays plutonium by counting the time correlated neutrons emitted by the spontaneous fissions of the even-even Pu isotopes in the presence of random neutron background, originating principally from (a,n) reactions in the material. The correlation technique discussed in this paper utilizes twofold neutron coincidence counting but the system is proposed to be enhanced for neutron multiplicity counting. A microcontroller based data acquisition system has been developed using a couple of fast FIFO 2kX9 bit memory ICs and a 16 bit counter for identifying time-correlated neutrons. Since the neutron pulses are arriving at a rapid rate, the incoming pulses are buffered in the FIFO and then transferred to PC by the microcontroller through the parallel port. The correlation analysis based on this time arrival information is done in the PC off-line. (author)

Measurement of angular distribution of neutron flux for the 6 MeV race-track microtron based pulsed neutron source

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-15

The 6 MeV race track microtron based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products, where the low neutron flux requirement is desirable. Electrons impinges on a e-{gamma} target to generate bremsstrahlung radiations, which further produces neutrons by photonuclear reaction in {gamma}-n target. The optimisation of these targets along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium at different scattering angles. Angular distribution of neutron flux indicates that the flux decreases with increase in the angle and are in good agreement with the FLUKA simulation.

INGDB-90. The International Neutron Nuclear Data Base for geophysics applications

International Nuclear Information System (INIS)

Kocherov, N.P.; McLaughline, P.K.

1991-01-01

This document describes the contents of the International Neutron Nuclear Data Base for applications in nuclear geophysics, such as borehole logging and mineral analysis. It contains neutron cross-section data from 19 elements and their isotopes of primary importance in geophysics, plus a data file with neutron spectra of three frequently used neutron sources. The INGDB-90 file is available, cost free, from the IAEA Nuclear Data Section on PC diskettes or on magnetic tape. (author). 9 refs

A neutron spectrum unfolding computer code based on artificial neural networks

International Nuclear Information System (INIS)

Ortiz-Rodríguez, J.M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J.M.; Vega-Carrillo, H.R.

2014-01-01

The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding

Demonstration of a high-intensity neutron source based on a liquid-lithium target for Accelerator based Boron Neutron Capture Therapy.

Science.gov (United States)

Halfon, S; Arenshtam, A; Kijel, D; Paul, M; Weissman, L; Berkovits, D; Eliyahu, I; Feinberg, G; Kreisel, A; Mardor, I; Shimel, G; Shor, A; Silverman, I; Tessler, M

2015-12-01

A free surface liquid-lithium jet target is operating routinely at Soreq Applied Research Accelerator Facility (SARAF), bombarded with a ~1.91 MeV, ~1.2 mA continuous-wave narrow proton beam. The experiments demonstrate the liquid lithium target (LiLiT) capability to constitute an intense source of epithermal neutrons, for Accelerator based Boron Neutron Capture Therapy (BNCT). The target dissipates extremely high ion beam power densities (>3 kW/cm(2), >0.5 MW/cm(3)) for long periods of time, while maintaining stable conditions and localized residual activity. LiLiT generates ~3×10(10) n/s, which is more than one order of magnitude larger than conventional (7)Li(p,n)-based near threshold neutron sources. A shield and moderator assembly for BNCT, with LiLiT irradiated with protons at 1.91 MeV, was designed based on Monte Carlo (MCNP) simulations of BNCT-doses produced in a phantom. According to these simulations it was found that a ~15 mA near threshold proton current will apply the therapeutic doses in ~1h treatment duration. According to our present results, such high current beams can be dissipated in a liquid-lithium target, hence the target design is readily applicable for accelerator-based BNCT. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

High neutronic efficiency, low current targets for accelerator-based BNCT applications

International Nuclear Information System (INIS)

Powell, J.R.; Ludewig, H.; Todosow, M.

1998-01-01

The neutronic efficiency of target/filters for accelerator-based BNCT applications is measured by the proton current required to achieve a desirable neutron current at the treatment port (10 9 n/cm 2 /s). In this paper the authors describe two possible targeyt/filter concepts wihch minimize the required current. Both concepts are based on the Li-7 (p,n)Be-7 reaction. Targets that operate near the threshold energy generate neutrons that are close tothe desired energy for BNCT treatment. Thus, the filter can be extremely thin (∼ 5 cm iron). However, this approach has an extremely low neutron yield (n/p ∼ 1.0(-6)), thus requiring a high proton current. The proposed solutino is to design a target consisting of multiple extremely thin targets (proton energy loss per target ∼ 10 keV), and re-accelerate the protons between each target. Targets operating at ihgher proton energies (∼ 2.5 MeV) have a much higher yield (n/p ∼ 1.0(-4)). However, at these energies the maximum neutron energy is approximately 800 keV, and thus a neutron filter is required to degrade the average neutron energy to the range of interest for BNCT (10--20 keV). A neutron filter consisting of fluorine compounds and iron has been investigated for this case. Typically a proton current of approximately 5 mA is required to generate the desired neutron current at the treatment port. The efficiency of these filter designs can be further increased by incorporating neutron reflectors that are co-axial with the neutron source. These reflectors are made of materials which have high scattering cross sections in the range 0.1--1.0 MeV

Pulsed neutron sources for epithermal neutrons

International Nuclear Information System (INIS)

Windsor, C.G.

1978-01-01

It is shown how accelerator based neutron sources, giving a fast neutron pulse of short duration compared to the neutron moderation time, promise to open up a new field of epithermal neutron scattering. The three principal methods of fast neutron production: electrons, protons and fission boosters will be compared. Pulsed reactors are less suitable for epithermal neutrons and will only be briefly mentioned. The design principle of the target producing fast neutrons, the moderator and reflector to slow them down to epithermal energies, and the cell with its beam tubes and shielding will all be described with examples taken from the new Harwell electron linac to be commissioned in 1978. A general comparison of pulsed neutron performance with reactors is fraught with difficulties but has been attempted. Calculation of the new pulsed source fluxes and pulse widths is now being performed but we have taken the practical course of basing all comparisons on extrapolations from measurements on the old 1958 Harwell electron linac. Comparisons for time-of-flight and crystal monochromator experiments show reactors to be at their best at long wavelengths, at coarse resolution, and for experiments needing a specific incident wavelength. Even existing pulsed sources are shown to compete with the high flux reactors in experiments where the hot neutron flux and the time-of-flight methods can be best exploited. The sources under construction can open a new field of inelastic neutron scattering based on energy transfer up to an electron volt and beyond

Unfolding code for neutron spectrometry based on neural nets technology

International Nuclear Information System (INIS)

Ortiz R, J. M.; Vega C, H. R.

2012-10-01

The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Neural Networks have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This unfolding code called Neutron Spectrometry and Dosimetry by means of Artificial Neural Networks was designed in a graphical interface under LabVIEW programming environment. The core of the code is an embedded neural network architecture, previously optimized by the R obust Design of Artificial Neural Networks Methodology . The main features of the code are: is easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 Lil(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, only seven rate counts measurement with a Bonner spheres spectrometer are required for simultaneously unfold the 60 energy bins of the neutron spectrum and to calculate 15 dosimetric quantities, for radiation protection porpoises. This code generates a full report in html format with all relevant information. (Author)

Fast neutron spectrometry based on proton detection in CR-39 detector

Energy Technology Data Exchange (ETDEWEB)

Dajko, G.; Somogyi, G.

1986-01-01

The authors have developed a home-made proton-sensitive CR-39 track detector called MA-ND/p. Using this and the n-p scattering process the performance of a fast neutron spectrometer has been studied by applying two different methods. These are based on track density determinations by using varying radiator thicknesses at constant etching time and by using varying etching times at fixed radiator thickness, respectively. For both methods studied a computer programme is made to calculate the theoretically expected neutron sensitivity as a function of neutron energy. For both methods the neutron sensitivities, expressed in terms of observable etched proton tracks per neutron, are determined experimentally for 3.3 and 14.7 MeV neutron energies. The theoretical and experimental data obtained are compared.

Fast neutron spectrometry based on proton detection in CR-39 detector

International Nuclear Information System (INIS)

Dajko, G.; Somogyi, G.

1986-01-01

The authors have developed a home-made proton-sensitive CR-39 track detector called MA-ND/p. Using this and the n-p scattering process the performance of a fast neutron spectrometer has been studied by applying two different methods. These are based on track density determinations by using varying radiator thicknesses at constant etching time and by using varying etching times at fixed radiator thickness, respectively. For both methods studied a computer programme is made to calculate the theoretically expected neutron sensitivity as a function of neutron energy. For both methods the neutron sensitivities, expressed in terms of observable etched proton tracks per neutron, are determined experimentally for 3.3 and 14.7 MeV neutron energies. The theoretical and experimental data obtained are compared. (author)

Calibration of PADC-based neutron area dosemeters in the neutron field produced in the treatment room of a medical LINAC

International Nuclear Information System (INIS)

Bedogni, R.; Domingo, C.; Esposito, A.; Gentile, A.; García-Fusté, M.J.; San-Pedro, M. de; Tana, L.; D’Errico, F.; Ciolini, R.; Di Fulvio, A.

2013-01-01

PADC-based nuclear track detectors have been widely used as convenient ambient dosemeters in many working places. However, due to the large energy dependence of their response in terms of ambient dose equivalent (H ∗ (10)) and to the diversity of workplace fields in terms of energy distribution, the appropriate calibration of these dosemeters is a delicate task. These are among the reasons why ISO has introduced the 12789 Series of Standards, where the simulated workplace neutron fields are introduced and their use to calibrate neutron dosemeters is recommended. This approach was applied in the present work to the UAB PADC-based nuclear track detectors. As a suitable workplace, the treatment room of a 15 MV Varian CLINAC DHX medical accelerator, located in the Ospedale S. Chiara (Pisa), was chosen. Here the neutron spectra in two points of tests (1.5 m and 2 m from the isocenter) were determined with the INFN-LNF Bonner Sphere Spectrometer equipped with Dysprosium activation foils (Dy-BSS), and the values of H ∗ (10) were derived on this basis. The PADC dosemeters were exposed in these points. Their workplace specific H*(10) responses were determined and compared with those previously obtained in different simulated workplace or reference (ISO 8529) neutron fields. - Highlights: ► The neutron field of a medical LINAC was used to calibrate PADC neutron dosemeters. ► The neutron spectra were derived with a Dy-foil based Bonner Sphere Spectrometer. ► Workplace specific calibration factor were derived for the PADC dosemeters. ► These factors were compared with those obtained in reference neutron fields

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

Science.gov (United States)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; Folsom, Micah; Kouzes, Richard; Kukharev, Vladislav; Lintereur, Azaree; Robinson, Sean; Siciliano, Edward; Stave, Sean; Valdez, Patrick

2018-04-01

The feasibility of a full-scale lithium fluoride zinc sulfide (LiF/ZnS) based neutron multiplicity counter has been demonstrated. The counter was constructed of modular neutron detecting stacks that each contain five sheets of LiF/ZnS interleaved between six sheets of wavelength shifting plastic with a photomultiplier tube on each end. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high-density polyethylene blocks in the corners to reflect high-energy neutrons and capture low-energy neutrons. The final system design was optimized via modeling and small-scale test. Measuring neutrons from a 252Cf source, the counter achieved a 36% neutron detection efficiency (ɛ) and an 11 . 7 μs neutron die-away time (τ) for a doubles figure-of-merit (ɛ2 / τ) of 109. This is the highest doubles figure-of-merit measured to-date for a 3He-free neutron multiplicity counter.

Optimum filter-based discrimination of neutrons and gamma rays

International Nuclear Information System (INIS)

Amiri, Moslem; Prenosil, Vaclav; Cvachovec, Frantisek

2015-01-01

An optimum filter-based method for discrimination of neutrons and gamma-rays in a mixed radiation field is presented. The existing filter-based implementations of discriminators require sample pulse responses in advance of the experiment run to build the filter coefficients, which makes them less practical. Our novel technique creates the coefficients during the experiment and improves their quality gradually. Applied to several sets of mixed neutron and photon signals obtained through different digitizers using stilbene scintillator, this approach is analyzed and its discrimination quality is measured. (authors)

Evaluate existing radiation fields

International Nuclear Information System (INIS)

Aldrich, J.M.; Haggard, D.L.; Endres, G.W.R.; Fix, J.J.

1981-01-01

Knowledge of the spectrum of energies for beta, gamma, and neutron radiation experienced in the field is crucial to the proper interpretation of personnel dose. Calibration sources and techniques are determined on the basis of their relationship to field exposure. Selected techniques were used to obtain neutron, photon, and beta energy spectra data at several Hanford locations. Four neutron energy spectra and dose measurement methods were used: (1) multisphere spectrometer system; (2) tissue equivalent proportional counter (TEPC); (3) RASCAL (9'' to 3'' sphere ratios); and (4) helium-3 neutron spectrometer. Gamma spectroscopy was done using standard techniques. A specially designed TLD dosimeter was used to obtain beta spectrum measurements. The design and use of each of these instruments is described in the body of this report. Data collected and analyzed for each of the Hanford locations are included

Unfolding code for neutron spectrometry based on neural nets technology

Energy Technology Data Exchange (ETDEWEB)

Ortiz R, J. M.; Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Apdo. Postal 336, 98000 Zacatecas (Mexico)

2012-10-15

The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Neural Networks have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This unfolding code called Neutron Spectrometry and Dosimetry by means of Artificial Neural Networks was designed in a graphical interface under LabVIEW programming environment. The core of the code is an embedded neural network architecture, previously optimized by the {sup R}obust Design of Artificial Neural Networks Methodology{sup .} The main features of the code are: is easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a {sup 6}Lil(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, only seven rate counts measurement with a Bonner spheres spectrometer are required for simultaneously unfold the 60 energy bins of the neutron spectrum and to calculate 15 dosimetric quantities, for radiation protection porpoises. This code generates a full report in html format with all relevant information. (Author)

A dense plasma focus-based neutron source for a single-shot detection of illicit materials and explosives by a nanosecond neutron pulse

International Nuclear Information System (INIS)

Gribkov, V A; Latyshev, S V; Miklaszewski, R A; Chernyshova, M; Drozdowicz, K; Wiacek, U; Tomaszewski, K; Lemeshko, B D

2010-01-01

Recent progress in a single-pulse Nanosecond Impulse Neutron Investigation System (NINIS) intended for interrogation of hidden objects by means of measuring elastically scattered neutrons is presented in this paper. The method uses very bright neutron pulses having duration of the order of 10 ns only, which are generated by dense plasma focus (DPF) devices filled with pure deuterium or DT mixture as a working gas. The small size occupied by the neutron bunch in space, number of neutrons per pulse and mono-chromaticity (ΔE/E∼1%) of the neutron spectrum provides the opportunity to use a time-of-flight (TOF) technique with flying bases of about a few metres. In our researches we used DPF devices having bank energy in the range 2-7 kJ. The devices generate a neutron yield of the level of 10 8 -10 9 2.45 MeV and 10 10 -10 11 14 MeV neutrons per pulse with pulse duration ∼10-20 ns. TOF base in the tests was 2.2-18.5 m. We have demonstrated the possibility of registering of neutrons scattered by the substances under investigation-1 litre bottles with methanol (CH 3 OH), phosphoric (H 2 PO 4 ) and nitric (HNO 3 ) acids as well as a long object-a 1 m gas tank filled with deuterium at high pressure. It is shown that the above mentioned short TOF bases and relatively low neutron yields are enough to distinguish different elements' nuclei composing the substance under interrogation and to characterize the geometry of lengthy objects in some cases. The wavelet technique was employed to 'clean' the experimental data registered. The advantages and restrictions of the proposed and tested NINIS technique in comparison with other methods are discussed.

A novel fast-neutron tomography system based on a plastic scintillator array and a compact D–D neutron generator

International Nuclear Information System (INIS)

Adams, Robert; Zboray, Robert; Prasser, Horst-Michael

2016-01-01

Very few experimental imaging studies using a compact neutron generator have been published, and to the knowledge of the authors none have included tomography results using multiple projection angles. Radiography results with a neutron generator, scintillator screen, and camera can be seen in Bogolubov et al. (2005), Cremer et al. (2012), and Li et al. (2014). Comparable results with a position-sensitive photomultiplier tube can be seen in Popov et al. (2011). One study using an array of individual fast neutron detectors in the context of cargo scanning for security purposes is detailed in Eberhardt et al. (2005). In that case, however, the emphasis was on very large objects with a resolution on the order of 1 cm, whereas this study focuses on less massive objects and a finer spatial resolution. In Andersson et al. (2014) three fast neutron counters and a D–T generator were used to perform attenuation measurements of test phantoms. Based on the axisymmetry of the test phantoms, the single-projection information was used to calculate radial attenuation distributions of the object, which was compared with the known geometry. In this paper a fast-neutron tomography system based on an array of individual detectors and a purpose-designed compact D–D neutron generator is presented. Each of the 88 detectors consists of a plastic scintillator read out by two Silicon photomultipliers and a dedicated pulse-processing board. Data acquisition for all channels was handled by four single-board microcontrollers. Details of the individual detector design and testing are elaborated upon. Using the complete array, several fast-neutron images of test phantoms were reconstructed, one of which was compared with results using a Co-60 gamma source. The system was shown to be capable of 2 mm resolution, with exposure times on the order of several hours per reconstructed tomogram. Details about these measurements and the analysis of the reconstructed images are given, along with a

Design and simulation of an optimized e-linac based neutron source for BNCT research

International Nuclear Information System (INIS)

Durisi, E.; Alikaniotis, K.; Borla, O.; Bragato, F.; Costa, M.; Giannini, G.; Monti, V.; Visca, L.; Vivaldo, G.; Zanini, A.

2015-01-01

The paper is focused on the study of a novel photo-neutron source for BNCT preclinical research based on medical electron Linacs. Previous studies by the authors already demonstrated the possibility to obtain a mixed thermal and epithermal neutron flux of the order of 10"7 cm"−"2 s"−"1. This paper investigates possible Linac’s modifications and a new photo-converter design to rise the neutron flux above 5 10"7 cm"−"2 s"−"1, also reducing the gamma contamination. - Highlights: • Proposal of a mixed thermal and epithermal (named hyperthermal) neutron source based on medical high energy electron Linac. • Photo-neutron production via Giant Dipole Resonance on high Z materials. • MCNP4B-GN simulations to design the photo-converter geometry maximizing the hyperthermal neutron flux and minimizing the fast neutron and gamma contaminations. Hyperthermal neutron field suitable for BNCT preclinical research.

Block-Based Compressed Sensing for Neutron Radiation Image Using WDFB

Directory of Open Access Journals (Sweden)

Wei Jin

2015-01-01

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

New neutron-based isotopic analytical methods; An explorative study of resonance capture and incoherent scattering

NARCIS (Netherlands)

Perego, R.C.

2004-01-01

Two novel neutron-based analytical techniques have been treated in this thesis, Neutron Resonance Capture Analysis (NRCA), employing a pulsed neutron source, and Neutron Incoherent Scattering (NIS), making use of a cold neutron source. With the NRCA method isotopes are identified by the

Spectral correction factors for conventional neutron dose meters used in high-energy neutron environments improved and extended results based on a complete survey of all neutron spectra in IAEA-TRS-403

International Nuclear Information System (INIS)

Oparaji, U.; Tsai, Y. H.; Liu, Y. C.; Lee, K. W.; Patelli, E.; Sheu, R. J.

2017-01-01

This paper presents improved and extended results of our previous study on corrections for conventional neutron dose meters used in environments with high-energy neutrons (E n > 10 MeV). Conventional moderated-type neutron dose meters tend to underestimate the dose contribution of high-energy neutrons because of the opposite trends of dose conversion coefficients and detection efficiencies as the neutron energy increases. A practical correction scheme was proposed based on analysis of hundreds of neutron spectra in the IAEA-TRS-403 report. By comparing 252 Cf-calibrated dose responses with reference values derived from fluence-to-dose conversion coefficients, this study provides recommendations for neutron field characterization and the corresponding dose correction factors. Further sensitivity studies confirm the appropriateness of the proposed scheme and indicate that (1) the spectral correction factors are nearly independent of the selection of three commonly used calibration sources: 252 Cf, 241 Am-Be and 239 Pu-Be; (2) the derived correction factors for Bonner spheres of various sizes (6''-9'') are similar in trend and (3) practical high-energy neutron indexes based on measurements can be established to facilitate the application of these correction factors in workplaces. (authors)

Introduction to Neutron Coincidence Counter Design Based on Boron-10

Energy Technology Data Exchange (ETDEWEB)

Kouzes, Richard T.; Ely, James H.; Lintereur, Azaree T.; Siciliano, Edward R.

2012-01-22

The Department of Energy Office of Nonproliferation Policy (NA-241) is supporting the project 'Coincidence Counting With Boron-Based Alternative Neutron Detection Technology' at Pacific Northwest National Laboratory (PNNL) for development of an alternative neutron coincidence counter. The goal of this project is ultimately to design, build and demonstrate a boron-lined proportional tube based alternative system in the configuration of a coincidence counter. This report, providing background information for this project, is the deliverable under Task 1 of the project.

Designing research of fast neutron radiation field based on the reactor

International Nuclear Information System (INIS)

Zhang Wenzhong; Zhang Xiaomin

2009-01-01

Based on the Tsinghua University experimental nuclear reactor neutron source, this research designed moderate theory technical scheme, and the thickness of materials in the scheme were selected by means of Monte Carlo simulating method. An fast neutron radiation field was gained. (authors)

Neutron Sources for Standard-Based Testing

Energy Technology Data Exchange (ETDEWEB)

Radev, Radoslav [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McLean, Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2014-11-10

The DHS TC Standards and the consensus ANSI Standards use ²⁵²Cf as the neutron source for performance testing because its energy spectrum is similar to the ²³⁵U and ²³⁹Pu fission sources used in nuclear weapons. An emission rate of 20,000 ± 20% neutrons per second is used for testing of the radiological requirements both in the ANSI standards and the TCS. Determination of the accurate neutron emission rate of the test source is important for maintaining consistency and agreement between testing results obtained at different testing facilities. Several characteristics in the manufacture and the decay of the source need to be understood and accounted for in order to make an accurate measurement of the performance of the neutron detection instrument. Additionally, neutron response characteristics of the particular instrument need to be known and taken into account as well as neutron scattering in the testing environment.

Development and characterization of a neutron detector based on a lithium glass–polymer composite

Energy Technology Data Exchange (ETDEWEB)

Mayer, M.; Nattress, J.; Kukharev, V.; Foster, A.; Meddeb, A. [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); Ounaies, Z. [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Jovanovic, I., E-mail: ijovanovic@psu.edu [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States)

2015-06-11

We report on the fabrication and characterization of a neutron scintillation detector based on a Li-glass–polymer composite that utilizes a combination of pulse height and pulse shape discrimination (PSD) to achieve high gamma rejection. In contrast to fast neutron detection in a PSD medium, we combine two scintillating materials that do not possess inherent neutron/gamma PSD properties to achieve effective PSD/pulse height discrimination in a composite material. Unlike recoil-based fast neutron detection, neutron/gamma discrimination can be robust even at low neutron energies due to the high Q-value neutron capture on {sup 6}Li. A cylindrical detector with a 5.05 cm diameter and 5.08 cm height was fabricated from scintillating 1 mm diameter Li-glass rods and scintillating polyvinyltoluene. The intrinsic efficiency for incident fission neutrons from {sup 252}Cf and gamma rejection of the detector were measured to be 0.33% and less than 10{sup −8}, respectively. These results demonstrate the high selectivity of the detector for neutrons and provide motivation for prototyping larger detectors optimized for specific applications, such as detection and event-by-event spectrometry of neutrons produced by fission.

Development Of A Method For Measurement Of Total Neutron Cross Sections Based On The Neutron Transmission Method Using A He-3 Counter On Filtered Neutron Beams At Dalat Research Reactor

International Nuclear Information System (INIS)

Tran Tuan Anh; Dang Lanh; Nguyen Canh Hai; Nguyen Xuan Hai; Pham Kien; Nguyen Thuy Nham; Pham Ngoc Son; Ho Huu Thang

2007-01-01

Determination of total neutron cross sections and average resonance parameters in the energy range from tens keV to hundreds keV is important for fast reactors calculations and designs because this energy range gives the most output of all neutron induced reactions in the spectrum of fast reactors. Besides, the total neutron cross section measurement is also one of the methods for determination of s, p and d-wave neutron strength functions. The purpose of this project is to develop a method for measurement of total neutron cross sections based on the neutron transmission technique using a He-3 counter. The average total neutron cross sections of 238 U were obtained from neutron transmission measurements on filtered neutron beams of 55 keV and 144 keV at the horizontal channel No.4 of the Dalat research reactor. The present results have been compared with the previous measurements, and the evaluated data from ENDF/B-6.8 library. (author)

Determination of the spatial response of neutron based analysers using a Monte Carlo based method

International Nuclear Information System (INIS)

Tickner, James

2000-01-01

One of the principal advantages of using thermal neutron capture (TNC, also called prompt gamma neutron activation analysis or PGNAA) or neutron inelastic scattering (NIS) techniques for measuring elemental composition is the high penetrating power of both the incident neutrons and the resultant gamma-rays, which means that large sample volumes can be interrogated. Gauges based on these techniques are widely used in the mineral industry for on-line determination of the composition of bulk samples. However, attenuation of both neutrons and gamma-rays in the sample and geometric (source/detector distance) effects typically result in certain parts of the sample contributing more to the measured composition than others. In turn, this introduces errors in the determination of the composition of inhomogeneous samples. This paper discusses a combined Monte Carlo/analytical method for estimating the spatial response of a neutron gauge. Neutron propagation is handled using a Monte Carlo technique which allows an arbitrarily complex neutron source and gauge geometry to be specified. Gamma-ray production and detection is calculated analytically which leads to a dramatic increase in the efficiency of the method. As an example, the method is used to study ways of reducing the spatial sensitivity of on-belt composition measurements of cement raw meal

Using Neutron-based techniques to investigate battery behaviour

International Nuclear Information System (INIS)

Pramudita, James C.; Goonetilleke, Damien; Sharma, Neeraj; Peterson, Vanessa K.

2016-01-01

The extensive use of portable electronic devices has given rise to increasing demand for reliable high energy density storage in the form of batteries. Today, lithium-ion batteries (LIBs) are the leading technology as they offer high energy density and relatively long lifetimes. Despite their widespread adoption, Li-ion batteries still suffer from significant degradation in their performance over time. The most obvious degradation in lithium-ion battery performance is capacity fade – where the capacity of the battery reduces after extended cycling. This talk will focus on how in situ time-resolved neutron powder diffraction (NPD) can be used to gain a better understanding of the structural changes which contribute to the observed capacity fade. The commercial batteries studied each feature different electrochemical and storage histories that are precisely known, allowing us to elucidate the tell-tale signs of battery degradation using NPD and relate these to battery history. Moreover, this talk will also showcase the diverse use of other neutron-based techniques such as neutron imaging to study electrolyte concentrations in lead-acid batteries, and the use of quasi-elastic neutron scattering to study Na-ion dynamics in sodium-ion batteries.

Liquid Li based neutron source for BNCT and science application

International Nuclear Information System (INIS)

Horiike, H.; Murata, I.; Iida, T.; Yoshihashi, S.; Hoashi, E.; Kato, I.; Hashimoto, N.; Kuri, S.; Oshiro, S.

2015-01-01

Liquid lithium (Li) is a candidate material for a target of intense neutron source, heat transfer medium in space engines and charges stripper. For a medical application of BNCT, epithermal neutrons with least energetic neutrons and γ-ray are required so as to avoid unnecessary doses to a patient. This is enabled by lithium target irradiated by protons at 2.5 MeV range, with utilizing the threshold reaction of "7Li(p,n)"7Be at 1.88 MeV. In the system, protons at 2.5 MeV penetrate into Li layer by 0.25 mm with dissipating heat load near the surface. To handle it, thin film flow of high velocity is important for stable operation. For the proton accelerator, electrostatic type of the Schnkel or the tandem is planned to be employed. Neutrons generated at 0.6 MeV are gently moderated to epithermal energy while suppressing accompanying γ-ray minimum by the dedicated moderator assembly. - Highlights: • Liquid lithium (Li) is a candidate material for a target of intense neutron source. • An accelerator based neutron source with p-liquid Li target for boron neutron capture therapy is under development in Osaka University, Japan. • In our system, the harmful radiation dose due to rays and fast neutrons will be suppressed very low. • The system performance are very promising as a state of art cancer treatment system. • The project is planned as a joint undertaking between industries and Osaka University.

Accelerator-based cold neutron sources and their cooling system

International Nuclear Information System (INIS)

Inoue, Kazuhiko; Yanai, Masayoshi; Ishikawa, Yoshikazu.

1985-01-01

We have developed and installed two accelerator-based cold neutron sources within a electron linac at Hokkaido University and a proton synchrotoron at National Laboratory for High Energy Physics. Solid methane at 20K was adopted as the cold moderator. The methane condensing heat exchangers attached directly to the moderator chambers were cooled by helium gas, which was kept cooled in refrigerators and circulated by ventilation fans. Two cold neutron sources have operated smoothly and safely for the past several years. In this paper we describe some of the results obtained in the preliminary experiments by using a modest capacity refrigerator, the design philosophy of the cooling system for the pulsed cold neutron sources, and outline of two facilities. (author)

Development of a Fresnel lens for cold neutrons based on neutron refractive optics

International Nuclear Information System (INIS)

Oku, T.; Morita, S.; Moriyasu, S.; Yamagata, Y.; Ohmori, H.; Takizawa, Y.; Shimizu, H.M.; Hirota, T.; Kiyanagi, Y.; Ino, T.; Furusaka, M.; Suzuki, J.

2001-01-01

We have developed compound refractive lenses (CRLs) for cold neutrons, which are made of vitreous silica and have an effective potential of (90.1-2.7x10 -4 i) neV. In the case of compound refractive optics, neutron absorption by the material deteriorates lens performance. Thus, to prevent an increase in neutron absorption with increasing beam size, we have developed Fresnel lenses using the electrolytic in-process dressing grinding technique. The lens characteristics were carefully investigated with experimental and numerical simulation studies. The lenses functioned as a neutron focusing lens, and the focal length of 14 m was obtained with a 44-element series of the Fresnel lenses for 10 A neutrons. Moreover, good neutron transmission of 0.65 for 15 A neutrons was obtained due to the shape effect. According to comprehensive analysis of the obtained results, it is possible to realize a CRL for practical use by choosing a suitable lens shape and material

Development of a Fresnel lens for cold neutrons based on neutron refractive optics

CERN Document Server

Oku, T; Moriyasu, S; Yamagata, Y; Ohmori, H; Takizawa, Y; Shimizu, H M; Hirota, T; Kiyanagi, Y; Ino, T; Furusaka, M; Suzuki, J

2001-01-01

We have developed compound refractive lenses (CRLs) for cold neutrons, which are made of vitreous silica and have an effective potential of (90.1-2.7x10 sup - sup 4 i) neV. In the case of compound refractive optics, neutron absorption by the material deteriorates lens performance. Thus, to prevent an increase in neutron absorption with increasing beam size, we have developed Fresnel lenses using the electrolytic in-process dressing grinding technique. The lens characteristics were carefully investigated with experimental and numerical simulation studies. The lenses functioned as a neutron focusing lens, and the focal length of 14 m was obtained with a 44-element series of the Fresnel lenses for 10 A neutrons. Moreover, good neutron transmission of 0.65 for 15 A neutrons was obtained due to the shape effect. According to comprehensive analysis of the obtained results, it is possible to realize a CRL for practical use by choosing a suitable lens shape and material.

Electronic neutron sensor based on coincidence detection

International Nuclear Information System (INIS)

Barelaud, B.; Decossas, J.L.; Mokhtari, F.; Vareille, J.C.

1996-01-01

The last symposium on neutron dosimetry which took place in Paris in November 1995 have shown again that it doesn't exist any individual active neutron dosemeter. The state of art on electronic device, the needs of the nuclear power industry in individual neutron monitoring and the new trends of The last symposium on neutron dosimetry which took place in Paris in November 1995 have shown again that it doesn't exist any individual active neutron dosemeter. The state of art on electronic device, the needs of the nuclear power industry in individual neutron monitoring and the new trends of researches were presented. They confirm the relevance of our studies in progress in the C2M team of the University of Limoges. The aim of this work is to realize an individual electronic neutron dosemeter. The device in the progress of being development will operate either as a dosemeter or as ratemeter giving H p (10) and H p (10) either as a spectrometer permitting to characterize the primary neutron beam. (author)

Dosimetric response evaluation of tooth enamel for accelerator-based neutron radiation

International Nuclear Information System (INIS)

Khan, R.F.H.; Rink, W.J.; Boreham, D.R.

2003-01-01

To study the neutron response of human tooth enamel, a number of experiments with an accelerator-based neutron source have been designed. The neutron beam was produced with the low gamma yield, 7 Li(p,n) 7 Be type thick target, using the 3 MV McMaster K.N. Van de Graaff accelerator. The dosimetry was done using a pre-calibrated snoopy type neutron dosimeter. Neutron irradiation induces a dosimetric signal in the tooth enamel at the same defect site as gamma produced damage with the same g-values (g parallel =1.9973, width 0.4 mT g perpendicular =2.002, width 0.3 mT). The dosimetric signal grows linearly with neutron dose from 6-35 Gy tissue dose. Dosimetric response in two different grain sizes (300-500 μm, and grains <4 mm) has shown increased dosimetric amplitude in the larger grains. Dose build up effect on tooth inside the mouth due to cheek was simulated by placing a 4 mm thick paraffin wax layer between the beam and tooth, but had little effect. These results show that for mean neutron energy of 280 keV, the relative neutron response of the human tooth enamel ranges from 8% to 12% of the equivalent gamma ray response

A D-D/D-T fusion reaction based neutron generator system for liver tumor BNCT

International Nuclear Information System (INIS)

Koivunoro, H.; Lou, T.P.; Leung, K. N.; Reijonen, J.

2003-01-01

Boron-neutron capture therapy (BNCT) is an experimental radiation treatment modality used for highly malignant tumor treatments. Prior to irradiation with low energetic neutrons, a 10B compound is located selectively in the tumor cells. The effect of the treatment is based on the high LET radiation released in the 10 B(n,α) 7 Li reaction with thermal neutrons. BNCT has been used experimentally for brain tumor and melanoma treatments. Lately applications of other severe tumor type treatments have been introduced. Results have shown that liver tumors can also be treated by BNCT. At Lawrence Berkeley National Laboratory, various compact neutron generators based on D-D or D-T fusion reactions are being developed. The earlier theoretical studies of the D-D or D-T fusion reaction based neutron generators have shown that the optimal moderator and reflector configuration for brain tumor BNCT can be created. In this work, the applicability of 2.5 MeV neutrons for liver tumor BNCT application was studied. The optimal neutron energy for external liver treatments is not known. Neutron beams of different energies (1eV < E < 100 keV) were simulated and the dose distribution in the liver was calculated with the MCNP simulation code. In order to obtain the optimal neutron energy spectrum with the D-D neutrons, various moderator designs were performed using MCNP simulations. In this article the neutron spectrum and the optimized beam shaping assembly for liver tumor treatments is presented

Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

1990-09-01

Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

Simulations of Lithium-Based Neutron Coincidence Counter for Gd-Loaded Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowles, Christian C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kouzes, Richard T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Siciliano, Edward R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2014-10-01

The Department of Energy Office of Nuclear Safeguards and Security (NA-241) is supporting the project Lithium-Based Alternative Neutron Detection Technology Coincidence Counting for Gd-loaded Fuels at Pacific Northwest National Laboratory for the development of a lithium-based neutron coincidence counter for nondestructively assaying Gd loaded nuclear fuel. This report provides results from MCNP simulations of a lithium-based coincidence counter for the possible measurement of Gd-loaded nuclear fuel. A comparison of lithium-based simulations and UNCL-II simulations with and without Gd loaded fuel is provided. A lithium-based model, referred to as PLNS3A-R1, showed strong promise for assaying Gd loaded fuel.

Spectrum shaping assessment of accelerator-based fusion neutron sources to be used in BNCT treatment

Science.gov (United States)

Cerullo, N.; Esposito, J.; Daquino, G. G.

2004-01-01

Monte Carlo modelling of an irradiation facility, for boron neutron capture therapy (BNCT) application, using a set of advanced type, accelerator based, 3H(d,n) 4He (D-T) fusion neutron source device is presented. Some general issues concerning the design of a proper irradiation beam shaping assembly, based on very hard energy neutron source spectrum, are reviewed. The facility here proposed, which represents an interesting solution compared to the much more investigated Li or Be based accelerator driven neutron source could fulfil all the medical and safety requirements to be used by an hospital environment.

Two reports: (i) Correlation properties of delayed neutrons from fast neutron induced fission. (ii) Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting

International Nuclear Information System (INIS)

Piksaikin, V.M.; Isaev, S.G.; Goverdovski, A.A.; Pshakin, G.M.

1998-10-01

The document includes the following two reports: 'Correlation properties of delayed neutrons from fast neutron induced fission' and 'Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting. A separate abstract was prepared for each report

A feasibility study of a deuterium-deuterium neutron generator-based boron neutron capture therapy system for treatment of brain tumors.

Science.gov (United States)

Hsieh, Mindy; Liu, Yingzi; Mostafaei, Farshad; Poulson, Jean M; Nie, Linda H

2017-02-01

Boron neutron capture therapy (BNCT) is a binary treatment modality that uses high LET particles to achieve tumor cell killing. Deuterium-deuterium (DD) compact neutron generators have advantages over nuclear reactors and large accelerators as the BNCT neutron source, such as their compact size, low cost, and relatively easy installation. The purpose of this study is to design a beam shaping assembly (BSA) for a DD neutron generator and assess the potential of a DD-based BNCT system using Monte Carlo (MC) simulations. The MC model consisted of a head phantom, a DD neutron source, and a BSA. The head phantom had tally cylinders along the centerline for computing neutron and photon fluences and calculating the dose as a function of depth. The head phantom was placed at 4 cm from the BSA. The neutron source was modeled to resemble the source of our current DD neutron generator. A BSA was designed to moderate and shape the 2.45-MeV DD neutrons to the epithermal (0.5 eV to 10 keV) range. The BSA had multiple components, including moderator, reflector, collimator, and filter. Various materials and configurations were tested for each component. Each BSA layout was assessed in terms of the in-air and in-phantom parameters. The maximum brain dose was limited to 12.5 Gray-Equivalent (Gy-Eq) and the skin dose to 18 Gy-Eq. The optimized BSA configuration included 30 cm of lead for reflector, 45 cm of LiF, and 10 cm of MgF 2 for moderator, 10 cm of lead for collimator, and 0.1 mm of cadmium for thermal neutron filter. Epithermal flux at the beam aperture was 1.0 × 10 5  n epi /cm 2 -s; thermal-to-epithermal neutron ratio was 0.05; fast neutron dose per epithermal was 5.5 × 10 -13  Gy-cm 2 /φ epi , and photon dose per epithermal was 2.4 × 10 -13  Gy-cm 2 /φ epi . The AD, AR, and the advantage depth dose rate were 12.1 cm, 3.7, and 3.2 × 10 -3  cGy-Eq/min, respectively. The maximum skin dose was 0.56 Gy-Eq. The DD neutron yield that is needed to

A neutron spectrum unfolding computer code based on artificial neural networks

Science.gov (United States)

Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

2014-02-01

The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in

A neutron spectrometer based on temperature variations in superheated drop compositions

CERN Document Server

Apfel, R E

2002-01-01

The response of superheated drop detectors (SDDs) to neutron radiation varies in a self-consistent manner with variations in temperature and pressure, making such compositions suitable for neutron spectrometry. The advantage of this approach is that the response functions of candidate materials versus energy as the temperature or pressure is varied are nested and have distinct thresholds, with no thermal neutron response. These characteristics permit unfolding without the uncertainties associated with other spectrometry techniques, where multiple solutions are possible, thus requiring an initial guess of the spectrum. A spectrometer was developed based on the well-established technology for acoustic sensing of bubble events interfaced with a proportional-integral-derivative temperature controller. The active monitor for neutrons, called REMbrandt sup T sup M , was used as the platform for controlling temperature on a SDD probe and for data acquisition, thereby automating the process of measuring the neutron e...

Development of highly effective neutron shields and neutron absorbing materials

International Nuclear Information System (INIS)

Tsuda, K.; Matsuda, F.; Taniuchi, H.; Yuhara, T.; Iida, T.

1993-01-01

A wide range of materials, including polymers and hydrogen-occluded alloys that might be usable as the neutron shielding material were examined. And a wide range of materials, including aluminum alloys that might be usable as the neutron-absorbing material were examined. After screening, the candidate material was determined on the basis of evaluation regarding its adaptabilities as a high-performance neutron-shielding and neutron-absorbing material. This candidate material was manufactured for trial, after which material properties tests, neutron-shielding tests and neutron-absorbing tests were carried out on it. The specifications of this material were thus determined. This research has resulted in materials of good performance; a neutron-shielding material based on ethylene propylene rubber and titanium hydride, and a neutron-absorbing material based on aluminum and titanium hydride. (author)

Application of the multisphere technique. Calibration and use of a modified Multiple Probe Detector

International Nuclear Information System (INIS)

Lalande, R.

1966-11-01

The study concerns the search for a portable, compact device with a great autonomy of operation, able to carry out precise measurements of fast neutrons in exclusion zones. A DSM-type multi-probe detector, which is self-contained and fully transistorized, have been studied; it includes a storage battery with a 30 hour autonomy and a buffering capability, a pulse amplifier, an integrator (sensitivity 4 c / s - 200 c / s - 2000 c / s), a totalizer to carry out counting on 5 mm, and a SNR fast neutron probe equipped with its preamplifier. Slightly modified, this device perfectly fulfills the operating conditions. Designed to precisely define the relationship between the flow and the dose intensity, it allows to calibrate any type of fast neutrons detector (e.g. BF 3 or unmodified DSM) that will respond correctly and will provide routine monitoring at a facility

Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination.

Science.gov (United States)

Liu, B; Xu, J; Liu, T; Ouyang, X

2012-10-01

To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a (252)Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D-D neutron generator can create neutrons at up to 10(13) n s(-1) with current technology. All these enable an effective and low-cost method of killing anthrax spores. There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g (252)Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D-D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D-D neutron generator output >10(13) n s(-1) should be attainable in the near future. This indicates that we could use a D-D neutron generator to sterilise anthrax contamination within several seconds.

Analysis of accelerator based neutron spectra for BNCT using proton recoil spectroscopy

International Nuclear Information System (INIS)

Wielopolski, L.; Ludewig, H.; Powell, J.R.; Raparia, D.; Alessi, J.G.; Lowenstein, D.I.

1998-01-01

Boron Neutron Capture Therapy (BNCT) is a promising binary treatment modality for high-grade primary brain tumors (glioblastoma multiforme, GM) and other cancers. BNCT employs a boron-10 containing compound that preferentially accumulates in the cancer cells in the brain. Upon neutron capture by 10 B energetic alpha particles and triton released at the absorption site kill the cancer cell. In order to gain penetration depth in the brain Fairchild proposed, for this purpose, the use of energetic epithermal neutrons at about 10 keV. Phase I/II clinical trials of BNCT for GM are underway at the Brookhaven Medical Research Reactor (BMRR) and at the MIT Reactor, using these nuclear reactors as the source for epithermal neutrons. In light of the limitations of new reactor installations, e.g. cost, safety and licensing, and limited capability for modulating the reactor based neutron beam energy spectra alternative neutron sources are being contemplated for wider implementation of this modality in a hospital environment. For example, accelerator based neutron sources offer the possibility of tailoring the neutron beams, in terms of improved depth-dose distributions, to the individual and offer, with relative ease, the capability of modifying the neutron beam energy and port size. In previous work new concepts for compact accelerator/target configuration were published. In this work, using the Van de Graaff accelerator the authors have explored different materials for filtering and reflecting neutron beams produced by irradiating a thick Li target with 1.8 to 2.5 MeV proton beams. However, since the yield and the maximum neutron energy emerging from the Li-7(p,n)Be-7 reaction increase with increase in the proton beam energy, there is a need for optimization of the proton energy versus filter and shielding requirements to obtain the desired epithermal neutron beam. The MCNP-4A computer code was used for the initial design studies that were verified with benchmark experiments

Study on the novel neutron-to-proton convertor for improving the detection efficiency of a triple GEM based fast neutron detector

International Nuclear Information System (INIS)

Wang Xiaodong; Yang Lei; Zhang Chunhui; Hu Bitao; Yang Herun; Zhang Junwei; Ren Zhongguo; Ha Ri-Ba-La; An Luxing

2015-01-01

A high-efficiency fast neutron detector prototype based on a triple Gas Electron Multiplier (GEM) detector, which, coupled with a novel multi-layered high-density polyethylene (HDPE) as a neutron-to-proton converter for improving the neutron detection efficiency, is introduced and tested with the Am-Be neutron source in the Institute of Modern Physics (IMP) at Lanzhou in the present work. First, the developed triple GEM detector is tested by measuring its effective gain and energy resolution with "5"5Fe X-ray source to ensure that it has a good performance. The effective gain and obtained energy resolution is 5.0 × 10"4 and around 19.2%, respectively. Secondly, the novel multi-layered HDPE converter is coupled with the cathode of the triple GEM detector making it a high-efficiency fast neutron detector. Its effective neutron response is four times higher than that of the traditional single-layered conversion technique when the converter layer number is 38. (authors)

Spectrum unfolding, sensitivity analysis and propagation of uncertainties with the maximum entropy deconvolution code MAXED

CERN Document Server

Reginatto, M; Neumann, S

2002-01-01

MAXED was developed to apply the maximum entropy principle to the unfolding of neutron spectrometric measurements. The approach followed in MAXED has several features that make it attractive: it permits inclusion of a priori information in a well-defined and mathematically consistent way, the algorithm used to derive the solution spectrum is not ad hoc (it can be justified on the basis of arguments that originate in information theory), and the solution spectrum is a non-negative function that can be written in closed form. This last feature permits the use of standard methods for the sensitivity analysis and propagation of uncertainties of MAXED solution spectra. We illustrate its use with unfoldings of NE 213 scintillation detector measurements of photon calibration spectra, and of multisphere neutron spectrometer measurements of cosmic-ray induced neutrons at high altitude (approx 20 km) in the atmosphere.

TEPC performance in the CANDU workplace

International Nuclear Information System (INIS)

Waker, A.J.; Szornel, K.; Nunes, J.

1997-01-01

Tissue-equivalent proportional counters (TEPCs) have a number of features that make them an attractive option for neutron monitoring around power reactors. These features are principally the ability of the TEPC to operate in a mixed field environment, the direct determination of the dose equivalent from first principles and a reasonably well-understood response to the wide range of neutron energy encountered. A spherical TEPC from Far West Technology (5'' in diameter) and a commercial, TEPC-based, neutron monitor (REM 500) have been used to map the neutron fields at different locations in a CANDU 6 (Canadian Deuterium Uranium) power reactor operated by New Brunswick Power. Neutron ambient dose equivalent rates ranged between 8 μSv.h -1 and 500 μSv.h -1 as measured with a multisphere spectrometer and photon dose equivalent rates from 40 μSv.h -1 to 1.2 mSv.H -1 as determined with a TEPC. It is shown that, for the CANDU workplace, the energy response of TEPCs is still not known accurately enough to enable field correction factors to be derived from neutron fluence measurements or to avoid the the need for more appropriate calibration procedures based on simulated 'workplace' neutron fields. The sensitivity of TEPCs is sufficient for workplace monitoring, but needs consideration when counters of 2'' diameter or smaller are proposed for use. For mixed-field work, the use of a TEPC for photon dosimetry as opposed to photon discrimination requires attention to count rates and instrument dead time. (author)

Response of LET spectrometer based on track etching at some neutron sources

International Nuclear Information System (INIS)

Spurny, Frantisek; Brabcova, Katerina; Jadrnickova, Iva

2008-01-01

There is still need to develop upgrade, and test further methods able to characterise the external exposure to neutrons. This contribution presents further results obtained with the goal to enlarge and upgrade the possibility of neutron dosimetry and microdosimetry with a LET spectrometer based on the chemically etched track detectors (TED). As TED we have used several types of polyallyldiglycolcarbonates (PADC). The PADC detectors have been exposed in: high energy neutron beams at iThemba facility, Cape Town, South Africa, and in monoenergetic neutron beams at JRC Geel, Belgium. The studies have been performed in the frame of the ESA supported project DOBIES. (author)

A neutron production target for ESS based upon the Canned-rods concept

International Nuclear Information System (INIS)

Ghiglino, A.; Terrón, S.; Thomsen, K.; Wolters, J.; Magán, M.; Martínez, F.; Vicente, P.J. de; Vivanco, R.; Sordo, F.; Butzek, M.; Perlado, J.M.; Bermejo, F.J.

2014-01-01

The neutron production targets operating within the present day spallation neutron sources in the MW power range are either based on water-cooled solid state devices such as that implemented at the SINQ source at PSI or liquid metal loops such as those installed at SNS and MLSF. Here we describe a water-cooled rotating solid target as an option for the 5 MW ESS project as an alternative to the current design based upon a helium-cooled solid rotating target. Implementation of the proposed option would provide comparable neutronic performance to that of the gas-cooled concept and furthermore, it would involve a relatively straightforward adaptation of the current ESS baseline geometry

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)

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)

Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

Science.gov (United States)

Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A

2014-12-01

An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.

Delayed neutron yield from fast neutron induced fission of 238U

International Nuclear Information System (INIS)

Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Roshchenko, V.A.; Goverdovski, A.A.; Tertytchnyi, R.G.

2002-01-01

The measurements of the total delayed neutron yield from fast neutron induced fission of 238 U were made. The experimental method based on the periodic irradiation of the fissionable sample by neutrons from a suitable nuclear reaction had been employed. The preliminary results on the energy dependence of the total delayed neutron yield from fission of 238 U are obtained. According to the comparison of experimental data with our prediction based on correlation properties of delayed neutron characteristics, it is concluded that the value of the total delayed neutron yield near the threshold of (n,f) reaction is not a constant. (author)

Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

Science.gov (United States)

Čufar, Aljaž; Batistoni, Paola; Conroy, Sean; Ghani, Zamir; Lengar, Igor; Milocco, Alberto; Packer, Lee; Pillon, Mario; Popovichev, Sergey; Snoj, Luka; JET Contributors

2017-03-01

At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium-tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle-energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

Energy Technology Data Exchange (ETDEWEB)

Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Conroy, Sean [Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Ghani, Zamir [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lengar, Igor [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Milocco, Alberto; Packer, Lee [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pillon, Mario [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2017-03-01

At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium–tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle–energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

PELAN - a transportable, neutron-based UXO identification technique

International Nuclear Information System (INIS)

Vourvopoulos, G.

1998-01-01

An elemental characterization method is used to differentiate between inert projectiles and UXO's. This method identifies in a non-intrusive, nondestructive manner, the elemental composition of the projectile contents. Most major and minor chemical elements within the interrogated object (hydrogen, carbon, nitrogen, oxygen, fluorine, phosphorus, chlorine, arsenic, etc.) are identified and quantified. The method is based on PELAN - Pulsed Elemental Analysis with Neutrons. PELAN uses pulsed neutrons produced from a compact, sealed tube neutron generator. Using an automatic analysis computer program, the quantities of each major and minor chemical element are determined. A decision-making tree identifies the object by comparing its elemental composition with stored elemental composition libraries of substances that could be contained within the projectile. In a series of blind tests, PELAN was able to identify without failure, the contents of each shell placed in front of it. The PELAN probe does not need to be in contact with the interrogated projectile. If the object is buried, the interrogation can take place in situ provided the probe can be inserted a few centimeters from the object's surface. (author)

Cr-39 fast neutron dosemeter based on A (n, α) converter

International Nuclear Information System (INIS)

Widayati, S.; Budiantari, T.

1998-01-01

The aim of this experiment is to obtained the response of Cr-39 as fast neutron dosemeter based on an (n, α) converter. Cr-39 was irradiated to AmBe fast neutron flux from 0.10 mSv to 2.5 mSv. Cr-39 processed by chemical etching with NaOH 20 % at temperature of 60 oC in six hours. The results of experiment showed that the response of Cr-39 based on an (n, α) converter is 6 times bigger than the response of Cr-39 without (n, α) converter. (author)

ICF implosion hotspot ion temperature diagnostic techniques based on neutron time-of-flight method

International Nuclear Information System (INIS)

Tang Qi; Song Zifeng; Chen Jiabin; Zhan Xiayu

2013-01-01

Ion temperature of implosion hotspot is a very important parameter for inertial confinement fusion. It reflects the energy level of the hotspot, and it is very sensitive to implosion symmetry and implosion speed. ICF implosion hotspot ion temperature diagnostic techniques based on neutron time-of-flight method were described. A neutron TOF spectrometer was developed using a ultrafast plastic scintillator as the neutron detector. Time response of the spectrometer has 1.1 ns FWHM and 0.5 ns rising time. TOF spectrum resolving method based on deconvolution and low pass filter was illuminated. Implosion hotspot ion temperature in low neutron yield and low ion temperature condition at Shenguang-Ⅲ facility was acquired using the diagnostic techniques. (authors)

Extraction of the neutron-neutron scattering length ann from kinematically complete neutron-deuteron breakup experiments

International Nuclear Information System (INIS)

Witala, H.; Hueber, D.; Gloeckle, W.; Tornow, W.; Gonzalez Trotter, D.E.

1996-01-01

Data for the neutron-neutron final-state-interaction cross section obtained recently in a kinematically complete neutron-deuteron breakup experiment have been reanalyzed using rigorous solutions of the three-nucleon Faddeev equations with realistic nucleon-nucleon interactions. A discrepancy was found with respect to a recent analysis based on the W-matrix approximation to the Paris potential. We also estimate theoretical uncertainties in extracting the neutron-neutron scattering length resulting from the use of different nucleon-nucleon interactions and the possible action of the two pion-exchange three-nucleon force. We find that there exists a certain production angle for the interacting neutron-neutron pair where the uncertainties become minimal. (author)

Characterisation of an accelerator-based neutron source for BNCT versus beam energy

Science.gov (United States)

Agosteo, S.; Curzio, G.; d'Errico, F.; Nath, R.; Tinti, R.

2002-01-01

Neutron capture in 10B produces energetic alpha particles that have a high linear energy transfer in tissue. This results in higher cell killing and a higher relative biological effectiveness compared to photons. Using suitably designed boron compounds which preferentially localize in cancerous cells instead of healthy tissues, boron neutron capture therapy (BNCT) has the potential of providing a higher tumor cure rate within minimal toxicity to normal tissues. This clinical approach requires a thermal neutron source, generally a nuclear reactor, with a fluence rate sufficient to deliver tumorcidal doses within a reasonable treatment time (minutes). Thermal neutrons do not penetrate deeply in tissue, therefore BNCT is limited to lesions which are either superficial or otherwise accessible. In this work, we investigate the feasibility of an accelerator-based thermal neutron source for the BNCT of skin melanomas. The source was designed via MCNP Monte Carlo simulations of the thermalization of a fast neutron beam, generated by 7 MeV deuterons impinging on a thick target of beryllium. The neutron field was characterized at several deuteron energies (3.0-6.5 MeV) in an experimental structure installed at the Van De Graaff accelerator of the Laboratori Nazionali di Legnaro, in Italy. Thermal and epithermal neutron fluences were measured with activation techniques and fast neutron spectra were determined with superheated drop detectors (SDD). These neutron spectrometry and dosimetry studies indicated that the fast neutron dose is unacceptably high in the current design. Modifications to the current design to overcome this problem are presented.

On the e-linac-based neutron yield

International Nuclear Information System (INIS)

Bunatyan, G.G.; Nikolenko, V.G.; Popov, A.B.

2010-01-01

We treat neutron generating in high atomic number materials due to the photonuclear reactions induced by the Bremsstrahlung of an electron beam produced by linear electron accelerator (e-linac). The dependence of neutron yield on the electron energy and the irradiated sample size is considered for various sample materials. The calculations are performed without resort to the so-called 'numerical Monte Carlo simulation'. The acquired neutron yields are well correlated with the data asserted in investigations performed at a number of the e-linac-driven neutron sources

A research plan based on high intensity proton accelerator Neutron Science Research Center

International Nuclear Information System (INIS)

Mizumoto, Motoharu

1997-01-01

A plan called Neutron Science Research Center (NSRC) has been proposed in JAERI. The center is a complex composed of research facilities based on a proton linac with an energy of 1.5GeV and an average current of 10mA. The research facilities will consist of Thermal/Cold Neutron Facility, Neutron Irradiation Facility, Neutron Physics Facility, OMEGA/Nuclear Energy Facility, Spallation RI Beam Facility, Meson/Muon Facility and Medium Energy Experiment Facility, where high intensity proton beam and secondary particle beams such as neutron, pion, muon and unstable radio isotope (RI) beams generated from the proton beam will be utilized for innovative researches in the fields on nuclear engineering and basic sciences. (author)

A research plan based on high intensity proton accelerator Neutron Science Research Center

Energy Technology Data Exchange (ETDEWEB)

Mizumoto, Motoharu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

A plan called Neutron Science Research Center (NSRC) has been proposed in JAERI. The center is a complex composed of research facilities based on a proton linac with an energy of 1.5GeV and an average current of 10mA. The research facilities will consist of Thermal/Cold Neutron Facility, Neutron Irradiation Facility, Neutron Physics Facility, OMEGA/Nuclear Energy Facility, Spallation RI Beam Facility, Meson/Muon Facility and Medium Energy Experiment Facility, where high intensity proton beam and secondary particle beams such as neutron, pion, muon and unstable radio isotope (RI) beams generated from the proton beam will be utilized for innovative researches in the fields on nuclear engineering and basic sciences. (author)

A novel fast-neutron tomography system based on a plastic scintillator array and a compact D-D neutron generator.

Science.gov (United States)

Adams, Robert; Zboray, Robert; Prasser, Horst-Michael

2016-01-01

Very few experimental imaging studies using a compact neutron generator have been published, and to the knowledge of the authors none have included tomography results using multiple projection angles. Radiography results with a neutron generator, scintillator screen, and camera can be seen in Bogolubov et al. (2005), Cremer et al. (2012), and Li et al. (2014). Comparable results with a position-sensitive photomultiplier tube can be seen in Popov et al. (2011). One study using an array of individual fast neutron detectors in the context of cargo scanning for security purposes is detailed in Eberhardt et al. (2005). In that case, however, the emphasis was on very large objects with a resolution on the order of 1cm, whereas this study focuses on less massive objects and a finer spatial resolution. In Andersson et al. (2014) three fast neutron counters and a D-T generator were used to perform attenuation measurements of test phantoms. Based on the axisymmetry of the test phantoms, the single-projection information was used to calculate radial attenuation distributions of the object, which was compared with the known geometry. In this paper a fast-neutron tomography system based on an array of individual detectors and a purpose-designed compact D-D neutron generator is presented. Each of the 88 detectors consists of a plastic scintillator read out by two Silicon photomultipliers and a dedicated pulse-processing board. Data acquisition for all channels was handled by four single-board microcontrollers. Details of the individual detector design and testing are elaborated upon. Using the complete array, several fast-neutron images of test phantoms were reconstructed, one of which was compared with results using a Co-60 gamma source. The system was shown to be capable of 2mm resolution, with exposure times on the order of several hours per reconstructed tomogram. Details about these measurements and the analysis of the reconstructed images are given, along with a discussion

Neutron shielding material based on colemanite and epoxy resin

International Nuclear Information System (INIS)

Okuno, K.

2005-01-01

In recent years, there has been a need for compact shielding design such as self-shielding of a PET cyclotron or up-gradation of radiation machinery in existing facilities. In these cases, high performance shielding materials are needed. Concrete or polyethylene have been used for a neutron shield. However, for compact shielding, they fall short in terms of performance or durability. Therefore, a new type of neutron shielding material based on epoxy resin and colemanite has been developed. Slab attenuation experiments up to 40 cm for the new shielding material were carried out using a 252 Cf neutron source. Measurement was carried out using a REM-counter, and compared with calculation. The results show that the shielding performance is better than concrete and polyethylene mixed with 10 wt% boron oxide. From the result, we confirmed that the performance of the new material is suitable for practical use. (authors)

Logic based feature detection on incore neutron spectra

Energy Technology Data Exchange (ETDEWEB)

Racz, A.; Kiss, S.; Bende-Farkas, S. (Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics)

1993-04-01

A general framework for detecting features of incore neutron spectra with a rule-based methodology is presented. As an example, we determine the meaningful peaks in the APSD-s. This work is part of a larger project, aimed at developing a noise diagnostic expert system. (Author).

A neutron spectrum unfolding code based on generalized regression artificial neural networks

International Nuclear Information System (INIS)

Ortiz R, J. M.; Martinez B, M. R.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

2015-10-01

The most delicate part of neutron spectrometry, is the unfolding process. Then derivation of the spectral information is not simple because the unknown is not given directly as result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, as the optimum selection of the network topology and the long training time. Compared to BPNN, is usually much faster to train a generalized regression neural network (GRNN). That is mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum. In addition, often are more accurate than BPNN in prediction. These characteristics make GRNN be of great interest in the neutron spectrometry domain. In this work is presented a computational tool based on GRNN, capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. (Author)

A neutron spectrum unfolding code based on generalized regression artificial neural networks

Energy Technology Data Exchange (ETDEWEB)

Ortiz R, J. M.; Martinez B, M. R.; Castaneda M, R.; Solis S, L. O. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Ramon Lopez Velarde 801, Col. Centro, 98000 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

2015-10-15

The most delicate part of neutron spectrometry, is the unfolding process. Then derivation of the spectral information is not simple because the unknown is not given directly as result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, as the optimum selection of the network topology and the long training time. Compared to BPNN, is usually much faster to train a generalized regression neural network (GRNN). That is mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum. In addition, often are more accurate than BPNN in prediction. These characteristics make GRNN be of great interest in the neutron spectrometry domain. In this work is presented a computational tool based on GRNN, capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a {sup 6}LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. (Author)

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.

Evaluation of radioactivity in the bodies of mice induced by neutron exposure from an epi-thermal neutron source of an accelerator-based boron neutron capture therapy system

Science.gov (United States)

NAKAMURA, Satoshi; IMAMICHI, Shoji; MASUMOTO, Kazuyoshi; ITO, Masashi; WAKITA, Akihisa; OKAMOTO, Hiroyuki; NISHIOKA, Shie; IIJIMA, Kotaro; KOBAYASHI, Kazuma; ABE, Yoshihisa; IGAKI, Hiroshi; KURITA, Kazuyoshi; NISHIO, Teiji; MASUTANI, Mitsuko; ITAMI, Jun

2017-01-01

This study aimed to evaluate the residual radioactivity in mice induced by neutron irradiation with an accelerator-based boron neutron capture therapy (BNCT) system using a solid Li target. The radionuclides and their activities were evaluated using a high-purity germanium (HP-Ge) detector. The saturated radioactivity of the irradiated mouse was estimated to assess the radiation protection needs for using the accelerator-based BNCT system. 24Na, 38Cl, 80mBr, 82Br, 56Mn, and 42K were identified, and their saturated radioactivities were (1.4 ± 0.1) × 102, (2.2 ± 0.1) × 101, (3.4 ± 0.4) × 102, 2.8 ± 0.1, 8.0 ± 0.1, and (3.8 ± 0.1) × 101 Bq/g/mA, respectively. The 24Na activation rate at a given neutron fluence was found to be consistent with the value reported from nuclear-reactor-based BNCT experiments. The induced activity of each nuclide can be estimated by entering the saturated activity of each nuclide, sample mass, irradiation time, and proton current into the derived activation equation in our accelerator-based BNCT system. PMID:29225308

Time-frequency feature analysis and recognition of fission neutrons signal based on support vector machine

International Nuclear Information System (INIS)

Jin Jing; Wei Biao; Feng Peng; Tang Yuelin; Zhou Mi

2010-01-01

Based on the interdependent relationship between fission neutrons ( 252 Cf) and fission chain ( 235 U system), the paper presents the time-frequency feature analysis and recognition in fission neutron signal based on support vector machine (SVM) through the analysis on signal characteristics and the measuring principle of the 252 Cf fission neutron signal. The time-frequency characteristics and energy features of the fission neutron signal are extracted by using wavelet decomposition and de-noising wavelet packet decomposition, and then applied to training and classification by means of support vector machine based on statistical learning theory. The results show that, it is effective to obtain features of nuclear signal via wavelet decomposition and de-noising wavelet packet decomposition, and the latter can reflect the internal characteristics of the fission neutron system better. With the training accomplished, the SVM classifier achieves an accuracy rate above 70%, overcoming the lack of training samples, and verifying the effectiveness of the algorithm. (authors)

A Dosimetry Study of Deuterium-Deuterium Neutron Generator-based In Vivo Neutron Activation Analysis.

Science.gov (United States)

Sowers, Daniel; Liu, Yingzi; Mostafaei, Farshad; Blake, Scott; Nie, Linda H

2015-12-01

A neutron irradiation cavity for in vivo neutron activation analysis (IVNAA) to detect manganese, aluminum, and other potentially toxic elements in human hand bone has been designed and its dosimetric specifications measured. The neutron source is a customized deuterium-deuterium neutron generator that produces neutrons at 2.45 MeV by the fusion reaction 2H(d, n)3He at a calculated flux of 7 × 10(8) ± 30% s(-1). A moderator/reflector/shielding [5 cm high density polyethylene (HDPE), 5.3 cm graphite and 5.7 cm borated (HDPE)] assembly has been designed and built to maximize the thermal neutron flux inside the hand irradiation cavity and to reduce the extremity dose and effective dose to the human subject. Lead sheets are used to attenuate bremsstrahlung x rays and activation gammas. A Monte Carlo simulation (MCNP6) was used to model the system and calculate extremity dose. The extremity dose was measured with neutron and photon sensitive film badges and Fuji electronic pocket dosimeters (EPD). The neutron ambient dose outside the shielding was measured by Fuji NSN3, and the photon dose was measured by a Bicron MicroREM scintillator. Neutron extremity dose was calculated to be 32.3 mSv using MCNP6 simulations given a 10-min IVNAA measurement of manganese. Measurements by EPD and film badge indicate hand dose to be 31.7 ± 0.8 mSv for neutrons and 4.2 ± 0.2 mSv for photons for 10 min; whole body effective dose was calculated conservatively to be 0.052 mSv. Experimental values closely match values obtained from MCNP6 simulations. These are acceptable doses to apply the technology for a manganese toxicity study in a human population.

Neutron nuclear physics under the neutron science project

Energy Technology Data Exchange (ETDEWEB)

Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-11-01

The concept of fast neutron physics facility in the Neutron Science Research project is described. This facility makes use of an ultra-short proton pulse (width < 1 ns) for fast neutron time-of-flight works. The current design is based on an assumption of the maximum proton current of 100 {mu}A. Available neutron fluence and energy resolution are explained. Some of the research subjects to be performed at this facility are discussed. (author)

Optimisation of the neutron source based on gas dynamic trap for transmutation of radioactive wastes

Science.gov (United States)

Anikeev, Andrey V.

2012-06-01

The Budker Institute of Nuclear Physics in collaboration with the Russian and foreign organizations develop the project of 14 MeV neutron source, which can be used for fusion material studies and for other application. The projected neutron source of plasma type is based on the plasma Gas Dynamic Trap (GDT), which is a special magnetic mirror system for plasma confinement. Presented work continues the subject of development the GDT-based neutron source (GDT-NS) for hybrid fusion-fission reactors. The paper presents the results of recent numerical optimization of such neutron source for transmutation of the long-lives radioactive wastes in spent nuclear fuel.

Neutron distribution modeling based on integro-probabilistic approach of discrete ordinates method

International Nuclear Information System (INIS)

Khromov, V.V.; Kryuchkov, E.F.; Tikhomirov, G.V.

1992-01-01

In this paper is described the universal nodal method for the neutron distribution calculation in reactor and shielding problems, based on using of influence functions and factors of local-integrated volume and surface neutron sources in phase subregions. This method permits to avoid the limited capabilities of collision-probability method concerning with the detailed calculation of angular neutron flux dependence, scattering anisotropy and empty channels. The proposed method may be considered as modification of S n - method with advantage of ray-effects elimination. There are presented the description of method theory and algorithm following by the examples of method applications for calculation of neutron distribution in three-dimensional model of fusion reactor blanket and in highly heterogeneous reactor with empty channel

Mechanical strength evaluation of the glass base material in the JRR-3 neutron guide tube

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tetsuya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-02-01

The lifetime of the thermal neutron guide tube installed JRR-3 was investigated after 6 years from their first installation. And it was confirmed that a crack had been piercing into the glass base material of the side plate of the neutron guide tube. The cause of the crack was estimated as a static fatigue of the guide tube where an inside of the tube had been evacuated and stressed as well as an embrittlement of the glass base material by gamma ray irradiation. In this report, we evaluate the mechanical strength of the glass base material and estimate the time when the base material gets fatigue fracture. Furthermore, we evaluate a lifetime of the neutron guide tube and confirm the validity of update timing in 2000 and 2001 when the thermal neutron guide tubes T1 and T2 were exchanged into those using the super mirror. (author)

Proposal for an accelerator-based neutron generator

International Nuclear Information System (INIS)

Grand, P.

1975-07-01

An Accelerator-based Neutron Generator is described that consists of a 30-MeV deuteron linear accelerator using a flowing liquid lithium target. With a continuous deuteron current of 100 milliamperes, a source intensity of more than 10 16 neutrons per second will be produced. The neutrons will be emitted in a roughly collimated beam. The proposed facility can be divided into two areas: the 30-MeV linear accelerator and the multiple-target experimental area. The 30-MeV accelerator will consist of eight rf accelerating cavities in a single vacuum tank, each cavity being powered by its own rf power amplifier operating at 50 MHz. To shield the beam bunches from the rf field when it is in the decelerating direction, 66 ''drift tubes'' will be included; the drift-tube structures will include quadrupole magnets which will keep the beam focused. The accelerator will produce a continuous beam of 100 milliamperes. Beam power will thus be 3.0 megawatts; total power including rf losses in the accelerating cavities will be 4.5 megawatts. The injectors for the linear accelerator will be two 500-kV dc accelerators, one for injection of D + ions and the other for D - ions. They can be used simultaneously or one can serve as a spare in case of breakdown or maintenance of the other. (U.S.)

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.

Accelerator-based intense neutron source for materials R and D

International Nuclear Information System (INIS)

Jameson, R.A.

1990-01-01

Accelerator-based neutron sources for R and D of materials in nuclear energy systems, including fusion reactors, can provide sufficient neutron flux, flux-volume, fluence and other attractive features for many aspects of materials research. The neutron spectrum produced from the D-Li reaction has been judged useful for many basic materials research problems, and satisfactory as an approximation of the fusion process. A most interesting aspect for materials researchers is the increased flexibility and opportunities for experimental configurations that a modern accelerator-based source could add to the set of available tools. First, of course, is a high flux of neutrons. Four other tools are described: 1. The output energy of the deuteron beam can be varied to provide energy selectivity for the materials researcher. The energy would typically be varied in discrete steps; the number of steps can be adjusted depending on actual needs and costs. 2. The materials sample target chamber could be irradiated by more than one beam, from different angles. This would provide many possibilities for tailoring the flux distribution. 3. Advanced techniques in magnetic optics systems allow the density distribution of the deuteron beam at the target to be tailored. Controlled distributions from Gaussian to uniform to hollow can be provided. This affords further control of the distribution in the target chamber. 4. The accelerator and associated beam transport elements are all essentially electronic systems and, therefore, can be controlled and modulated on a time cycle basis. Therefore, all of the above tools could be varied in possibly complex patterns under computer control; this may open further experimental approaches for studying various rate-dependent effects. These considerations will be described in the context of the Energy Selective Neutron Irradiation Test (ESNIT) facility which is conceived at JAERI. (author)

Characterisation of an accelerator-based neutron source for BNCT versus beam energy

CERN Document Server

Agosteo, S; D'Errico, F; Nath, R; Tinti, R

2002-01-01

Neutron capture in sup 1 sup 0 B produces energetic alpha particles that have a high linear energy transfer in tissue. This results in higher cell killing and a higher relative biological effectiveness compared to photons. Using suitably designed boron compounds which preferentially localize in cancerous cells instead of healthy tissues, boron neutron capture therapy (BNCT) has the potential of providing a higher tumor cure rate within minimal toxicity to normal tissues. This clinical approach requires a thermal neutron source, generally a nuclear reactor, with a fluence rate sufficient to deliver tumorcidal doses within a reasonable treatment time (minutes). Thermal neutrons do not penetrate deeply in tissue, therefore BNCT is limited to lesions which are either superficial or otherwise accessible. In this work, we investigate the feasibility of an accelerator-based thermal neutron source for the BNCT of skin melanomas. The source was designed via MCNP Monte Carlo simulations of the thermalization of a fast ...

Accelerator-based intense neutron source for materials R ampersand D

International Nuclear Information System (INIS)

Jameson, R.A.

1990-01-01

Accelerator-based neutron sources for R ampersand D of materials in nuclear energy systems, including fusion reactors, can provide sufficient neutron flux, flux-volume, fluence and other attractive features for many aspects of materials research. The neutron spectrum produced from the D-Li reaction has been judged useful for many basic materials research problems, and to be a satisfactory approximation to that of the fusion process. The technology of high-intensity linear accelerators can readily be applied to provide the deuteron beam for the neutron source. Earlier applications included the Los Alamos Meson Physics Facility and the Fusion Materials Irradiation Test facility prototype. The key features of today's advanced accelerator technology are presented to illustrate the present state-of-the-art in terms of improved understanding of basic physical principles and engineering technique, and to show how these advances can be applied to present demands in a timely manner. These features include how to produce an intense beam current with the high quality required to minimize beam losses along the accelerator and transport system that could cause maintenance difficulties, by controlling the beam emittance through proper choice of the operating frequency, balancing of the forces acting on the beam, and realization in practical hardware. A most interesting aspect for materials researchers is the increased flexibility and opportunities for experimental configurations that a modern accelerator-based source could add to the set of available tools. 8 refs., 5 figs

Simulation of e-{gamma}-n targets by FLUKA and measurement of neutron flux at various angles for accelerator based neutron source

Energy Technology Data Exchange (ETDEWEB)

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

2010-10-15

A 6 MeV Race track Microtron (an electron accelerator) based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products where the low neutron flux requirement is desirable. The bremsstrahlung radiation emitted by impinging 6 MeV electron on the e-{gamma} primary target, was made to fall on the {gamma}-n secondary target to produce neutrons. The optimisation of bremsstrahlung and neutron producing target along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium and the measured fluxes were 1.1878 x 10{sup 5}, 0.9403 x 10{sup 5}, 0.7428 x 10{sup 5}, 0.6274 x 10{sup 5}, 0.5659 x 10{sup 5}, 0.5210 x 10{sup 5} n/cm{sup 2}/s at 0{sup o}, 30{sup o}, 60{sup o}, 90{sup o}, 115{sup o}, 140{sup o} respectively. The results indicate that the neutron flux was found to be decreased as increase in the angle and in good agreement with the FLUKA simulation.

Synthetic neutron camera and spectrometer in JET based on AFSI-ASCOT simulations

Science.gov (United States)

Sirén, P.; Varje, J.; Weisen, H.; Koskela, T.; contributors, JET

2017-09-01

The ASCOT Fusion Source Integrator (AFSI) has been used to calculate neutron production rates and spectra corresponding to the JET 19-channel neutron camera (KN3) and the time-of-flight spectrometer (TOFOR) as ideal diagnostics, without detector-related effects. AFSI calculates fusion product distributions in 4D, based on Monte Carlo integration from arbitrary reactant distribution functions. The distribution functions were calculated by the ASCOT Monte Carlo particle orbit following code for thermal, NBI and ICRH particle reactions. Fusion cross-sections were defined based on the Bosch-Hale model and both DD and DT reactions have been included. Neutrons generated by AFSI-ASCOT simulations have already been applied as a neutron source of the Serpent neutron transport code in ITER studies. Additionally, AFSI has been selected to be a main tool as the fusion product generator in the complete analysis calculation chain: ASCOT - AFSI - SERPENT (neutron and gamma transport Monte Carlo code) - APROS (system and power plant modelling code), which encompasses the plasma as an energy source, heat deposition in plant structures as well as cooling and balance-of-plant in DEMO applications and other reactor relevant analyses. This conference paper presents the first results and validation of the AFSI DD fusion model for different auxiliary heating scenarios (NBI, ICRH) with very different fast particle distribution functions. Both calculated quantities (production rates and spectra) have been compared with experimental data from KN3 and synthetic spectrometer data from ControlRoom code. No unexplained differences have been observed. In future work, AFSI will be extended for synthetic gamma diagnostics and additionally, AFSI will be used as part of the neutron transport calculation chain to model real diagnostics instead of ideal synthetic diagnostics for quantitative benchmarking.

Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

International Nuclear Information System (INIS)

Davidson, J.W.; Battat, M.E.

1993-01-01

Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

Neutron flux density data acquisition system based on LabVIEW

International Nuclear Information System (INIS)

Zhao Yanhui; Zhao Xiuliang; Li Zonglun; Liang Fengyan; Liu Liyan

2011-01-01

In the LabVIEW software, combined with PCI-6251 data acquisition card, VI of neutron flux density data acquisition is realized by DAQmx data acquisition functions. VI is composed of front panel and block diagram. The data collected can be displayed in the forms of the data curve and the data control, and saved in the form of files. Test results show that the frequency of output signal in NI ELVIS can be accurately measured by the system, realizing neutron flux density data acquisition based on LabVIEW. (authors)

High-flux neutron source based on a liquid-lithium target

Science.gov (United States)

Halfon, S.; Feinberg, G.; Paul, M.; Arenshtam, A.; Berkovits, D.; Kijel, D.; Nagler, A.; Eliyahu, I.; Silverman, I.

2013-04-01

A prototype compact Liquid Lithium Target (LiLiT), able to constitute an accelerator-based intense neutron source, was built. The neutron source is intended for nuclear astrophysical research, boron neutron capture therapy (BNCT) in hospitals and material studies for fusion reactors. The LiLiT setup is presently being commissioned at Soreq Nuclear research Center (SNRC). The lithium target will produce neutrons through the 7Li(p,n)7Be reaction and it will overcome the major problem of removing the thermal power generated by a high-intensity proton beam, necessary for intense neutron flux for the above applications. The liquid-lithium loop of LiLiT is designed to generate a stable lithium jet at high velocity on a concave supporting wall with free surface toward the incident proton beam (up to 10 kW). During off-line tests, liquid lithium was flown through the loop and generated a stable jet at velocity higher than 5 m/s on the concave supporting wall. The target is now under extensive test program using a high-power electron-gun. Up to 2 kW electron beam was applied on the lithium flow at velocity of 4 m/s without any flow instabilities or excessive evaporation. High-intensity proton beam irradiation will take place at SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator currently in commissioning at SNRC.

Development of lithium target for accelerator based neutron capture therapy

International Nuclear Information System (INIS)

Taskaev, Sergey; Bayanov, Boris; Belov, Victor; Zhoorov, Eugene

2006-01-01

Pilot innovative accelerator based neutron source for neutron capture therapy of cancer is now of the threshold of its operation at the BINP, Russia. One of the main elements of the facility is lithium target producing neutrons via threshold 7 Li(p,n) 7 Be reaction at 25 kW proton beam with energies 1.915 MeV or 2.5 MeV. The main problems of lithium target were determined to be: 7 Be radioactive isotope activation keeping lithium layer solid, presence of photons due to proton inelastic scattering on lithium nuclei, and radiation blistering. The results of thermal test of target prototype were presented as previous NCT Congress. It becomes clear that water is preferable for cooling the target, and that lithium target 10 cm in diameter is able to run before melting. In the present report, the conception of optimal target is proposed: thin metal disk 10 cm in diameter easy for detaching, with evaporated thin layer of pure lithium from the side of proton beam exposure, its back being intensively cooled with turbulent water flow to maintain lithium layer solid. Design of the target for the neutron source constructed at BINP is shown. The results of investigation of radiation blistering and lithium layer are presented. Target unit of facility is under construction now, and obtaining neutrons is expected in nearest future. (author)

IBM-PC-based reactor neutronics analysis package

International Nuclear Information System (INIS)

Nigg, D.W.; Wessol, D.E.; Grimesey, R.A.; Parsons, D.K.; Wheeler, F.J.; Yoon, W.Y.; Lake, J.A.

1985-01-01

Technical advances over the past few years have led to a situation where a wide range of complex scientific computations can now be done on properly configured microcomputers such as the IBM-PC (personal computer). For a number of reasons, including security, economy, and user convenience, the development of a comprehensive system of reactor neutronics codes suitable for operation on the IBM-PC has been undertaken at the Idaho National Engineering Laboratory (INEL). It is anticipated that a PC-based code system could also have wide applicability in the nuclear engineering education community since conversion of software generated by national laboratories and others to college and university mainframe hardware has historically been a time-consuming process that has sometimes met with only limited success. This paper discusses the philosophy behind the INEL reactor neutronics PC code system and describes those parts of the system that are currently complete, those that are now under development, and those that are still in the planning stage

Towards radiation hard converter material for SiC-based fast neutron detectors

Science.gov (United States)

Tripathi, S.; Upadhyay, C.; Nagaraj, C. P.; Venkatesan, A.; Devan, K.

2018-05-01

In the present work, Geant4 Monte-Carlo simulations have been carried out to study the neutron detection efficiency of the various neutron to other charge particle (recoil proton) converter materials. The converter material is placed over Silicon Carbide (SiC) in Fast Neutron detectors (FNDs) to achieve higher neutron detection efficiency as compared to bare SiC FNDs. Hydrogenous converter material such as High-Density Polyethylene (HDPE) is preferred over other converter materials due to the virtue of its high elastic scattering reaction cross-section for fast neutron detection at room temperature. Upon interaction with fast neutrons, hydrogenous converter material generates recoil protons which liberate e-hole pairs in the active region of SiC detector to provide a detector signal. The neutron detection efficiency offered by HDPE converter is compared with several other hydrogenous materials viz., 1) Lithium Hydride (LiH), 2) Perylene, 3) PTCDA . It is found that, HDPE, though providing highest efficiency among various studied materials, cannot withstand high temperature and harsh radiation environment. On the other hand, perylene and PTCDA can sustain harsh environments, but yields low efficiency. The analysis carried out reveals that LiH is a better material for neutron to other charge particle conversion with competent efficiency and desired radiation hardness. Further, the thickness of LiH has also been optimized for various mono-energetic neutron beams and Am-Be neutron source generating a neutron fluence of 109 neutrons/cm2. The optimized thickness of LiH converter for fast neutron detection is found to be ~ 500 μm. However, the estimated efficiency for fast neutron detection is only 0.1%, which is deemed to be inadequate for reliable detection of neutrons. A sensitivity study has also been done investigating the gamma background effect on the neutron detection efficiency for various energy threshold of Low-Level Discriminator (LLD). The detection

Measurement of the neutron spectrum in a room with an accelerator Varian 2300C/D Linac using the Bonner multisphere spectrometer; Medicao do espectro de neutrons em uma sala com um acelerador Varian 2300C/D Linav usando o espectrometro de multiesferas de Bonner

Energy Technology Data Exchange (ETDEWEB)

Cavalcante, D.B.S., E-mail: cavalcante@ird.gov.b [Universidade Federal do Rio de Janeiro (IF/UFRJ), RJ (Brazil). Inst. de Fisica; Fonseca, E.S. da, E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Lemos Junior, R.M. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil); Batista, D.V.S. [Instituto Nacional do Cancer (INCa), Rio de Janeiro, RJ (Brazil)

2009-07-01

The generated neutron field varies considerably and depends on the beam energy, on the shielding of the accelerator, on the filters for beam homogeneity, and also on the mobile collimators and geometry of irradiation. The estimation of the component relative to the photoneutrons has practical interest for evaluation of the radiological risks for the workers and for the patient as well. Due to the high frequency magnetic field, and to the photon abundance resulting of the escape and scattering at treatment room, those measurements present some difficulties. Measurements of the neutron fields can be made with a Bonner spectrometer. Those system was calibrated with referred neutron standard sources and used for make measurements on a spot of the room where a Variant 2300C/D Linac is installed. The unfolding process used the BUNKI computer code for determination of the neutron spectra at the measurement spot

Frequency spectrum analysis of 252Cf neutron source based on LabVIEW

International Nuclear Information System (INIS)

Mi Deling; Li Pengcheng

2011-01-01

The frequency spectrum analysis of 252 Cf Neutron source is an extremely important method in nuclear stochastic signal processing. Focused on the special '0' and '1' structure of neutron pulse series, this paper proposes a fast-correlation algorithm to improve the computational rate of the spectrum analysis system. And the multi-core processor technology is employed as well as multi-threaded programming techniques of LabVIEW to construct frequency spectrum analysis system of 252 Cf neutron source based on LabVIEW. It not only obtains the auto-correlation and cross correlation results, but also auto-power spectrum,cross-power spectrum and ratio of spectral density. The results show that: analysis tools based on LabVIEW improve the fast auto-correlation and cross correlation code operating efficiency about by 25% to 35%, also verify the feasibility of using LabVIEW for spectrum analysis. (authors)

Neutronics of the IFMIF neutron source: development and analysis

International Nuclear Information System (INIS)

Wilson, P.P.H.

1999-01-01

The accurate analysis of this system required the development of a code system and methodology capable of modelling the various physical processes. A generic code system for the neutronics analysis of neutron sources has been created by loosely integrating existing components with new developments: the data processing code NJOY, the Monte Carlo neutron transport code MCNP, and the activation code ALARA were supplemented by a damage data processing program, damChar, and integrated with a number of flexible and extensible modules for the Perl scripting language. Specific advances were required to apply this code system to IFMIF. Based on the ENDF-6 data format requirements of this system, new data evaluations have been implemented for neutron transport and activation. Extensive analysis of the Li(d, xn) reaction has led to a new MCNP source function module, M c DeLi, based on physical reaction models and capable of accurate and flexible modelling of the IFMIF neutron source term. In depth analyses of the neutron flux spectra and spatial distribution throughout the high flux test region permitted a basic validation of the tools and data. The understanding of the features of the neutron flux provided a foundation for the analyses of the other neutron responses. (orig./DGE) [de

A new, 13C-based material for neutron targets

International Nuclear Information System (INIS)

Romanenko, A.I.; Anikeeva, O.B.; Gorbachev, R.V.; Zhmurikov, E.I.; Gubin, K.V.; Logachev, P.V.; Avilov, M.S.; Tsybulya, S.V.; Kryukova, G.N.; Burgina, E.B.; Tecchio, L.

2005-01-01

A 13 C-based neutron-target material is investigated using X-ray diffraction, IR absorption and Raman scattering spectroscopies, transmission electron microscopy, and electrical (conductivity, magnetoresistance, and Hall effect) measurements before and after high-power electron irradiation for various lengths of time [ru

Neutron shielding material

International Nuclear Information System (INIS)

Nodaka, M.; Iida, T.; Taniuchi, H.; Yosimura, K.; Nagahama, H.

1993-01-01

From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

Feasibility evaluation of a neutron grating interferometer with an analyzer grating based on a structured scintillator

Science.gov (United States)

Kim, Youngju; Kim, Jongyul; Kim, Daeseung; Hussey, Daniel. S.; Lee, Seung Wook

2018-03-01

We introduce an analyzer grating based on a structured scintillator fabricated by a gadolinium oxysulfide powder filling method for a symmetric Talbot-Lau neutron grating interferometer. This is an alternative way to analyze the Talbot self-image of a grating interferometer without using an absorption grating to block neutrons. Since the structured scintillator analyzer grating itself generates the signal for neutron detection, we do not need an additional scintillator screen as an absorption analyzer grating. We have developed and tested an analyzer grating based on a structured scintillator in our symmetric Talbot-Lau neutron grating interferometer to produce high fidelity absorption, differential phase, and dark-field contrast images. The acquired images have been compared to results of a grating interferometer utilizing a typical absorption analyzer grating with two commercial scintillation screens. The analyzer grating based on the structured scintillator enhances interference fringe visibility and shows a great potential for economical fabrication, compact system design, and so on. We report the performance of the analyzer grating based on a structured scintillator and evaluate its feasibility for the neutron grating interferometer.

Monte Carlo based dosimetry and treatment planning for neutron capture therapy of brain tumors

International Nuclear Information System (INIS)

Zamenhof, R.G.; Brenner, J.F.; Wazer, D.E.; Madoc-Jones, H.; Clement, S.D.; Harling, O.K.; Yanch, J.C.

1990-01-01

Monte Carlo based dosimetry and computer-aided treatment planning for neutron capture therapy have been developed to provide the necessary link between physical dosimetric measurements performed on the MITR-II epithermal-neutron beams and the need of the radiation oncologist to synthesize large amounts of dosimetric data into a clinically meaningful treatment plan for each individual patient. Monte Carlo simulation has been employed to characterize the spatial dose distributions within a skull/brain model irradiated by an epithermal-neutron beam designed for neutron capture therapy applications. The geometry and elemental composition employed for the mathematical skull/brain model and the neutron and photon fluence-to-dose conversion formalism are presented. A treatment planning program, NCTPLAN, developed specifically for neutron capture therapy, is described. Examples are presented illustrating both one and two-dimensional dose distributions obtainable within the brain with an experimental epithermal-neutron beam, together with beam quality and treatment plan efficacy criteria which have been formulated for neutron capture therapy. The incorporation of three-dimensional computed tomographic image data into the treatment planning procedure is illustrated

Neutron generators at Purnima Lab

International Nuclear Information System (INIS)

Patel, Tarun; Sinha, Amar

2015-01-01

Neutron sources are in a great demand in many area like research, nuclear waste management, industrial process control, medical and also security. Major sources of neutrons are nuclear reactors, radioisotopes and accelerator based neutron generators. For many field applications, reactors cannot be used due to its large size, complicated system, high cost and also safety issues. Radioisotopes like Pu-Be, Am-Be, Cf, are extensively used for many industrial applications. But they are limited in their use due to their low source strength and also handling difficulties due to radioactivity. They are also not suitable for pulsed neutron applications. In contrast, compact size, pulsed operation, on/off operation etc.of accelerator based neutron generators make them very popular for many applications. Particle accelerators based on different types of neutron generators have been developed around the world. Among these deuteron accelerator based D-D and D-T neutron generators are widely used as they produce mono-energetic fast neutrons and in particular high yield of D-T neutron can be obtained with less than 300 KV of accelerating voltage

Nitrogen Detection in Bulk Samples Using a D-D Reaction-Based Portable Neutron Generator

Directory of Open Access Journals (Sweden)

A. A. Naqvi

2013-01-01

Full Text Available Nitrogen concentration was measured via 2.52 MeV nitrogen gamma ray from melamine, caffeine, urea, and disperse orange bulk samples using a newly designed D-D portable neutron generator-based prompt gamma ray setup. Inspite of low flux of thermal neutrons produced by D-D reaction-based portable neutron generator and interference of 2.52 MeV gamma rays from nitrogen in bulk samples with 2.50 MeV gamma ray from bismuth in BGO detector material, an excellent agreement between the experimental and calculated yields of nitrogen gamma rays indicates satisfactory performance of the setup for detection of nitrogen in bulk samples.

Neutron Imaging at Compact Accelerator-Driven Neutron Sources in Japan

Directory of Open Access Journals (Sweden)

Yoshiaki Kiyanagi

2018-03-01

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

Nanotubes based neutron generator for calibration of neutrino and dark matter detectors

Science.gov (United States)

Chepurnov, A. S.; Ionidi, V. Y.; Kirsanov, M. A.; Kitsyuk, E. P.; Klenin, A. A.; Kubankin, A. S.; Oleinik, A. N.; Pavlov, A. A.; Shchagin, A. V.

2017-12-01

The compact 2.45 MeV fast neutron generator with a reduced supply voltage for calibration of low-background neutrino and dark matter detectors was tested. The generator is based on an array of carbon nanotubes. Neutron generation is carried out by applying a high voltage in the range of +10 to + 25 kV to a nanotube array, which cause an ionization of deuterium molecules with the following acceleration of ions in the direction of the grounded target covered by a deuterated polyethylene film. The d(d,n)3He nuclear reaction happens as the result of ions collisions with the target. The dependences of the neutron yield as functions of the applied voltage were obtained for two different types of carbon nanotubes array. It is shown that the type of nanotubes array does not influence significantly on the neutron yield.

A neutron spectrum unfolding code based on generalized regression artificial neural networks

International Nuclear Information System (INIS)

Rosario Martinez-Blanco, Ma. del

2016-01-01

The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, i.e. the optimum selection of the network topology and the long training time. Compared to BPNN, it's usually much faster to train a generalized regression neural network (GRNN). That's mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum, provided that the optimal values of spread has been determined and that the dataset adequately represents the problem space. In addition, GRNN are often more accurate than BPNN in the prediction. These characteristics make GRNNs to be of great interest in the neutron spectrometry domain. This work presents a computational tool based on GRNN capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages using a k-fold cross validation of 3 folds, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a "6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. - Highlights: • Main drawback of neutron spectrometry with BPNN is network topology optimization. • Compared to BPNN, it’s usually much faster to train a (GRNN). • GRNN are often more accurate than BPNN in the prediction. These characteristics make GRNNs to be of great interest. • This computational code, automates the pre-processing, training

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-01-15

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

A proposed experiment for studying the direct neutron-neutron interaction

International Nuclear Information System (INIS)

Hassan Fikry, A.R.; Maayouf, R.M.A.

1979-01-01

An experiment for studying the direct neutron-neutron interaction is suggested. The experiment is based on the combined use of an accelerator, e.g., an electron linear accelerator, together with a mobile pulsed reactor; or using a pulsed beam reactor together with a mobile neutron generator

An introduction to NH-A neutron earth base moisture gage

International Nuclear Information System (INIS)

Zhu Huaian; Jiang Yulan; Yin Xilin; Yu Peiying; Luo Pinjie

1988-01-01

NH-A neutron earth base moisture gage is an accurate instrument which can measure earth moisture rapidly and non-destructively and display moisture results immediately. The deviation is estimated at ±0.012g/cm

Comparison of Experiment and Simulation of the triple GEM-Based Fast Neutron Detector

International Nuclear Information System (INIS)

Wang Xiao-Dong; Luo Wen; Zhang Jun-Wei; Yang He-Run; Duan Li-Min; Lu Chen-Gui; Hu Rong-Jiang; Hu Bi-Tao; Zhang Chun-Hui; Yang Lei; Zhou Jian-Rong; An Lv-Xing

2015-01-01

A detector for fast neutrons based on a 10 × 10 cm"2 triple gas electron multiplier (GEM) device is developed and tested. A neutron converter, which is a high density polyethylene (HDPE) layer, is combined with the triple GEM detector cathode and placed inside the detector, in the path of the incident neutrons. The detector is tested by obtaining the energy deposition spectrum with an Am Be neutron source in the Institute of Modern Physics (IMP) at Lanzhou. In the present work we report the results of the tests and compare them with those of simulations. The transport of fast neutrons and their interactions with the different materials in the detector are simulated with the GEANT4 code, to understand the experimental results. The detector displays a clear response to the incident fast neutrons. However, an unexpected disagreement in the energy dependence of the response between the simulated and measured spectra is observed. The neutron sources used in our simulation include deuterium-tritium (DT, 14 MeV), deuterium-deuterium (DD, 2.45 MeV), and Am Be sources. The simulation results also show that among the secondary particles generated by the incident neutron, the main contributions to the total energy deposition are from recoil protons induced in hydrogen-rich HDPE or Kapton (GEM material), and activation photons induced by neutron interaction with Ar atoms. Their contributions account for 90% of the total energy deposition. In addition, the dependence of neutron deposited energy spectrum on the composition of the gas mixture is presented. (paper)

Prototype Stilbene Neutron Collar

Energy Technology Data Exchange (ETDEWEB)

Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shumaker, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Verbeke, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-10-26

A neutron collar using stilbene organic scintillator cells for fast neutron counting is described for the assay of fresh low enriched uranium (LEU) fuel assemblies. The prototype stilbene collar has a form factor similar to standard He-3 based collars and uses an AmLi interrogation neutron source. This report describes the simulation of list mode neutron correlation data on various fuel assemblies including some with neutron absorbers (burnable Gd poisons). Calibration curves (doubles vs ²³⁵U linear mass density) are presented for both thermal and fast (with Cd lining) modes of operation. It is shown that the stilbene collar meets or exceeds the current capabilities of He-3 based neutron collars. A self-consistent assay methodology, uniquely suited to the stilbene collar, using triples is described which complements traditional assay based on doubles calibration curves.

Study of neutron fields around an intense neutron generator.

Science.gov (United States)

Kicka, L; Machrafi, R; Miller, A

2017-12-01

Neutron fields in the vicinity of the newly built neutron facility, at the University of Ontario Institute of Technology (UOIT), have been investigated in a series of Monte Carlo simulations and measurements. The facility hosts a P-385 neutron generator based on a deuterium-deuterium fusion reaction. The neutron fluence at different locations around the neutron generator facility has been simulated using MCNPX 2.7E Monte Carlo particle transport program. To characterize neutron fields, three neutron sources were modeled with distributions corresponding to different incident deuteron energies of 90kV, 110kV, and 130kV. Measurements have been carried out to determine the dose rate at locations adjacent to the generator using bubble detectors (BDs). The neutron intensity was evaluated and the total dose rates corresponding to different applied acceleration potentials were estimated at various locations. Copyright © 2017 Elsevier Ltd. All rights reserved.

Study of a nTHGEM-based thermal neutron detector

Science.gov (United States)

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

2016-07-01

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

GPU-accelerated 3D neutron diffusion code based on finite difference method

Energy Technology Data Exchange (ETDEWEB)

Xu, Q.; Yu, G.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ. (China)

2012-07-01

Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

GPU-accelerated 3D neutron diffusion code based on finite difference method

International Nuclear Information System (INIS)

Xu, Q.; Yu, G.; Wang, K.

2012-01-01

Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

Progress report: determinations of the neutron-neutron scattering length ann from kinematically incomplete neutron-deuteron breakup data revisited

International Nuclear Information System (INIS)

Tornow, W.; Braun, R.T.; Witala, H.

1996-01-01

We review published analyses of the final-state-interaction enhancement observed in proton energy distributions obtained from kinematically incomplete neutron-deuteron breakup experiments. We compare the results derived from these analyses for the neutron-neutron scattering length, a nn with our results based on a rigorous treatment of the three-nucleon Faddeev equations in conjunction with the use of realistic nucleon-nucleon potentials. Our values for a nn deviate outside the quoted uncertainties from the ones obtained in the previous analyses where simplified nucleon-nucleon interaction models were employed. In contrast to the previous determinations, the present results for a nn are in clear disagreement with the values for a nn based on π - -deuteron capture experiments. Unless inconsistencies in the experimental neutron-deuteron breakup data at low energies can be resolved and the influence of possible three-nucleon-force effects can be reliably determined, we recommend that one not resort to the kinematically incomplete neutron-deuteron breakup reaction as a tool for determining a quantity as important for nuclear and particle physics as is the neutron-neutron scattering length a nn . (author)

Workplace testing of the new single sphere neutron spectrometer based on Dysprosium activation foils (Dy-SSS)

International Nuclear Information System (INIS)

Bedogni, R.; Gómez-Ros, J.M.; Esposito, A.; Gentile, A.; Chiti, M.; Palacios-Pérez, L.; Angelone, M.; Tana, L.

2012-01-01

A photon insensitive passive neutron spectrometer consisting of a single moderating polyethylene sphere with Dysprosium activation foils arranged along three perpendicular axes was designed by CIEMAT and INFN. The device is called Dy-SSS (Dy foil-based Single Sphere Spectrometer). It shows nearly isotropic response in terms of neutron fluence up to 20 MeV. The first prototype, previously calibrated with 14 MeV neutrons, has been recently tested in workplaces having different energy and directional distributions. These are a 2.5 MeV nearly mono-chromatic and mono-directional beam available at the ENEA Frascati Neutron Generator (FNG) and the photo-neutron field produced in a 15 MV Varian CLINAC DHX medical accelerator, located in the Ospedale S. Chiara (Pisa). Both neutron spectra are known through measurements with a Bonner Sphere Spectrometer. In both cases the experimental response of the Dy-SSS agrees with the reference data. Moreover, it is demonstrated that the spectrometric capability of the new device are independent from the directional distribution of the neutron field. This opens the way to a new generation of moderation-based neutron instruments, presenting all advantages of the Bonner sphere spectrometer without the disadvantage of the repeated exposures. This concept is being developed within the NESCOFI@BTF project of INFN (Commissione Scientifica Nazionale 5).

Workplace testing of the new single sphere neutron spectrometer based on Dysprosium activation foils (Dy-SSS)

Science.gov (United States)

Bedogni, R.; Gómez-Ros, J. M.; Esposito, A.; Gentile, A.; Chiti, M.; Palacios-Pérez, L.; Angelone, M.; Tana, L.

2012-08-01

A photon insensitive passive neutron spectrometer consisting of a single moderating polyethylene sphere with Dysprosium activation foils arranged along three perpendicular axes was designed by CIEMAT and INFN. The device is called Dy-SSS (Dy foil-based Single Sphere Spectrometer). It shows nearly isotropic response in terms of neutron fluence up to 20 MeV. The first prototype, previously calibrated with 14 MeV neutrons, has been recently tested in workplaces having different energy and directional distributions. These are a 2.5 MeV nearly mono-chromatic and mono-directional beam available at the ENEA Frascati Neutron Generator (FNG) and the photo-neutron field produced in a 15 MV Varian CLINAC DHX medical accelerator, located in the Ospedale S. Chiara (Pisa). Both neutron spectra are known through measurements with a Bonner Sphere Spectrometer. In both cases the experimental response of the Dy-SSS agrees with the reference data. Moreover, it is demonstrated that the spectrometric capability of the new device are independent from the directional distribution of the neutron field. This opens the way to a new generation of moderation-based neutron instruments, presenting all advantages of the Bonner sphere spectrometer without the disadvantage of the repeated exposures. This concept is being developed within the NESCOFI@BTF project of INFN (Commissione Scientifica Nazionale 5).

Workplace testing of the new single sphere neutron spectrometer based on Dysprosium activation foils (Dy-SSS)

Energy Technology Data Exchange (ETDEWEB)

Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN-LNF (Frascati National Laboratories), Via E. Fermi n. 40-00044 Frascati (Italy); Gomez-Ros, J.M. [INFN-LNF (Frascati National Laboratories), Via E. Fermi n. 40-00044 Frascati (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Esposito, A.; Gentile, A.; Chiti, M.; Palacios-Perez, L. [INFN-LNF (Frascati National Laboratories), Via E. Fermi n. 40-00044 Frascati (Italy); Angelone, M. [ENEA C.R. Frascati, C.P. 65, 00044 Frascati (Italy); Tana, L. [A.O. Universitaria Pisana-Ospedale S. Chiara, Via Bonanno Pisano, Pisa (Italy)

2012-08-21

A photon insensitive passive neutron spectrometer consisting of a single moderating polyethylene sphere with Dysprosium activation foils arranged along three perpendicular axes was designed by CIEMAT and INFN. The device is called Dy-SSS (Dy foil-based Single Sphere Spectrometer). It shows nearly isotropic response in terms of neutron fluence up to 20 MeV. The first prototype, previously calibrated with 14 MeV neutrons, has been recently tested in workplaces having different energy and directional distributions. These are a 2.5 MeV nearly mono-chromatic and mono-directional beam available at the ENEA Frascati Neutron Generator (FNG) and the photo-neutron field produced in a 15 MV Varian CLINAC DHX medical accelerator, located in the Ospedale S. Chiara (Pisa). Both neutron spectra are known through measurements with a Bonner Sphere Spectrometer. In both cases the experimental response of the Dy-SSS agrees with the reference data. Moreover, it is demonstrated that the spectrometric capability of the new device are independent from the directional distribution of the neutron field. This opens the way to a new generation of moderation-based neutron instruments, presenting all advantages of the Bonner sphere spectrometer without the disadvantage of the repeated exposures. This concept is being developed within the NESCOFI@BTF project of INFN (Commissione Scientifica Nazionale 5).

Secondary Neutron Production from Space Radiation Interactions: Advances in Model and Experimental Data Base Development

Science.gov (United States)

Heilbronn, Lawrence H.; Townsend, Lawrence W.; Braley, G. Scott; Iwata, Yoshiyuki; Iwase, Hiroshi; Nakamura, Takashi; Ronningen, Reginald M.; Cucinotta, Francis A.

2003-01-01

For humans engaged in long-duration missions in deep space or near-Earth orbit, the risk from exposure to galactic and solar cosmic rays is an important factor in the design of spacecraft, spacesuits, and planetary bases. As cosmic rays are transported through shielding materials and human tissue components, a secondary radiation field is produced. Neutrons are an important component of that secondary field, especially in thickly-shielded environments. Calculations predict that 50% of the dose-equivalent in a lunar or Martian base comes from neutrons, and a recent workshop held at the Johnson Space Center concluded that as much as 30% of the dose in the International Space Station may come from secondary neutrons. Accelerator facilities provide a means for measuring the effectiveness of various materials in their ability to limit neutron production, using beams and energies that are present in cosmic radiation. The nearly limitless range of beams, energies, and target materials that are present in space, however, means that accelerator-based experiments will not provide a complete database of cross sections and thick-target yields that are necessary to plan and design long-duration missions. As such, accurate nuclear models of neutron production are needed, as well as data sets that can be used to compare with, and verify, the predictions from such models. Improvements in a model of secondary neutron production from heavy-ion interactions are presented here, along with the results from recent accelerator-based measurements of neutron-production cross sections. An analytical knockout-ablation model capable of predicting neutron production from high-energy hadron-hadron interactions (both nucleon-nucleus and nucleus-nucleus collisions) has been previously developed. In the knockout stage, the collision between two nuclei result in the emission of one or more nucleons from the projectile and/or target. The resulting projectile and target remnants, referred to as

9Be(d,n)10B-based neutron sources for BNCT

International Nuclear Information System (INIS)

Capoulat, M.E.; Herrera, M.S.; Minsky, D.M.; González, S.J.; Kreiner, A.J.

2014-01-01

In the frame of accelerator-based BNCT, the 9 Be(d,n) 10 B reaction was investigated as a possible source of epithermal neutrons. In order to determine the configuration in terms of bombarding energy, target thickness and Beam Shaping Assembly (BSA) design that results in the best possible beam quality, a systematic optimization study was carried out. From this study, the optimal configuration resulted in tumor doses ≥40 Gy-Eq, with a maximum value of 51 Gy-Eq at a depth of about 2.7 cm, in a 60 min treatment. The optimal configuration was considered for the treatment planning assessment of a real Glioblastoma Multiforme case. From this, the resulted dose performances were comparable to those obtained with an optimized 7 Li(p,n)-based neutron source, under identical conditions and subjected to the same clinical protocol. - Highlights: • Study of the 9 Be(d,n) 10 B reaction as a source of epithermal neutrons for BNCT. • Evaluation of the optimal configuration of target thickness, deuteron energy and BSA design. • Computational dose assessment for brain tumor treatments using the MCNP code. • Treatment planning assessment of a particular clinical Glioblastoma Multiforme case. • Dose performances were comparable to those obtained with an optimized 7 Li(p,n)-based source

Biomedical neutron research at the Californium User Facility for neutron science

International Nuclear Information System (INIS)

Martin, R.C.; Byrne, T.E.; Miller, L.F.

1997-01-01

The Californium User Facility for Neutron Science has been established at Oak Ridge National Laboratory (ORNL). The Californium User Facility (CUF) is a part of the larger Californium Facility, which fabricates and stores compact 252 Cf neutron sources for worldwide distribution. The CUF can provide a cost-effective option for research with 252 Cf sources. Three projects at the CUF that demonstrate the versatility of 252 Cf for biological and biomedical neutron-based research are described: future establishment of a 252 Cf-based neutron activation analysis system, ongoing work to produce miniature high-intensity, remotely afterloaded 252 Cf sources for tumor therapy, and a recent experiment that irradiated living human lung cancer cells impregnated with experimental boron compounds to test their effectiveness for boron neutron capture therapy

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

International Nuclear Information System (INIS)

Kiyanagi, Yoshiaki

2012-01-01

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

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

International Nuclear Information System (INIS)

Shaikh, A.M.

2013-01-01

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

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.

Lithium-based neutron detectors

International Nuclear Information System (INIS)

Yursova, L.

1977-01-01

The problems of using scintillation lithium-based detectors (LiJ(Eu) and 6 LiJ(Eu)), as well as lithium glasses for neutron detection are described. As compared with the glasses the LiJ(Eu) monocrystal possesses substantially higher energy resolution, its luminescence yield is considerably higher (in some cases ten fold), its application makes possible gamma radiation discrimination with the energy approximately four times higher and its higher specific mass ensures better efficiency of gamma radiation counting. The only 6 LiJ(Eu) drawback is its high hydroscopicity as well as its possibility to be used only in a limited temperature range (maximum temperature +35 deg C). The lithium glass can be used (with the exception of spectrometric measurements and radiation mixed regions measurement) with more than 1 MeV gamma radiation energy in a wide temperature range, in agressive, corroding and acid media

Characteristics of SiC neutron sensor spectrum unfolding process based on Bayesian inference

Energy Technology Data Exchange (ETDEWEB)

Cetnar, Jerzy; Krolikowski, Igor [Faculty of Energy and Fuels AGH - University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Ottaviani, L. [IM2NP, UMR CNRS 7334, Aix-Marseille University, Case 231 -13397 Marseille Cedex 20 (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

2015-07-01

This paper deals with SiC detector signal interpretation in neutron radiation measurements in mixed neutron gamma radiation fields, which is called the detector inverse problem or the spectrum unfolding, and it aims in finding a representation of the primary radiation, based on the measured detector signals. In our novel methodology we resort to Bayesian inference approach. In the developed procedure the resultant spectra is unfolded form detector channels reading, where the estimated neutron fluence in a group structure is obtained with its statistical characteristic comprising of standard deviation and correlation matrix. In the paper we present results of unfolding process for case of D-T neutron source in neutron moderating environment. Discussions of statistical properties of obtained results are presented as well as of the physical meaning of obtained correlation matrix of estimated group fluence. The presented works has been carried out within the I-SMART project, which is part of the KIC InnoEnergy R and D program. (authors)

The $\\mu$TPC Method: Improving the Position Resolution of Neutron Detectors Based on MPGDs

CERN Document Server

Pfeiffer, Dorothea; Birch, Jens; Hall-Wilton, Richard; Höglund, Carina; Hultman, Lars; Iakovidis, George; Oliveri, Eraldo; Oksanen, Esko; Ropelewski, Leszek; Thuiner, Patrik

2015-01-01

Due to the Helium-3 crisis, alternatives to the standard neutron detection techniques are becoming urgent. In addition, the instruments of the European Spallation Source (ESS) require advances in the state of the art of neutron detection. The instruments need detectors with excellent neutron detection efficiency, high-rate capabilities and unprecedented spatial resolution. The Macromolecular Crystallography instrument (NMX) requires a position resolution in the order of 200 um over a wide angular range of incoming neutrons. Solid converters in combination with Micro Pattern Gaseous Detectors (MPGDs) are proposed to meet the new requirements. Charged particles rising from the neutron capture have usually ranges larger than several millimetres in gas. This is apparently in contrast with the requirements for the position resolution. In this paper, we present an analysis technique, new in the field of neutron detection, based on the Time Projection Chamber (TPC) concept. Using a standard Single-GEM with the catho...

Neutron Optics: Towards Applications for Hot Neutrons

International Nuclear Information System (INIS)

Schanzer, C; Schneider, M; Böni, P

2016-01-01

Supermirrors with large critical angles of reflection, i.e. large index m are an essential ingredient to transport, focus and polarise neutrons over a wide range of energy. Here we summarise the recent developments of supermirror with very large critical angles of reflection and high reflectivity that were conducted at SwissNeutronics as well as their implementation in devices. Approaching critical angles m = 8 times the critical angle of natural nickel makes new applications possible and extends the use of reflection optics towards the regime of hot and epithermal neutrons. Based on comparisons of simulations with experiment we demonstrate future possibilities of applications of large-m supermirrors towards devices for neutrons with short wavelength. (paper)

Basic concepts underlying fast-neutron-based contraband interrogation technology

International Nuclear Information System (INIS)

Fink, C.L.; Guenther, P.T.; Smith, D.L.

1992-01-01

All accelerator-based fast-neutron contraband interrogation systems have many closely interrelated subsystems, whose performance parameters will be critically interdependent. For optimal overall performance, a systems analysis design approach is required. This paper provides a general overview of the interrelationships and the tradeoffs to be considered for optimization of nonaccelerator subsystems

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.

Isotopic neutron sources for neutron activation analysis

International Nuclear Information System (INIS)

Hoste, J.

1988-06-01

This User's Manual is an attempt to provide for teaching and training purposes, a series of well thought out demonstrative experiments in neutron activation analysis based on the utilization of an isotopic neutron source. In some cases, these ideas can be applied to solve practical analytical problems. 19 refs, figs and tabs

Biomedical neutron research at the Californium User Facility for Neutron Science

International Nuclear Information System (INIS)

Martin, R.C.; Byrne, T.E.; Miller, L.F.

1998-01-01

The Californium User Facility for Neutron Science has been established at Oak Ridge National Laboratory (ORNL). The Californium User Facility (CUF) is a part of the larger Californium Facility, which fabricates and stores compact 252 Cf neutron sources for worldwide distribution. The CUF can provide a cost-effective option for research with 252 Cf sources. Three projects at the CUF that demonstrate the versatility of 252 Cf for biological and biomedical neutron-based research are described: future establishment of a 252 Cf-based neutron activation analysis system, ongoing work to produce miniature high-intensity, remotely afterloaded 252 Cf sources for tumor therapy, and a recent experiment that irradiated living human lung cancer cells impregnated with experimental boron compounds to test their effectiveness for boron neutron capture therapy. (author)

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

Science.gov (United States)

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

2018-03-01

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

Neutron rich matter, neutron stars, and their crusts

International Nuclear Information System (INIS)

Horowitz, C J

2011-01-01

Neutron rich matter is at the heart of many fundamental questions in Nuclear Physics and Astrophysics. What are the high density phases of QCD? Where did the chemical elements come from? What is the structure of many compact and energetic objects in the heavens, and what determines their electromagnetic, neutrino, and gravitational-wave radiations? Moreover, neutron rich matter is being studied with an extraordinary variety of new tools such as Facility for Rare Isotope Beams (FRIB) and the Laser Interferometer Gravitational Wave Observatory (LIGO). We describe the Lead Radius Experiment (PREX) that is using parity violation to measure the neutron radius in 208Pb. This has important implications for neutron stars and their crusts. Using large scale molecular dynamics, we model the formation of solids in both white dwarfs and neutron stars. We find neutron star crust to be the strongest material known, some 10 billion times stronger than steel. It can support mountains on rotating neutron stars large enough to generate detectable gravitational waves. Finally, we describe a new equation of state for supernova and neutron star merger simulations based on the Virial expansion at low densities, and large scale relativistic mean field calculations.

Neutron recognition in the LAND detector for large neutron multiplicity

Energy Technology Data Exchange (ETDEWEB)

Pawlowski, P., E-mail: piotr.pawlowski@ifj.edu.pl [Institute of Nuclear Physics, PAN, Radzikowskiego 152, 31-342 Krakow (Poland); Brzychczyk, J. [Institute of Physics, Jagiellonian University, Reymonta 4, 30-059 Krakow (Poland); Leifels, Y.; Trautmann, W. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Adrich, P. [National Centre for Nuclear Research, PL-00681 Warsaw (Poland); GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Aumann, T. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Bacri, C.O. [Institut de Physique Nucleaire, IN2P3-CNRS et Universite, F-91406 Orsay (France); Barczyk, T. [Institute of Physics, Jagiellonian University, Reymonta 4, 30-059 Krakow (Poland); Bassini, R. [Istituto di Scienze Fisiche, Universita degli Studi and INFN, I-20133 Milano (Italy); Bianchin, S. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Boiano, C. [Istituto di Scienze Fisiche, Universita degli Studi and INFN, I-20133 Milano (Italy); Boretzky, K. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Boudard, A. [IRFU/SPhN, CEA/Saclay, F-91191 Gif-sur-Yvette (France); Chbihi, A. [GANIL, CEA et IN2P3-CNRS, F-14076 Caen (France); Cibor, J.; Czech, B. [Institute of Nuclear Physics, PAN, Radzikowskiego 152, 31-342 Krakow (Poland); De Napoli, M. [Dipartimento di Fisica e Astronomia-Universita and INFN-CT and LNS, I-95123 Catania (Italy); and others

2012-12-01

The performance of the LAND neutron detector is studied. Using an event-mixing technique based on one-neutron data obtained in the S107 experiment at the GSI laboratory, we test the efficiency of various analytic tools used to determine the multiplicity and kinematic properties of detected neutrons. A new algorithm developed recently for recognizing neutron showers from spectator decays in the ALADIN experiment S254 is described in detail. Its performance is assessed in comparison with other methods. The properties of the observed neutron events are used to estimate the detection efficiency of LAND in this experiment.

Neutron dosimetry inside the containment building of Spanish nuclear power plants with PADC based dosemeters

Energy Technology Data Exchange (ETDEWEB)

Garcia-Fuste, M.J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Domingo, C., E-mail: carles.domingo@uab.ca [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Amgarou, K.; Bouassoule, T.; Castelo, J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain)

2009-10-15

The Spanish Nuclear Safety Council (Consejo de Seguridad Nuclear, CSN) recommends performing neutron individual dose assignments at workplaces based on ambient dose equivalent measurements using area monitors and by estimating the amount of time that workers spend in the different monitored environments. In addition, some Spanish nuclear power plants estimate the neutron dose equivalent using albedo thermoluminescence dosemeters (TLD). In the period 2004-2006, our group, together with other research centers, participated in a project, funded by the CSN, with the support of the Spanish Nuclear Power Plants Association (UNESA), to investigate in situ which could be the best practical procedure for individual neutron dose monitoring in nuclear power plants. As part of this survey, several units of the UAB PADC based neutron dosemeter were exposed, on a methacrylate phantom simulating a human body, at four different places inside the containment building of the Asco I nuclear power plant. The influence of different types of calibration neutron fields is analysed and the dose equivalent for each point is estimated.

A neutron amplifier: prospects for reactor-based waste transmutation

International Nuclear Information System (INIS)

Blanovsky, A.

2004-01-01

A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 10 15 n/s (0.7 MW D-T or 2.5 MW electron beam power) is needed to control the HFSR that produces 300 MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode. (author)

Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.

Science.gov (United States)

Shan, Qing; Chu, Shengnan; Jia, Wenbao

2015-11-01

Coal is one of the most popular fuels in the world. The use of coal not only produces carbon dioxide, but also contributes to the environmental pollution by heavy metals. In prompt gamma-ray neutron activation analysis (PGNAA)-based coal analyzer, the characteristic gamma rays of C and O are mainly induced by fast neutrons, whereas thermal neutrons can be used to induce the characteristic gamma rays of H, Si, and heavy metals. Therefore, appropriate thermal and fast neutrons are beneficial in improving the measurement accuracy of heavy metals, and ensure that the measurement accuracy of main elements meets the requirements of the industry. Once the required yield of the deuterium-tritium (d-T) neutron generator is determined, appropriate thermal and fast neutrons can be obtained by optimizing the neutron source term. In this article, the Monte Carlo N-Particle (MCNP) Transport Code and Evaluated Nuclear Data File (ENDF) database are used to optimize the neutron source term in PGNAA-based coal analyzer, including the material and shape of the moderator and neutron reflector. The optimized targets include two points: (1) the ratio of the thermal to fast neutron is 1:1 and (2) the total neutron flux from the optimized neutron source in the sample increases at least 100% when compared with the initial one. The simulation results show that, the total neutron flux in the sample increases 102%, 102%, 85%, 72%, and 62% with Pb, Bi, Nb, W, and Be reflectors, respectively. Maximum optimization of the targets is achieved when the moderator is a 3-cm-thick lead layer coupled with a 3-cm-thick high-density polyethylene (HDPE) layer, and the neutron reflector is a 27-cm-thick hemispherical lead layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

Reactor neutron dosimetry

International Nuclear Information System (INIS)

Najzer, M.; Pauko, M.; Glumac, B.; Acquah, I.N.; Moskon, F.

1977-01-01

An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

A polarizing neutron periscope for neutron imaging

International Nuclear Information System (INIS)

Schulz, Michael; Boeni, Peter; Calzada, Elbio; Muehlbauer, Martin; Neubauer, Andreas; Schillinger, Burkhard

2009-01-01

Optical neutron polarizers like guides or benders destroy the collimation of a neutron beam due to multiple reflections or scattering. This makes them unsuitable for their use in polarized neutron radiography, because the beam collimation is essential to obtain high spatial resolution. We have developed a neutron polarizer based on the principle of an optical periscope with a zigzag double reflection on two parallel high-m supermirror polarizers. If the supermirrors are perfectly parallel and flat, the beam collimation is left unchanged by such a device. A first proof of concept version of this type of polarizer was built and tested. We expect to achieve a beam polarization of up to 99% with an improved version yet to be built.

In vivo evaluation of neutron capture therapy effectivity using calcium phosphate-based nanoparticles as Gd-DTPA delivery agent.

Science.gov (United States)

Dewi, Novriana; Mi, Peng; Yanagie, Hironobu; Sakurai, Yuriko; Morishita, Yasuyuki; Yanagawa, Masashi; Nakagawa, Takayuki; Shinohara, Atsuko; Matsukawa, Takehisa; Yokoyama, Kazuhito; Cabral, Horacio; Suzuki, Minoru; Sakurai, Yoshinori; Tanaka, Hiroki; Ono, Koji; Nishiyama, Nobuhiro; Kataoka, Kazunori; Takahashi, Hiroyuki

2016-04-01

A more immediate impact for therapeutic approaches of current clinical research efforts is of major interest, which might be obtained by developing a noninvasive radiation dose-escalation strategy, and neutron capture therapy represents one such novel approach. Furthermore, some recent researches on neutron capture therapy have focused on using gadolinium as an alternative or complementary for currently used boron, taking into account several advantages that gadolinium offers. Therefore, in this study, we carried out feasibility evaluation for both single and multiple injections of gadolinium-based MRI contrast agent incorporated in calcium phosphate nanoparticles as neutron capture therapy agent. In vivo evaluation was performed on colon carcinoma Col-26 tumor-bearing mice irradiated at nuclear reactor facility of Kyoto University Research Reactor Institute with average neutron fluence of 1.8 × 10(12) n/cm(2). Antitumor effectivity was evaluated based on tumor growth suppression assessed until 27 days after neutron irradiation, followed by histopathological analysis on tumor slice. The experimental results showed that the tumor growth of irradiated mice injected beforehand with Gd-DTPA-incorporating calcium phosphate-based nanoparticles was suppressed up to four times higher compared to the non-treated group, supported by the results of histopathological analysis. The results of antitumor effectivity observed on tumor-bearing mice after neutron irradiation indicated possible effectivity of gadolinium-based neutron capture therapy treatment.

First experiments with a liquid-lithium based high-intensity 25-keV neutron source

International Nuclear Information System (INIS)

Paul, M.

2014-01-01

A high-intensity neutron source based on a Liquid-Lithium Target (LiLiT) and the 7 Li(p,n) reaction was developed at SARAF (Soreq Applied Research Accelerator Facility, Israel) and is used for nuclear astrophysics experiments. The setup was commissioned with a 1.3 mA proton beam at 1.91 MeV, producing a neutron yield of ~ 2 ×10 10 n/s, more than one order of magnitude larger than conventional 7 Li(p,n)-based neutron sources and peaked at ~25 keV. The LiLiT device consists of a high-velocity (> 4 m/s) vertical jet of liquid lithium (~200 °C) whose free surface is bombarded by the proton beam. The lithium jet acts both as the neutron-producing target and as a power beam dump. The target dissipates a peak power areal density of 2.5 kW/cm 2 and peak volume density of 0.5 MW/cm 3 with no change of temperature or vacuum regime in the vacuum chamber. Preliminary results of Maxwellian-averaged cross section measurements for stable isotopes of Zr and Ce, performed by activation in the neutron flux of LiLiT, and nuclear-astrophysics experiments in planning will be described. (author)

The Los Alamos Neutron Science Center Spallation Neutron Sources

International Nuclear Information System (INIS)

Nowicki, Suzanne F.; Wender, Stephen A.; Mocko, Michael

2017-01-01

The Los Alamos Neutron Science Center (LANSCE) provides the scientific community with intense sources of neutrons, which can be used to perform experiments supporting civilian and national security research. These measurements include nuclear physics experiments for the defense program, basic science, and the radiation effect programs. This paper focuses on the radiation effects program, which involves mostly accelerated testing of semiconductor parts. When cosmic rays strike the earth's atmosphere, they cause nuclear reactions with elements in the air and produce a wide range of energetic particles. Because neutrons are uncharged, they can reach aircraft altitudes and sea level. These neutrons are thought to be the most important threat to semiconductor devices and integrated circuits. The best way to determine the failure rate due to these neutrons is to measure the failure rate in a neutron source that has the same spectrum as those produced by cosmic rays. Los Alamos has a high-energy and a low-energy neutron source for semiconductor testing. Both are driven by the 800-MeV proton beam from the LANSCE accelerator. The high-energy neutron source at the Weapons Neutron Research (WNR) facility uses a bare target that is designed to produce fast neutrons with energies from 100 keV to almost 800 MeV. The measured neutron energy distribution from WNR is very similar to that of the cosmic-ray-induced neutrons in the atmosphere. However, the flux provided at the WNR facility is typically 5×107 times more intense than the flux of the cosmic-ray-induced neutrons. This intense neutron flux allows testing at greatly accelerated rates. An irradiation test of less than an hour is equivalent to many years of neutron exposure due to cosmic-ray neutrons. The low-energy neutron source is located at the Lujan Neutron Scattering Center. It is based on a moderated source that provides useful neutrons from subthermal energies to ~100 keV. The characteristics of these sources

The Los Alamos Neutron Science Center Spallation Neutron Sources

Science.gov (United States)

Nowicki, Suzanne F.; Wender, Stephen A.; Mocko, Michael

The Los Alamos Neutron Science Center (LANSCE) provides the scientific community with intense sources of neutrons, which can be used to perform experiments supporting civilian and national security research. These measurements include nuclear physics experiments for the defense program, basic science, and the radiation effect programs. This paper focuses on the radiation effects program, which involves mostly accelerated testing of semiconductor parts. When cosmic rays strike the earth's atmosphere, they cause nuclear reactions with elements in the air and produce a wide range of energetic particles. Because neutrons are uncharged, they can reach aircraft altitudes and sea level. These neutrons are thought to be the most important threat to semiconductor devices and integrated circuits. The best way to determine the failure rate due to these neutrons is to measure the failure rate in a neutron source that has the same spectrum as those produced by cosmic rays. Los Alamos has a high-energy and a low-energy neutron source for semiconductor testing. Both are driven by the 800-MeV proton beam from the LANSCE accelerator. The high-energy neutron source at the Weapons Neutron Research (WNR) facility uses a bare target that is designed to produce fast neutrons with energies from 100 keV to almost 800 MeV. The measured neutron energy distribution from WNR is very similar to that of the cosmic-ray-induced neutrons in the atmosphere. However, the flux provided at the WNR facility is typically 5×107 times more intense than the flux of the cosmic-ray-induced neutrons. This intense neutron flux allows testing at greatly accelerated rates. An irradiation test of less than an hour is equivalent to many years of neutron exposure due to cosmic-ray neutrons. The low-energy neutron source is located at the Lujan Neutron Scattering Center. It is based on a moderated source that provides useful neutrons from subthermal energies to ∼100 keV. The characteristics of these sources, and

Fission reactor based epithermal neutron irradiation facilities for routine clinical application in BNCT-Hatanaka memorial lecture

International Nuclear Information System (INIS)

Harling, Otto K.

2009-01-01

Based on experience gained in the recent clinical studies at MIT/Harvard, the desirable characteristics of epithermal neutron irradiation facilities for eventual routine clinical BNCT are suggested. A discussion of two approaches to using fission reactors for epithermal neutron BNCT is provided. This is followed by specific suggestions for the performance and features needed for high throughput clinical BNCT. An example of a current state-of-the-art, reactor based facility, suited for routine clinical use is discussed. Some comments are provided on the current status of reactor versus accelerator based epithermal neutron sources for BNCT. This paper concludes with a summary and a few personal observations on BNCT by the author.

RADIATION PERFORMANCE OF GAN AND INAS/GAAS QUANTUM DOT BASED DEVICES SUBJECTED TO NEUTRON RADIATION

Directory of Open Access Journals (Sweden)

Dhiyauddin Ahmad Fauzi

2017-05-01

Full Text Available In addition to their useful optoelectronics functions, gallium nitride (GaN and quantum dots (QDs based structures are also known for their radiation hardness properties. With demands on such semiconductor material structures, it is important to investigate the differences in reliability and radiation hardness properties of these two devices. For this purpose, three sets of GaN light-emitting diode (LED and InAs/GaAs dot-in-a well (DWELL samples were irradiated with thermal neutron of fluence ranging from 3×1013 to 6×1014 neutron/cm2 in PUSPATI TRIGA research reactor. The radiation performances for each device were evaluated based on the current-voltage (I-V and capacitance-voltage (C-V electrical characterisation method. Results suggested that the GaN based sample is less susceptible to electrical changes due to the thermal neutron radiation effects compared to the QD based sample.

Boron-coated straws as a replacement for 3He-based neutron detectors

International Nuclear Information System (INIS)

Lacy, Jeffrey L.; Athanasiades, Athanasios; Sun, Liang; Martin, Christopher S.; Lyons, Tom D.; Foss, Michael A.; Haygood, Hal B.

2011-01-01

US and international government efforts to equip major seaports with large area neutron detectors, aimed to intercept the smuggling of nuclear materials, have precipitated a critical shortage of 3 He gas. It is estimated that the annual demand of 3 He for US security applications alone is more than the worldwide supply. This is strongly limiting the prospects of neutron science, safeguards, and other applications that rely heavily on 3 He-based detectors. Clearly, alternate neutron detection technologies that can support large sensitive areas, and have low gamma sensitivity and low cost must be developed. We propose a low-cost technology based on long copper tubes (straws), coated on the inside with a thin layer of 10 B-enriched boron carbide ( 10 B 4 C). In addition to the high abundance of boron on Earth and low cost of 10 B enrichment, the boron-coated straw (BCS) detector offers distinct advantages over conventional 3 He-based detectors, and alternate technologies such as 10 BF 3 tubes and 10 B-coated rigid tubes. These include better distribution inside moderator assemblies, many-times faster electronic signals, no pressurization, improved gamma-ray rejection, no toxic or flammable gases, and ease of serviceability. We present the performance of BCS detectors dispersed in a solid plastic moderator to address the need for portal monitoring. The design adopts the outer dimensions of currently deployed 3 He-based monitors, but takes advantage of the small BCS diameter to achieve a more uniform distribution of neutron converter throughout the moderating material. We show that approximately 63 BCS detectors, each 205 cm long, distributed inside the moderator, can match or exceed the detection efficiency of typical monitors fitted with a 5 cm diameter 3 He tube, 187 cm long, pressurized to 3 atm.

Boron-coated straws as a replacement for 3He-based neutron detectors

Science.gov (United States)

Lacy, Jeffrey L.; Athanasiades, Athanasios; Sun, Liang; Martin, Christopher S.; Lyons, Tom D.; Foss, Michael A.; Haygood, Hal B.

2011-10-01

US and international government efforts to equip major seaports with large area neutron detectors, aimed to intercept the smuggling of nuclear materials, have precipitated a critical shortage of 3He gas. It is estimated that the annual demand of 3He for US security applications alone is more than the worldwide supply. This is strongly limiting the prospects of neutron science, safeguards, and other applications that rely heavily on 3He-based detectors. Clearly, alternate neutron detection technologies that can support large sensitive areas, and have low gamma sensitivity and low cost must be developed. We propose a low-cost technology based on long copper tubes (straws), coated on the inside with a thin layer of 10B-enriched boron carbide ( 10B 4C). In addition to the high abundance of boron on Earth and low cost of 10B enrichment, the boron-coated straw (BCS) detector offers distinct advantages over conventional 3He-based detectors, and alternate technologies such as 10BF 3 tubes and 10B-coated rigid tubes. These include better distribution inside moderator assemblies, many-times faster electronic signals, no pressurization, improved gamma-ray rejection, no toxic or flammable gases, and ease of serviceability. We present the performance of BCS detectors dispersed in a solid plastic moderator to address the need for portal monitoring. The design adopts the outer dimensions of currently deployed 3He-based monitors, but takes advantage of the small BCS diameter to achieve a more uniform distribution of neutron converter throughout the moderating material. We show that approximately 63 BCS detectors, each 205 cm long, distributed inside the moderator, can match or exceed the detection efficiency of typical monitors fitted with a 5 cm diameter 3He tube, 187 cm long, pressurized to 3 atm.

Neutron based evaluation in support of NEAMS

Energy Technology Data Exchange (ETDEWEB)

Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bourke, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian Simon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-06-07

The primary objective of the Advanced Non-Destructive fuel Examination (ANDE) work package is to develop capability that has the potential to accelerate insight and development of ceramic and metallic fuels. Establishing unique validation opportunities for new models is a key component of this effort. To explore opportunities a series of interactions were held with NEAMS modelers at LANL. The focus was to identify experiments that draw on the unique capabilities of neutron scattering and imaging for studies of nuclear fuel particularly in areas where experimental data can be valuable for of models validation. The neutron characterization techniques applied in the ANDE program span length scales from millimeter to micrometer to angstroms. Spatial heterogeneities of interest include cracks, pores and inclusions, crystal structure, phase composition, stoichiometry texture, chemistry and atomic thermal motion. Neutrons offer characterization opportunities that are distinct from other probes such as X-rays, electrons or protons. This report describes a variety of opportunities whereby neutron data can be related to models and lists some opportunities.

Development of a two-dimensional imaging detector based on a neutron scintillator with wavelength-shifting fibers

CERN Document Server

Sakai, K; Oku, T; Morimoto, K; Shimizu, H M; Tokanai, F; Gorin, A; Manuilov, I V; Ryazantsev, A; Ino, T; Kuroda, K; Suzuki, J

2002-01-01

For evaluating neutron optical devices, a two-dimensional (2D) detector based on a neutron scintillator with wavelength-shifting fibers has been developed at RIKEN. We have investigated a ZnS(Ag)+LiF and a Li glass plate as neutron scintillators with the coding technique for realizing the large sensitive area of 50 x 50 mm sup 2. After fabricating the 2D detector, its performance was tested using cold neutrons at JAERI. As a result, a spatial resolution of propor to 1.0 mm was obtained. (orig.)

CORE-COLLAPSE SUPERNOVA EQUATIONS OF STATE BASED ON NEUTRON STAR OBSERVATIONS

International Nuclear Information System (INIS)

Steiner, A. W.; Hempel, M.; Fischer, T.

2013-01-01

Many of the currently available equations of state for core-collapse supernova simulations give large neutron star radii and do not provide large enough neutron star masses, both of which are inconsistent with some recent neutron star observations. In addition, one of the critical uncertainties in the nucleon-nucleon interaction, the nuclear symmetry energy, is not fully explored by the currently available equations of state. In this article, we construct two new equations of state which match recent neutron star observations and provide more flexibility in studying the dependence on nuclear matter properties. The equations of state are also provided in tabular form, covering a wide range in density, temperature, and asymmetry, suitable for astrophysical simulations. These new equations of state are implemented into our spherically symmetric core-collapse supernova model, which is based on general relativistic radiation hydrodynamics with three-flavor Boltzmann neutrino transport. The results are compared with commonly used equations of state in supernova simulations of 11.2 and 40 M ☉ progenitors. We consider only equations of state which are fitted to nuclear binding energies and other experimental and observational constraints. We find that central densities at bounce are weakly correlated with L and that there is a moderate influence of the symmetry energy on the evolution of the electron fraction. The new models also obey the previously observed correlation between the time to black hole formation and the maximum mass of an s = 4 neutron star

Measurement of accelerator-based neutron distributions using nuclear track detectors

International Nuclear Information System (INIS)

Al-Jarallah, M.I.; Abu-Jarad, F.; Rehman, Fazal-ur-; Khiari, F.Z.; Aksoy, A.; Nassar, R.

2000-01-01

Nuclear track detectors were used to measure the longitudinal and transverse distributions of slow neutrons in a moderated neutron field as well as the longitudinal and transverse distributions of fast neutrons produced on the 0 deg. beam line of the KFUPM 350 keV ion accelerator. The neutrons were first produced from the T(d,n) 4 He reaction with a neutron energy of approximately 14 MeV and were then moderated in a cylindrical polyethylene moderator placed at the end of the 0 deg. beam line. The optimal transverse slow neutron distribution was found to be uniform within ±4.5% at a 3 cm depth inside the moderator. The fast neutron distribution component along the moderator central axis exhibited an exponential-like drop in intensity with depth. Linearity checks of alpha and proton recoil track density with irradiation time for the nuclear track detectors were verified for both slow and fast neutrons

Measurement of accelerator-based neutron distributions using nuclear track detectors

Energy Technology Data Exchange (ETDEWEB)

Al-Jarallah, M.I. E-mail: mibrahim@kfupm.edu.sa; Abu-Jarad, F.; Rehman, Fazal-ur-; Khiari, F.Z.; Aksoy, A.; Nassar, R

2000-12-01

Nuclear track detectors were used to measure the longitudinal and transverse distributions of slow neutrons in a moderated neutron field as well as the longitudinal and transverse distributions of fast neutrons produced on the 0 deg. beam line of the KFUPM 350 keV ion accelerator. The neutrons were first produced from the T(d,n){sup 4}He reaction with a neutron energy of approximately 14 MeV and were then moderated in a cylindrical polyethylene moderator placed at the end of the 0 deg. beam line. The optimal transverse slow neutron distribution was found to be uniform within {+-}4.5% at a 3 cm depth inside the moderator. The fast neutron distribution component along the moderator central axis exhibited an exponential-like drop in intensity with depth. Linearity checks of alpha and proton recoil track density with irradiation time for the nuclear track detectors were verified for both slow and fast neutrons.

System and apparatus for neutron radiography

International Nuclear Information System (INIS)

Whittemore, W.L.

1991-01-01

This patent describes a neutron radiography apparatus. It comprises an imaging plane; a neutron moderator having a cavity defining a convergent collimator, the cavity having a base and converging walls of neutron moderating material terminating at an aperture; a divergent collimator coaxially joined to the cavity at the aperture, the divergent collimator having diverging walls of radiation- absorbing material extending from the aperture to an expanded distal opening for irradiating the imaging plane; sources of neutrons disposed symmetrically about the base of the cavity; a neutron moderating material disposed for maximum neutron thermalization between the sources and the base of the cavity; and means for substantially shielding the plane from electromagnetic energy

Neutron beams for therapy

International Nuclear Information System (INIS)

Kuplenikov, Eh.L.; Dovbnya, A.N.; Telegin, Yu.N.; Tsymbal, V.A.; Kandybej, S.S.

2011-01-01

It was given the analysis and generalization of the study results carried out during some decades in many world countries on application of thermal, epithermal and fast neutrons for neutron, gamma-neutron and neutron-capture therapy. The main attention is focused on the practical application possibility of the accumulated experience for the base creation for medical research and the cancer patients effective treatment.

Recent advances in neutron tomography

International Nuclear Information System (INIS)

McFarland, E.; Massachusetts Inst. of Technology, Cambridge, MA; Lanza, R.

1993-01-01

Neutron imaging has been shown to be an excellent imaging tool for many nondestructive evaluation applications. Significantly improved contrast over X-ray images is possible for materials commonly found in engineering assemblies. The major limitations have been the neutron source and detection. A low cost, position sensitive neutron tomography detector system has been designed and built based on an electro-optical detector system using a LiF-ZnS scintillator screen and a cooled charge coupled device. This detector system can be used for neutron radiography as well as two and three-dimensional neutron tomography. Calculated performance of the system predicted near-quantum efficiency for position sensitive neutron detection. Experimental data was recently taken using this system at McClellan Air Force Base, Air Logistics Center, Sacramento, CA. With increased availability of low cost neutron sources and advanced image processing, neutron tomography will become an increasingly important nondestructive imaging method

Synovectomy by Neutron capture; Sinovectomia por captura de neutrones

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Torres M, C. [Centro Regional de Estudios Nucleares, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98000 Zacatecas (Mexico)

1998-12-31

The Synovectomy by Neutron capture has as purpose the treatment of the rheumatoid arthritis, illness which at present does not have a definitive curing. This therapy requires a neutron source for irradiating the articulation affected. The energy spectra and the intensity of these neutrons are fundamental since these neutrons induce nuclear reactions of capture with Boron-10 inside the articulation and the freely energy of these reactions is transferred at the productive tissue of synovial liquid, annihilating it. In this work it is presented the neutron spectra results obtained with moderator packings of spherical geometry which contains in its center a Pu{sup 239} Be source. The calculations were realized through Monte Carlo method. The moderators assayed were light water, heavy water base and the both combination of them. The spectra obtained, the average energy, the neutron total number by neutron emitted by source, the thermal neutron percentage and the dose equivalent allow us to suggest that the moderator packing more adequate is what has a light water thickness 0.5 cm (radius 2 cm) and 24.5 cm heavy water (radius 26.5 cm). (Author)

Exclusive data-based modeling of neutron-nuclear reactions below 20 MeV

Science.gov (United States)

Savin, Dmitry; Kosov, Mikhail

2017-09-01

We are developing CHIPS-TPT physics library for exclusive simulation of neutron-nuclear reactions below 20 MeV. Exclusive modeling reproduces each separate scattering and thus requires conservation of energy, momentum and quantum numbers in each reaction. Inclusive modeling reproduces only selected values while averaging over the others and imposes no such constraints. Therefore the exclusive modeling allows to simulate additional quantities like secondary particle correlations and gamma-lines broadening and avoid artificial fluctuations. CHIPS-TPT is based on the formerly included in Geant4 CHIPS library, which follows the exclusive approach, and extends it to incident neutrons with the energy below 20 MeV. The NeutronHP model for neutrons below 20 MeV included in Geant4 follows the inclusive approach like the well known MCNP code. Unfortunately, the available data in this energy region is mostly presented in ENDF-6 format and semi-inclusive. Imposing additional constraints on secondary particles complicates modeling but also allows to detect inconsistencies in the input data and to avoid errors that may remain unnoticed in inclusive modeling.

Online monitoring of fast neutron (DT/DD) at Purnima neutron generator

International Nuclear Information System (INIS)

Bishnoi, S.; Patel, T.; Shukla, M.; Adhikari, P.S.; Sinha, A.

2012-01-01

A neutron generator (NG) at Purnima Labs, BARC has been developed for DT accelerator driven zero power subcritical (ADSS) system. Subcritical core of ADSS will be coupled to the NG for benchmarking experiments. Kinetic parameters of ADSS such as K-source, flux, power etc depends on this external neutron source strength injected to the core. However the neutron emission rate of NG does not remain stable throughout its operation. In view of this a reliable, precise and online monitoring of NG's neutron emission rate is required. An online neutron monitoring system based on associated particle method has been designed, developed and installed at NG. The monitoring unit consists of an ion implanted planar silicon detector, placed inside the drift tube of NG at an angle with respect to D + beam direction. A series of experiments were carried out with increasing neutron yield to optimize the position of detector such that it has sufficient counting statistics and minimum pileup. A complementary calibration procedure for validating these results based on activation technique was also carried out with standard Cu foil. The reaction rate monitored with online monitor and foil activation technique were compared, their variations with the predicted (theoretical) results were within 16%. This paper deals with the development and performance of online neutron monitoring system for DT and DD neutrons

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

Neutron accident dosimeter based on SSNTDs

International Nuclear Information System (INIS)

Palfalvi, J.; Sajo-Bohus, L.

1998-01-01

A sandwich type track etch detector of CR-39 was developed utilizing neutron-proton recoil and (n,α) reactions. Applying gold and Cd filters this system turns into a threshold detector and also it combines the albedo and the direct detection methods; thus it becomes possible to detect neutrons in three or more energy ranges depending on the number of gold degraders of different thickness allowing dose assessment with an uncertainty of about 20%, as blank tests have proved when a single gold foil of 20 μm thick was used. (author)

Accelerator-Based Boron Neutron Capture Therapy and the Development of a Dedicated Tandem-Electrostatic-Quadrupole

International Nuclear Information System (INIS)

Kreiner, A. J.; Di Paolo, H.; Burlon, A. A.; Valda, A. A.; Debray, M. E.; Somacal, H. R.; Minsky, D. M.; Kesque, J. M.; Giboudot, Y.; Levinas, P.; Fraiman, M.; Romeo, V.

2007-01-01

There is a generalized perception that the availability of suitable particle accelerators installed in hospitals, as neutron sources, may be crucial for the advancement of Boron Neutron Capture Therapy (BNCT). Progress on an ongoing project to develop a Tandem-ElectroStatic-Quadrupole (TESQ) accelerator for Accelerator-Based (AB)-BNCT is described here. The project goal is a machine capable of delivering 30 mA of 2.5 MeV protons to be used in conjunction with a neutron production target based on the 7 Li(p,n) 7 Be reaction slightly beyond its resonance at 2.25 MeV. A folded tandem, with 1.25 MV terminal voltage, combined with an ESQ chain is being designed and constructed. A 30 mA proton beam of 2.5 MeV are the specifications needed to produce sufficiently intense and clean epithermal neutron beams, based on the 7 Li(p,n) 7 Be reaction, to perform BNCT treatment for deep-seated tumors in less than an hour. The first design and construction of an ESQ module is discussed and its electrostatic fields are investigated theoretically and experimentally. Also new beam transport calculations through the accelerator are presented

Response matrix calculation of a Bonner Sphere Spectrometer using ENDF/B-VII libraries

Energy Technology Data Exchange (ETDEWEB)

Morató, Sergio; Juste, Belén; Miró, Rafael; Verdú, Gumersindo [Instituto de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain); Guardia, Vicent, E-mail: bejusvi@iqn.upv.es [GD Energy Services, Valencia (Spain). Grupo dominguis

2017-07-01

The present work is focused on the reconstruction of a neutron spectra using a multisphere spectrometer also called Bonner Spheres System (BSS). To that, the determination of the response detector curves is necessary therefore we have obtained the response matrix of a neutron detector by Monte Carlo (MC) simulation with MCNP6 where the use of unstructured mesh geometries is introduced as a novelty. The aim of these curves was to study the theoretical response of a widespread neutron spectrometer exposed to neutron radiation. A neutron detector device has been used in this work which is formed by a multispheres spectrometer (BSS) that uses 6 high density polyethylene spheres with different diameters. The BSS consists of a set of 0.95 g/cm{sup 3} high density polyethylene spheres. The detector is composed of a lithium iodide 6LiI cylindrical scintillator crystal 4mm x 4mm size LUDLUM Model 42 coupled to a photomultiplier tube. Thermal tables are required to include polyethylene cross section in the simulation. These data are essential to get correct and accurate results in problems involving neutron thermalization. Nowadays available literature present the response matrix calculated with ENDF.B.V cross section libraries (V.Mares et al 1993) or with ENDF.B.VI (R.Vega Carrillo et al 2007). This work uses two novelties to calculate the response matrix. On the one hand the use of unstructured meshes to simulate the geometry of the detector and the Bonner Spheres and on the other hand the use of the updated ENDF.B.VII cross sections libraries. A set of simulations have been performed to obtain the detector response matrix. 29 mono energetic neutron beams between 10 KeV to 20 MeV were used as source for each moderator sphere up to a total of 174 simulations. Each mono energetic source was defined with the same diameter as the moderating sphere used in its corresponding simulation and the spheres were uniformly irradiated from the top of the photomultiplier tube. Some

The IAEA collaborating centre for neutron activation based methodologies of research reactors

International Nuclear Information System (INIS)

Bode, P.

2010-01-01

The Reactor Institute Delft of the Delft University of Technology houses the Netherlands' only academic nuclear research reactor, with associated instrumentation and laboratories, for scientific education and research with ionizing radiation. The Institute's swimming pool type research reactor reached first criticality in 1963 and is currently operated at 2MW thermal powers on a 100 h/week basis. The reactor is equipped with neutron mirror guides serving ultra modern neutron beam physics instruments and with a very bright positron facility. Fully automated gamma-ray spectrometry systems are used by the laboratory for neutron activation analysis, providing large scale services under an ISO/IEC 17025:2005 compliant management system, being (since 1993) the first accredited laboratory of its kind in the world. Already for several years, this laboratory is sustainable by rendering these services to both the public and the private sector. The prime user of the Institute's fac ilities is the scientific Research Department of Radiation, Radionuclide and Reactors of the Faculty of Applied Sciences, housed inside the building. All reactor facilities are also made available for use by or for services to, external clients (industry, government, private sector, other (international research institutes and universities). The Reactor Institute Delft was inaugurated in May 2009 as a new lAEA Collaborating Centre for Neutron Activation Based Methodologies of Research Reactors. The collaboration involves education, research and development in (I) Production of reactor-produced, no-carrier added radioisotopes of high specific activity via neutron activation; (II) Neutron activation analysis with emphasis on automation as well as analysis of large samples, and radiotracer techniques and as a cross-cutting activity, (IIl) Quality assurance and management in research and application of research reactor based techniques and in research reactor operations. This c ollaboration will

Measurement channel of neutron flow based on software; Canal de medicion de flujo neutronico basado en software

Energy Technology Data Exchange (ETDEWEB)

Rivero G, T.; Benitez R, J. S. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: trg@nuclear.inin.mx

2008-07-01

The measurement of the thermal power in nuclear reactors is based mainly on the measurement of the neutron flow. The presence of these in the reactor core is associated to neutrons released by the fission reaction of the uranium-235. Once moderate, these neutrons are precursors of new fissions. This process it is known like chain reaction. Thus, the power to which works a nuclear reactor, he is proportional to the number of produced fissions and as these depend on released neutrons, also the power is proportional to the number of present neutrons. The measurement of the thermal power in a reactor is realized with called instruments nuclear channels. To low power (level source), these channels measure the individual counts of detected neutrons, whereas to a medium and high power, they measure the electrical current or fluctuation of the same one that generate the fission neutrons in ionization chambers especially designed to detect neutrons. For the case of TRIGA reactors, the measurement channels of neutron flow use discreet digital electronic technology makes some decades already. Recently new technological tools have arisen that allow developing new versions of nuclear channels of simple form and compacts. The present work consists of the development of a nuclear channel for TRIGA reactors based on the use of the correlated signal of a fission chamber for ample interval. This new measurement channel uses a data acquisition card of high speed and the data processing by software that to the being installed in a computer is created a virtual instrument, with what spreads in real time, in graphic and understandable form for the operator, the power indication to which it operates the nuclear reactor. This system when being based on software, offers a major versatility to realize changes in the signal processing and power monitoring algorithms. The experimental tests of neutronic power measurement show a reliable performance through seven decades of power, with a

An accelerator-based Boron Neutron Capture Therapy (BNCT) facility based on the 7Li(p,n)7Be

Science.gov (United States)

Musacchio González, Elizabeth; Martín Hernández, Guido

2017-09-01

BNCT (Boron Neutron Capture Therapy) is a therapeutic modality used to irradiate tumors cells previously loaded with the stable isotope 10B, with thermal or epithermal neutrons. This technique is capable of delivering a high dose to the tumor cells while the healthy surrounding tissue receive a much lower dose depending on the 10B biodistribution. In this study, therapeutic gain and tumor dose per target power, as parameters to evaluate the treatment quality, were calculated. The common neutron-producing reaction 7Li(p,n)7Be for accelerator-based BNCT, having a reaction threshold of 1880.4 keV, was considered as the primary source of neutrons. Energies near the reaction threshold for deep-seated brain tumors were employed. These calculations were performed with the Monte Carlo N-Particle (MCNP) code. A simple but effective beam shaping assembly (BSA) was calculated producing a high therapeutic gain compared to previously proposed facilities with the same nuclear reaction.

A wide dynamic range BF3 neutron monitor with front-end electronics based on a logarithmic amplifier

International Nuclear Information System (INIS)

Ferrarini, M.; Varoli, V.; Favalli, A.; Caresana, M.; Pedersen, B.

2010-01-01

This paper describes a wide dynamic range neutron monitor based on a BF 3 neutron detector. The detector is used in current mode, and front-end electronics based on a logarithmic amplifier are used in order to have a measurement capability ranging over many orders of magnitude. The system has been calibrated at the Polytechnic of Milan, CESNEF, with an AmBe neutron source, and has been tested in a pulsed field at the PUNITA facility at JRC, Ispra. The detector has achieved a dynamic range of over 6 orders of magnitude, being able to measure single neutron pulses and showing saturation-free response for a reaction rate up to 10 6 s -1 . It has also proved effective in measuring the PUNITA facility pulse integral fluence.

Neutron transportation simulator

International Nuclear Information System (INIS)

Uenohara, Yuzo.

1995-01-01

In the present invention, problems in an existent parallelized monte carlo method is solved, and behaviors of neutrons in a large scaled system are accurately simulated at a high speed. Namely, a neutron transportation simulator according to the monte carlo method simulates movement of each of neutrons by using a parallel computer. In this case, the system to be processed is divided based on a space region and an energy region to which neutrons belong. Simulation of neutrons in the divided regions is allotted to each of performing devices of the parallel computer. Tarry data and nuclear data of the neutrons in each of the regions are memorized dispersedly to memories of each of the performing devices. A transmission means for simulating the behaviors of the neutrons in the region by each of the performing devices, as well as transmitting the information of the neutrons, when the neutrons are moved to other region, to the performing device in a transported portion are disposed to each of the performing devices. With such procedures, simulation for the neutrons in the allotted region can be conducted with small capacity of memories. (I.S.)

Present status of fast neutron personnel dosimetry system based on CR-39 solid state nuclear track detectors

International Nuclear Information System (INIS)

Pal, Rupali; Sathian, Deepa; Jayalakshmi, V.; Bakshi, A.K.; Chougaonkar, M.P.; Mayya, Y.S.; Kumar, Valli; Babu, Rajesh; Kar, S.; Joshi, V.M.

2011-08-01

Neutron sources are of different types depending upon the method of production such as nuclear reactors, particle accelerators and laboratory sources. Neutron sources depending upon their energy, flux, size etc. are used for variety of applications in basic and applied sciences, neutron scattering experiments and in industry such as oil well - digging, coal mining and processing, ore processing etc. Personnel working in nuclear installations such as reactors, accelerators, spent fuel processing plants, nuclear fuel cycle operations and those working in various industries such as oil refining, oil well-digging, coal mining and processing, ore processing, etc. need to be monitored for neutron exposures, if any. Neutron monitoring is especially necessary in view of the fact that the radiation weighting factor for neutron is much higher than gamma rays and also it varies with energy. Radiological Physics and Advisory Division is involved in monitoring of personnel working in neutron fields. Around 2100 workers from 70 institutions (DAE and Non-DAE) are monitored on a quarterly basis. Neutron personnel monitoring, carried out in the country is based on Solid State Nuclear Track Detection (SSNTD) technique. In this technique, neutrons interact with hydrogen in CR-39 polymer to produce recoil protons. These protons create damages in the polymer, which are enlarged and appear as tracks when subjected to electrochemical etching (ECE). These tracks are counted in an optical system to evaluate the neutron dose. The neutron dosimetry system based on SSNTD has undergone a significant development, since it was started in 1990. The development includes upgradation of image analysis system for counting tracks, introduction of chemical etching (CE) at elevated temperatures for evaluation of dose equivalents above 10 mSv and use of carbon laser for cutting of CR-39 detectors. The entire dose evaluation process has been standardized, which includes calibration and performance tests

Proposal of a New Method for Neutron Dosimetry Based on Spectral Information Obtained by Application of Artificial Neural Networks

International Nuclear Information System (INIS)

Fehrenbacher, G.; Schuetz, R.; Hahn, K.; Sprunck, M.; Cordes, E.; Biersack, J.P.; Wahl, W.

1999-01-01

A new method for the monitoring of neutron radiation is proposed. It is based on the determination of spectral information on the neutron field in order to derive dose quantities like the ambient dose equivalent, the dose equivalent, or other dose quantities which depend on the neutron energy. The method uses a multi-element system consisting of converter type silicon detectors. The unfolding procedure is based on an artificial neural network (ANN). The response function of each element is determined by a computational model considering the neutron interaction with the dosemeter layers and the subsequent transport of produced ions. An example is given for a multi-element system. The ANN is trained by a given set of neutron spectra and then applied to count responses obtained in neutron fields. Four examples of spectra unfolded using the ANN are presented. (author)

Neutron excess generation by fusion neutron source for self-consistency of nuclear energy system

International Nuclear Information System (INIS)

Saito, Masaki; Artisyuk, V.; Chmelev, A.

1999-01-01

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities. (author)

Methods of neutron spectrometry

International Nuclear Information System (INIS)

Doerschel, B.

1981-01-01

The different methods of neutron spectrometry are based on the direct measurement of neutron velocity or on the use of suitable energy-dependent interaction processes. In the latter case the measuring effect of a detector is connected with the searched neutron spectrum by an integral equation. The solution needs suitable unfolding procedures. The most important methods of neutron spectrometry are the time-of-flight method, the crystal spectrometry, the neutron spectrometry by use of elastic collisions with hydrogen nuclei, and neutron spectrometry with the aid of nuclear reactions, especially of the neutron-induced activation. The advantages and disadvantages of these methods are contrasted considering the resolution, the measurable energy range, the sensitivity, and the experimental and computational efforts. (author)

Grazing Incidence Neutron Optics

Science.gov (United States)

Gubarev, Mikhail V. (Inventor); Ramsey, Brian D. (Inventor); Engelhaupt, Darell E. (Inventor)

2013-01-01

Neutron optics based on the two-reflection geometries are capable of controlling beams of long wavelength neutrons with low angular divergence. The preferred mirror fabrication technique is a replication process with electroform nickel replication process being preferable. In the preliminary demonstration test an electroform nickel optics gave the neutron current density gain at the focal spot of the mirror at least 8 for neutron wavelengths in the range from 6 to 20.ANG.. The replication techniques can be also be used to fabricate neutron beam controlling guides.

Neutron irradiation effects on magnetic properties of Fe-based ferromagnetic metallic glasses

International Nuclear Information System (INIS)

Miglierini, M.; Nasu, Saburo; Skorvanek, I.; Sitek, J.

1992-01-01

Transmission 57 Fe Moessbauer spectroscopy, J-H quasistatic hysteresis loop and AC susceptibility measurements are used to study effects of neutron irradiation on magnetic properties of Fe-based-ferromagnetic metallic glasses. Elastic stress centers are produced during the process of neutron irradiation as a result of atom mixing. Rearrangement of the atoms causes changes in the average value of the hyperfine field distribution and orientation of the net magnetic moment. They are shown to depend on the composition of the investigated samples. Cr-doped metallic glasses depict transition from the ferromagnetic to paramagnetic state at room temperature after neutron irradiation implying changes in the Curie temperature. The presence of Ni in the samples reduces the effects of radiation damage as revealed also from position lifetime data. Possible sources of a radiation damage are discussed using the results of γ-ray spectroscopy. (author)

Neutron irradiation effects on magnetic properties of Fe-based ferromagnetic metallic glasses

Energy Technology Data Exchange (ETDEWEB)

Miglierini, M.; Nasu, Saburo (Osaka Univ., Toyonaka (Japan). Faculty of Science); Skorvanek, I.; Sitek, J.

1992-04-01

Transmission {sup 57}Fe Moessbauer spectroscopy, J-H quasistatic hysteresis loop and AC susceptibility measurements are used to study effects of neutron irradiation on magnetic properties of Fe-based-ferromagnetic metallic glasses. Elastic stress centers are produced during the process of neutron irradiation as a result of atom mixing. Rearrangement of the atoms causes changes in the average value of the hyperfine field distribution and orientation of the net magnetic moment. They are shown to depend on the composition of the investigated samples. Cr-doped metallic glasses depict transition from the ferromagnetic to paramagnetic state at room temperature after neutron irradiation implying changes in the Curie temperature. The presence of Ni in the samples reduces the effects of radiation damage as revealed also from position lifetime data. Possible sources of a radiation damage are discussed using the results of {gamma}-ray spectroscopy. (author).

Development of a SQUID-based 3He Co-magnetometer Readout for a Neutron Electric Dipole Moment Experiment

OpenAIRE

Kim, Young Jin; Clayton, Steven M.

2012-01-01

A discovery of a permanent electric dipole moment (EDM) of the neutron would provide one of the most important low energy tests of the discrete symmetries beyond the Standard Model of particle physics. A new search of neutron EDM, to be conducted at the spallation neutron source (SNS) at ORNL, is designed to improve the present experimental limit of ~10^-26 e-cm by two orders of magnitude. The experiment is based on the magnetic-resonance technique in which polarized neutrons precess at the L...

Neutron detector based on lithiated sol-gel glass

CERN Document Server

Wallace, S; Miller, L F; Dai, S

2002-01-01

A neutron detector technology is demonstrated based on sup 6 Li/ sup 1 sup 0 B doped sol-gel glass. The detector is a sol-gel glass film coated silicon surface barrier detector (SBD). The ionized charged particles from (n, alpha) reactions in the sol-gel film enter the SBD and are counted. Data showing that gamma-ray pulse amplitudes interfere with identifying charged particles that exit the film layer with energies below the gamma-ray energy is presented. Experiments were performed showing the effect of sup 1 sup 3 sup 7 Cs and sup 6 sup 0 Co gamma rays on the SBD detector. The reaction product energies of the triton and alpha particles from sup 6 Li are significantly greater than the energies of the Compton electrons from high-energy gamma rays, allowing the measurement of neutrons in a high gamma background. The sol-gel radiation detection technology may be applicable to the characterization of transuranic waste, spent nuclear fuel and to the monitoring of stored plutonium.

Calibration of a special neutron dosemeter based on solid-state track detectors and fission radiators in various neutron fields

International Nuclear Information System (INIS)

Doerschel, B.; Krusche, M.; Schuricht, V.

1980-01-01

The calibration of a personnel neutron dosemeter in different neutron fields is described. The badge-like dosemeter contains 5 detectors: polycarbonate foil (10 μm, Makrofol KG), 232 Th, natural uranium, natural uranium with boron, and natural uranium with cadmium. Detector sensitivity and calibration factors have been calculated and measured in radiation fields of 252 Cf fission neutrons, WWR-S reactor neutrons with and without Cd and Fe shielding, 3-MeV (d,t) generator neutrons, and 238 PuBe neutrons. Measurement range and achievable accuracy are discussed from the point of view of applying the dosemeter in routine and emergency uses

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.

Improved neutron-gamma discrimination for a 3He neutron detector using subspace learning methods

Science.gov (United States)

Wang, C. L.; Funk, L. L.; Riedel, R. A.; Berry, K. D.

2017-05-01

3He gas based neutron Linear-Position-Sensitive Detectors (LPSDs) have been used for many neutron scattering instruments. Traditional Pulse-height Analysis (PHA) for Neutron-Gamma Discrimination (NGD) resulted in the neutron-gamma efficiency ratio (NGD ratio) on the order of 105-106. The NGD ratios of 3He detectors need to be improved for even better scientific results from neutron scattering. Digital Signal Processing (DSP) analyses of waveforms were proposed for obtaining better NGD ratios, based on features extracted from rise-time, pulse amplitude, charge integration, a simplified Wiener filter, and the cross-correlation between individual and template waveforms of neutron and gamma events. Fisher Linear Discriminant Analysis (FLDA) and three Multivariate Analyses (MVAs) of the features were performed. The NGD ratios are improved by about 102-103 times compared with the traditional PHA method. Our results indicate the NGD capabilities of 3He tube detectors can be significantly improved with subspace-learning based methods, which may result in a reduced data-collection time and better data quality for further data reduction.

Neutron generator based on adiabatic trap

International Nuclear Information System (INIS)

Golovin, I.N.; Zhil'tsov, V.A.; Panov, D.A.; Skovoroda, A.A.; Shatalov, G.E.; Shcherbakov, A.G.

1988-01-01

A possibility of 14 MeV neutron generator (NG) production on the basis of axial-symmetric adiabatic trap with MHD cusped armature for the testing of materials and elements of the DT reactor first wall and blanket structure is discussed. General requirements to NG are formulated. It is shown that the NG variant discussed meets the requirements formulated. Approximate calculation of the NG parameters has shown that total energy consumption by the generator does not exceed 220 MW at neutron flux specific capacity of 2.5 MW/m 2 and radiation test area of 5-6 m 2

Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem

International Nuclear Information System (INIS)

William Charlton

2007-01-01

Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions

A novel wide range, real-time neutron fluence monitor based on commercial off the shelf gallium arsenide light emitting diodes

Energy Technology Data Exchange (ETDEWEB)

Mukherjee, B., E-mail: bhaskar.mukherjee@uk-essen.de [Westdeutsches Protonentherapiezentrum Essen (WPE) gGmbH, Hufelandstrasse 55, D-45147 Essen (Germany); Hentschel, R. [Strahlenklinik, University Hospital Essen (Germany); Lambert, J. [Westdeutsches Protonentherapiezentrum Essen (WPE) gGmbH, Hufelandstrasse 55, D-45147 Essen (Germany); Deya, W. [Strahlenklinik, University Hospital Essen (Germany); Farr, J. [Westdeutsches Protonentherapiezentrum Essen (WPE) gGmbH, Hufelandstrasse 55, D-45147 Essen (Germany)

2011-10-01

Displacement damage produced by high-energy neutrons in gallium arsenide (GaAs) light emitting diodes (LED) results in the reduction of light output. Based on this principle we have developed a simple, cost effective, neutron detector using commercial off the shelf (COTS) GaAs-LED for the assessment of neutron fluence and KERMA at critical locations in the vicinity of the 230 MeV proton therapy cyclotron operated by Westdeutsches Protonentherapiezentrum Essen (WPE). The LED detector response (mV) was found to be linear within the neutron fluence range of 3.0x10{sup 8}-1.0x10{sup 11} neutron cm{sup -2}. The response of the LED detector was proportional to neutron induced displacement damage in LED; hence, by using the differential KERMA coefficient of neutrons in GaAs, we have rescaled the calibration curve for two mono-energetic sources, i.e. 1 MeV neutrons and 14 MeV neutrons generated by D+T fusion reaction. In this paper we present the principle of the real-time GaAs-LED based neutron fluence monitor as mentioned above. The device was calibrated using fast neutrons produced by bombarding a thick beryllium target with 14 MeV deuterons from a TCC CV 28 medical cyclotron of the Strahlenklinik University Hospital Essen.

Additive effect of BPA and Gd-DTPA for application in accelerator-based neutron source

International Nuclear Information System (INIS)

Yoshida, F.; Yamamoto, T.; Nakai, K.; Zaboronok, A.; Matsumura, A.

2015-01-01

Because of its fast metabolism gadolinium as a commercial drug was not considered to be suitable for neutron capture therapy. We studied additive effect of gadolinium and boron co-administration using colony forming assay. As a result, the survival of tumor cells with additional 5 ppm of Gd-DTPA decreased to 1/10 compared to the cells with boron only. Using gadolinium to increase the effect of BNCT instead of additional X-ray irradiation might be beneficial, as such combination complies with the short-time irradiation regimen at the accelerator-based neutron source. - Highlights: • Gd-DTPA is widely clinically used as a contrast medium for MRI. • Shift to an accelerator-based neutron source is advantageous for gadolinium NCT. • Boron–gadolinium NCT effects on tumor cell lines were significant. • Additional administration of Gd-DTPA might enhance the effect of BPA–BNCT.

Interaction of Water with Cement Based Repository Materials - Application of Neutron Imaging

International Nuclear Information System (INIS)

Mcglinn, P.J.; Brew, D.R.M.; Beer, F.C. De; Radebe, M.J.; Nshimirimana, R.

2013-01-01

Cementitious materials are conventionally used in conditioning intermediate and low level radioactive waste. In this study, a candidate cement-based wasteform and a series of barrier materials have been investigated using neutron imaging to: 1) characterise the wasteform for disposal in a repository for radioactive materials, and 2) characterise the compositon of the barrier materials in assessing their potential to transmit water. Imaging showed both the pore size distribution and the extent of the cracking that had occurred in the wasteform samples. The rate of the water penetration measured both by conventional sorptivity measurements and neutron imaging was greater than in pastes made from Ordinary Portland Cement. The ability of the cracks to distribute the water through the sample in a very short time was also evident. Macro-pore volume distributions of barrier samples, also acquired using neutron tomography, are shown to relate to water/cement ratio, composition and sorptivity data. The study highlights the significant potential of neutron imaging in the investigation of cementitious materials. The technique has the advantage of visualising and measuring, non-destructively, material distribution within macroscopic samples and is particularly useful in defining movement of water through the cementitious materials. (author)

Tests of a silicon wafer based neutron collimator

International Nuclear Information System (INIS)

Cussen, L.D.; Vale, C.J.; Anderson, I.S.; Hoeghoj, P.

2001-01-01

A Soller slit neutron collimator has been prepared by stacking 160 μm thick single crystal silicon wafers coated on one surface with 4 μm of gadolinium metal. The collimator has an angular width of 20 min full width at half maximum and an effective length of 2.75 cm. The collimator has beam dimensions of 1 cm wide by 5.3 cm high. Tests at neutron wavelengths 7.5A and 1.8A showed a peak transmission of 88% within 2% of the optimum theoretical possibility. The background suppression in the wings is comparable with that of conventional neutron collimators

Tests of a silicon wafer based neutron collimator

CERN Document Server

Cussen, L D; Anderson, I S; Hoeghoj, P

2001-01-01

A Soller slit neutron collimator has been prepared by stacking 160 mu m thick single crystal silicon wafers coated on one surface with 4 mu m of gadolinium metal. The collimator has an angular width of 20 min full width at half maximum and an effective length of 2.75 cm. The collimator has beam dimensions of 1 cm wide by 5.3 cm high. Tests at neutron wavelengths 7.5A and 1.8A showed a peak transmission of 88% within 2% of the optimum theoretical possibility. The background suppression in the wings is comparable with that of conventional neutron collimators.

Neutron multimonochromator-bipolarizer based on magnetic multilayer Fe/Co and new scheme for the total neutron polarization analysis

International Nuclear Information System (INIS)

Syromyatnikov, V.G.; Zaw Lin, Kyaw

2017-01-01

In this paper, we present a new neutron-optical element, Neutron Multimonochromator-Bipolarizer (NMB). It consists of a multimultilayer structure made of 12 periodic multilayer Fe/Co magnetic nanostructures whose period increases with distance from the substrate. Results are presented of calculations of the reflection coefficients from the NMB. We propose a new scheme of the total neutron polarization analysis for the time-of-flight method in the reflectometry. In this scheme, double NMB is used as a polarizer and there is no spin-flipper before the sample. NMB can be used in polarized neutron reflectometry, in SESANS, and for research of low-angle and inelastic scattering of polarized neutrons. (paper)

Prompt gamma-based neutron dosimetry for Am-Be and other workplace neutron spectra

International Nuclear Information System (INIS)

Udupi, Ashwini; Panikkath, Priyada; Sarkar, P.K.

2016-01-01

A new field-deployable technique for estimating the neutron ambient dose equivalent H*(10) by using the measured prompt gamma intensities emitted from borated high-density polyethylene (BHDPE) and the combination of normal HDPE and BHDPE with different configurations have been evaluated in this work. Monte Carlo simulations using the FLUKA code has been employed to calculate the responses from the prompt gammas emitted due to the monoenergetic neutrons interacting with boron, hydrogen, and carbon nuclei. A suitable linear combination of these prompt gamma responses (dose conversion coefficient (DCC)-estimated) is generated to approximate the International Commission on Radiological Protection provided DCC using the cross-entropy minimization technique. In addition, the shape and configurations of the HDPE and BHDPE combined system are optimized using the FLUKA code simulation results. The proposed method is validated experimentally, as well as theoretically, using different workplace neutron spectra with a satisfactory outcome. (author)

Calculations of neutron spectra after neutron-neutron scattering

Energy Technology Data Exchange (ETDEWEB)

Crawford, B E [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S L [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Howell, C R [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G E [North Carolina State University, Raleigh, NC 27695-8202 (United States); Tornow, W [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Furman, W I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Muzichka, A Yu [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V N [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation)

2004-09-01

A direct neutron-neutron scattering length, a{sub nn}, measurement with the goal of 3% accuracy (0.5 fm) is under preparation at the aperiodic pulsed reactor YAGUAR. A direct measurement of a{sub nn} will not only help resolve conflicting results of a{sub nn} by indirect means, but also in comparison to the proton-proton scattering length, a{sub pp}, shed light on the charge-symmetry of the nuclear force. We discuss in detail the analysis of the nn-scattering data in terms of a simple analytical expression. We also discuss calibration measurements using the time-of-flight spectra of neutrons scattered on He and Ar gases and the neutron activation technique. In particular, we calculate the neutron velocity and time-of-flight spectra after scattering neutrons on neutrons and after scattering neutrons on He and Ar atoms for the proposed experimental geometry, using a realistic neutron flux spectrum-Maxwellian plus epithermal tail. The shape of the neutron spectrum after scattering is appreciably different from the initial spectrum, due to collisions between thermal-thermal and thermal-epithermal neutrons. At the same time, the integral over the Maxwellian part of the realistic scattering spectrum differs by only about 6 per cent from that of a pure Maxwellian nn-scattering spectrum.

A genetic algorithm based method for neutron spectrum unfolding

International Nuclear Information System (INIS)

Suman, Vitisha; Sarkar, P.K.

2013-03-01

An approach to neutron spectrum unfolding based on a stochastic evolutionary search mechanism - Genetic Algorithm (GA) is presented. It is tested to unfold a set of simulated spectra, the unfolded spectra is compared to the output of a standard code FERDOR. The method was then applied to a set of measured pulse height spectrum of neutrons from the AmBe source as well as of emitted neutrons from Li(p,n) and Ag(C,n) nuclear reactions carried out in the accelerator environment. The unfolded spectra compared to the output of FERDOR show good agreement in the case of AmBe spectra and Li(p,n) spectra. In the case of Ag(C,n) spectra GA method results in some fluctuations. Necessity of carrying out smoothening of the obtained solution is also studied, which leads to approximation of the solution yielding an appropriate solution finally. Few smoothing techniques like second difference smoothing, Monte Carlo averaging, combination of both and gaussian based smoothing methods are also studied. Unfolded results obtained after inclusion of the smoothening criteria are in close agreement with the output obtained from the FERDOR code. The present method is also tested on a set of underdetermined problems, the outputs of which is compared to the unfolded spectra obtained from the FERDOR applied to a completely determined problem, shows a good match. The distribution of the unfolded spectra is also studied. Uncertainty propagation in the unfolded spectra due to the errors present in the measurement as well as the response function is also carried out. The method appears to be promising for unfolding the completely determined as well as underdetermined problems. It also has provisions to carry out the uncertainty analysis. (author)

Boron-coated straws as a replacement for {sup 3}He-based neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Lacy, Jeffrey L., E-mail: jlacy@proportionaltech.com [Proportional Technologies, Inc., 8022 El Rio Street, Houston, TX 77054 (United States); Athanasiades, Athanasios; Sun, Liang; Martin, Christopher S.; Lyons, Tom D.; Foss, Michael A.; Haygood, Hal B. [Proportional Technologies, Inc., 8022 El Rio Street, Houston, TX 77054 (United States)

2011-10-01

US and international government efforts to equip major seaports with large area neutron detectors, aimed to intercept the smuggling of nuclear materials, have precipitated a critical shortage of {sup 3}He gas. It is estimated that the annual demand of {sup 3}He for US security applications alone is more than the worldwide supply. This is strongly limiting the prospects of neutron science, safeguards, and other applications that rely heavily on {sup 3}He-based detectors. Clearly, alternate neutron detection technologies that can support large sensitive areas, and have low gamma sensitivity and low cost must be developed. We propose a low-cost technology based on long copper tubes (straws), coated on the inside with a thin layer of {sup 10}B-enriched boron carbide ({sup 10}B{sub 4}C). In addition to the high abundance of boron on Earth and low cost of {sup 10}B enrichment, the boron-coated straw (BCS) detector offers distinct advantages over conventional {sup 3}He-based detectors, and alternate technologies such as {sup 10}BF{sub 3} tubes and {sup 10}B-coated rigid tubes. These include better distribution inside moderator assemblies, many-times faster electronic signals, no pressurization, improved gamma-ray rejection, no toxic or flammable gases, and ease of serviceability. We present the performance of BCS detectors dispersed in a solid plastic moderator to address the need for portal monitoring. The design adopts the outer dimensions of currently deployed {sup 3}He-based monitors, but takes advantage of the small BCS diameter to achieve a more uniform distribution of neutron converter throughout the moderating material. We show that approximately 63 BCS detectors, each 205 cm long, distributed inside the moderator, can match or exceed the detection efficiency of typical monitors fitted with a 5 cm diameter {sup 3}He tube, 187 cm long, pressurized to 3 atm.

Gamma-Free Neutron Detector Based upon Lithium Phosphate Nanoparticles

International Nuclear Information System (INIS)

Steven Wallace

2007-01-01

A gamma-free neutron-sensitive scintillator is needed to enhance radiation sensing and detection for nonproliferation applications. Such a scintillator would allow very large detectors to be placed at the perimeter of spent-fuel storage facilities at commercial nuclear power plants, so that any movement of spontaneously emitted neutrons from spent nuclear fuel or weapons grade plutonium would be noted in real-time. This task is to demonstrate that the technology for manufacturing large panels of fluor-doped plastic containing lithium-6 phosphate nanoparticles can be achieved. In order to detect neutrons, the nanoparticles must be sufficiently small so that the plastic remains transparent. In this way, the triton and alpha particles generated by the capture of the neutron will result in a photon burst that can be coupled to a wavelength shifting fiber (WLS) producing an optical signal of about ten nanoseconds duration signaling the presence of a neutron emitting source

A neutron monitor for D-T neutron generator in the PGNAA-based online measurement system

Science.gov (United States)

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

2017-06-01

A new type of neutron detector, which consists of polyethylene, an EJ200 plastic scintillator and fused silica, was proposed and optimized by the GEANT4 Monte Carlo simulation toolkit in our previous studies. The calculation method was also described for calculating the neutron flux in the preset condition. This paper reports the manufacturing of the prototype detector. Experiments are conducted to validate the feasibility of this detector. A D-T neutron generator and a 60Co gamma-ray source are used in the experiments. The designed detector and a He-3 proportional counter are simultaneously used to monitor the yield of the D-T neutron generator. A more universal calculation method is developed to enable the application of this detector to common conditions. The experimental results show that the performance of the designed detector is comparable to that of the He-3 proportional counter. The relative deviations between their normalized counts are less than 5%.

ATR neutron spectral characterization

Energy Technology Data Exchange (ETDEWEB)

Rogers, J.W.; Anderl, R.A.

1995-11-01

The Advanced Test Reactor (ATR) at INEL provides intense neutron fields for irradiation-effects testing of reactor material samples, for production of radionuclides used in industrial and medical applications, and for scientific research. Characterization of the neutron environments in the irradiation locations of the ATR has been done by means of neutronics calculations and by means of neutron dosimetry based on the use of neutron activation monitors that are placed in the various irradiation locations. The primary purpose of this report is to present the results of an extensive characterization of several ATR irradiation locations based on neutron dosimetry measurements and on least-squares-adjustment analyses that utilize both neutron dosimetry measurements and neutronics calculations. This report builds upon the previous publications, especially the reference 4 paper. Section 2 provides a brief description of the ATR and it tabulates neutron spectral information for typical irradiation locations, as derived from the more historical neutron dosimetry measurements. Relevant details that pertain to the multigroup neutron spectral characterization are covered in section 3. This discussion includes a presentation on the dosimeter irradiation and analyses and a development of the least-squares adjustment methodology, along with a summary of the results of these analyses. Spectrum-averaged cross sections for neutron monitoring and for displacement-damage prediction in Fe, Cr, and Ni are given in section 4. In addition, section4 includes estimates of damage generation rates for these materials in selected ATR irradiation locations. In section 5, the authors present a brief discussion of the most significant conclusions of this work and comment on its relevance to the present ATR core configuration. Finally, detailed numerical and graphical results for the spectrum-characterization analyses in each irradiation location are provided in the Appendix.

Monte Carlo based dosimetry and treatment planning for neutron capture therapy of brain tumors

International Nuclear Information System (INIS)

Zamenhof, R.G.; Clement, S.D.; Harling, O.K.; Brenner, J.F.; Wazer, D.E.; Madoc-Jones, H.; Yanch, J.C.

1990-01-01

Monte Carlo based dosimetry and computer-aided treatment planning for neutron capture therapy have been developed to provide the necessary link between physical dosimetric measurements performed on the MITR-II epithermal-neutron beams and the need of the radiation oncologist to synthesize large amounts of dosimetric data into a clinically meaningful treatment plan for each individual patient. Monte Carlo simulation has been employed to characterize the spatial dose distributions within a skull/brain model irradiated by an epithermal-neutron beam designed for neutron capture therapy applications. The geometry and elemental composition employed for the mathematical skull/brain model and the neutron and photon fluence-to-dose conversion formalism are presented. A treatment planning program, NCTPLAN, developed specifically for neutron capture therapy, is described. Examples are presented illustrating both one and two-dimensional dose distributions obtainable within the brain with an experimental epithermal-neutron beam, together with beam quality and treatment plan efficacy criteria which have been formulated for neutron capture therapy. The incorporation of three-dimensional computed tomographic image data into the treatment planning procedure is illustrated. The experimental epithermal-neutron beam has a maximum usable circular diameter of 20 cm, and with 30 ppm of B-10 in tumor and 3 ppm of B-10 in blood, it produces a beam-axis advantage depth of 7.4 cm, a beam-axis advantage ratio of 1.83, a global advantage ratio of 1.70, and an advantage depth RBE-dose rate to tumor of 20.6 RBE-cGy/min (cJ/kg-min). These characteristics make this beam well suited for clinical applications, enabling an RBE-dose of 2,000 RBE-cGy/min (cJ/kg-min) to be delivered to tumor at brain midline in six fractions with a treatment time of approximately 16 minutes per fraction

A neutron dynamic therapy with a boron tracedrug UTX-51 using a compact neutron generator.

Science.gov (United States)

Hori, Hitoshi; Tada, Ryu; Uto, Yoshihiro; Nakata, Eiji; Morii, Takashi; Masuda, Kai

2014-08-01

We are developing a neutron dynamic therapy (NDT) with boron tracedrugs for a new mechanical-clearance treatment of pathotoxic misfolded, aggregated, and self-propagating prion-associated disease proteins. We present a compact neutron generator-based NDT using a boron tracedrug UTX-51. Our NDT is based on the weak thermal neutron-bombarded destructive action of UTX-51 on bovine serum albumin (BSA) using the neutron beams produced from a compact inertial electrostatic confinement fusion (IECF) neutron generator. BSA as an NDT molecular target was subjected to thermal neutron irradiation for eight hours using a compact neutron generator. The sodium dodecyl sulfate-polyacrylamide gel electrophoresis pattern showed no protein band when 2 nmoles of BSA were irradiated with more than 100 nmoles of UTX-51, while BSA was not affected when irradiated without UTX-51. For the first time, we have succeeded in the molecular destruction of a prion-disease model protein, BSA, by NDT with a boron tracedrug, UTX-51, using a compact neutron generator. Copyright© 2014 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.

Kriging-based algorithm for nuclear reactor neutronic design optimization

International Nuclear Information System (INIS)

Kempf, Stephanie; Forget, Benoit; Hu, Lin-Wen

2012-01-01

Highlights: ► A Kriging-based algorithm was selected to guide research reactor optimization. ► We examined impacts of parameter values upon the algorithm. ► The best parameter values were incorporated into a set of best practices. ► Algorithm with best practices used to optimize thermal flux of concept. ► Final design produces thermal flux 30% higher than other 5 MW reactors. - Abstract: Kriging, a geospatial interpolation technique, has been used in the present work to drive a search-and-optimization algorithm which produces the optimum geometric parameters for a 5 MW research reactor design. The technique has been demonstrated to produce an optimal neutronic solution after a relatively small number of core calculations. It has additionally been successful in producing a design which significantly improves thermal neutron fluxes by 30% over existing reactors of the same power rating. Best practices for use of this algorithm in reactor design were identified and indicated the importance of selecting proper correlation functions.

The neutron silicon lens. An update of the thermal neutron lens results

International Nuclear Information System (INIS)

Johnson, M.W.; Daymond, M.R.

2001-01-01

This paper introduces the concept of the Neutron Silicon Lens (NSL) and provides and update on the experimental results achieved to date. The NSL design is a cylindrical neutron lens based on the use of multiple neutron mirrors supported and separated by silicon wafers. Such lenses would have many applications in both the primary and scattered beams on neutron instruments, and would lead to immediate improvements where the sample to be illuminated is small, as in high pressure or engineering strain scanning instruments. (author)

The neutron silicon lens. An update of the thermal neutron lens results

Energy Technology Data Exchange (ETDEWEB)

Johnson, M.W.; Daymond, M.R. [Rutherford Appleton Laboratory, Chilton, Didcot, Oxfordshire (United Kingdom)

2001-03-01

This paper introduces the concept of the Neutron Silicon Lens (NSL) and provides and update on the experimental results achieved to date. The NSL design is a cylindrical neutron lens based on the use of multiple neutron mirrors supported and separated by silicon wafers. Such lenses would have many applications in both the primary and scattered beams on neutron instruments, and would lead to immediate improvements where the sample to be illuminated is small, as in high pressure or engineering strain scanning instruments. (author)

Options for a next generation neutron source for neutron scattering based on the projected linac facility at JAERI

International Nuclear Information System (INIS)

Mezei, F.; Watanabe, Noboru; Niimura, Nobuo; Morii, Yukio; Aizawa, Kazuya; Suzuki, Jun-ichi.

1997-03-01

Japan Atomic Energy Research Institute (JAERI) has a project to construct a high intensity proton accelerator to promote wide basic science using neutrons and nuclear power technologies such as radioactive nuclide transmutation. One of the most important field for utilization of neutron beam is neutron scattering. The energy and the averaged current obtained by the proton accelerator are 1.5 GeV and 4-5.3 mA, respectively and these provide 6-8 MW power. The repetition frequency is 50-60 Hz. Evaluation of options for the use of accelerators for neutron production for neutron scattering research and investigation of the neutron research opportunities offered by sharing the superconducting linac planned at JAERI were discussed. There are two ways of the utilization of proton beams for neutron scattering experiment. One is for long pulse spallation source (LPSS) and the other is for short pulse spallation source (SPSS). Quantitative evaluation of instrument performance with LPSS and SPSS was examined in the intensive discussion, calculations, workshop on this topics with Prof. F. Mezei who stayed at JAERI from October 24 to November 6, 1996. A report of the collaborative workshop will be also published separately. (author)

Small accelerator-based pulsed cold neutron sources

International Nuclear Information System (INIS)

Lanza, Richard C.

1997-09-01

Small neutron sources could be used by individual researchers with the convenience of an adequate local facility. Although these sources would produce lower fluxes than the national facilities, for selected applications, the convenience and availability may overcome the limitations on source strength. Such sources might also be useful for preliminary testing of ideas before going to a larger facility. Recent developments in small, high-current pulsed accelerators makes possible such a local source for pulsed cold neutrons.

A review of neutron scattering correction for the calibration of neutron survey meters using the shadow cone method

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang In; Kim, Bong Hwan; Kim, Jang Lyul; Lee, Jung Il [Health Physics Team, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-12-15

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a {sup 252}Californium ({sup 252}Cf) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1 - 9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

A review of neutron scattering correction for the calibration of neutron survey meters using the shadow cone method

International Nuclear Information System (INIS)

Kim, Sang In; Kim, Bong Hwan; Kim, Jang Lyul; Lee, Jung Il

2015-01-01

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a 252 Californium ( 252 Cf) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1 - 9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered

Development of high flux thermal neutron generator for neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-01

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

Investigation and optimisation of mobile NaI(Tl) and 3He-based neutron detectors for finding point sources

International Nuclear Information System (INIS)

Nilsson, Jonas M.C.; Finck, Robert R.; Rääf, Christopher

2015-01-01

Neutron radiation produces high-energy gamma radiation through (n,γ) reactions in matter. This can be used to detect neutron sources indirectly using gamma spectrometers. The sensitivity of a gamma spectrometer to neutrons can be amplified by surrounding it with polyvinyl chloride (PVC). The hydrogen in the PVC acts as a moderator and the chlorine emits prompt gammas when a neutron is captured. A 4.7-l 3 He-based mobile neutron detector was compared to a 4-l NaI(Tl)-detector covered with PVC using this principle. Methods were also developed to optimise the measurement parameters of the systems. The detector systems were compared with regard to their ability to find 241 AmBe, 252 Cf and 238 Pu– 13 C neutron sources. Results from stationary measurements were used to calculate optimal integration times as well as minimum detectable neutron emission rates. It was found that the 3 He-based detector was more sensitive to 252 Cf sources whereas the NaI(Tl) detector was more sensitive to 241 AmBe and 238 Pu– 13 C sources. The results also indicated that the sensitivity of the detectors to sources at known distances could theoretically be improved by 60% by changing from fixed integration times to list mode in mobile surveys

Polycapillary neutron lenses

International Nuclear Information System (INIS)

Mildner, D.F.R.

1997-01-01

The principle of multiple mirror reflection from smooth surfaces at small grazing angles enables the transport and guiding of high intensity slow neutron beams to locations of low background for neutron scattering and absorption experiments and to provide facilities for multiple instruments. Curved guides have been widely used at cold neutron facilities to remove the unwanted radiation (fast neutrons and gamma rays) from the beam without the use of filters. A typical guide has transverse dimensions of 50 mm and, with a radius of curvature of 1 km, transmits wavelengths longer than 5 A. Much tighter curves requires narrower transverse dimensions, otherwise there is little transmission. Typical neutron benders have a number of slots with transverse dimensions of ∼5 mm. Based on the same principle but using a different technology, recent developments in glass polycapillary fibers have produced miniature versions of neutron guides. Fibers with many thousands of channels having sizes of ∼ 10 μm enable beams of long wavelength neutrons (λ > 4 A) to be transmitted efficiently in a radius of curvature as small as a fraction of 1 m. A large collection of these miniature versions of neutron guides can be used to bend the neutron trajectories such that the incident beam can be focused. (author)

Neutron-gamma discrimination based on pulse shape discrimination in a Ce:LiCaAlF{sub 6} scintillator

Energy Technology Data Exchange (ETDEWEB)

Yamazaki, Atsushi, E-mail: a-yamazaki@nucl.nagoya-u.ac.jp [Department of Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University (Japan); Watanabe, Kenichi; Uritani, Akira [Department of Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University (Japan); Iguchi, Tetsuo [Department of Quantum Engineering, Graduate School of Engineering, Nagoya University (Japan); Kawaguchi, Noriaki [Tokuyama Corporation (Japan); Yanagida, Takayuki; Fujimoto, Yutaka; Yokota, Yuui; Kamada, Kei [Institute of Multidisciplinary Research for Advanced Materials (IMRAM), Tohoku University (Japan); Fukuda, Kentaro; Suyama, Toshihisa [Tokuyama Corporation (Japan); Yoshikawa, Akira [Institute of Multidisciplinary Research for Advanced Materials (IMRAM), Tohoku University (Japan); New Industry Creation Hatchery Center (NICHe), Tohoku University (Japan)

2011-10-01

We demonstrate neutron-gamma discrimination based on a pulse shape discrimination method in a Ce:LiCAF scintillator. We have tried neutron-gamma discrimination using a difference in the pulse shape or the decay time of the scintillation light pulse. The decay time is converted into the rise time through an integrating circuit. A {sup 252}Cf enclosed in a polyethylene container is used as the source of thermal neutrons and prompt gamma-rays. Obvious separation of neutron and gamma-ray events is achieved using the information of the rise time of the scintillation light pulse. In the separated neutron spectrum, the gamma-ray events are effectively suppressed with little loss of neutron events. The pulse shape discrimination is confirmed to be useful to detect neutrons with the Ce:LiCAF scintillator under an intense high-energy gamma-ray condition.

Neutron spectra characteristics for the intense neutron source, INS

International Nuclear Information System (INIS)

Battat, M.; Dierckx, R.; Emigh, C.R.

1977-01-01

The Intense Neutron Source, INS, facility is presently under construction at the Los Alamos Scientific Laboratory. Its purpose is to provide a broad base for research work related to the radiation effects produced by 14-MeV neutrons from a D-T burn of a fusion reactor. The INS facility produces a D-T burn-like reaction from the collision of an intense tritium-ion beam with a supersonic jet target of deuterium gas. The reaction produces a typical D-T 14-MeV neutron spectrum. By adding a fission blanket surrounding the D-T ''burn,'' the neutron spectral shape may be tailored to match almost perfectly the anticipated first-wall spectra from presently proposed fusion reactors. With a blanket in place, the total production of neutrons can be as large as 3 x 10 16 n/s and experimental volumes of the order of 1000 cm 3 can be available at flux levels greater than 0.6 x 10 14 n/cm 2 s

Neutron flux monitoring device

International Nuclear Information System (INIS)

Goto, Yasushi; Mitsubori, Minehisa; Ohashi, Kazunori.

1997-01-01

The present invention provides a neutron flux monitoring device for preventing occurrence of erroneous reactor scram caused by the elevation of the indication of a start region monitor (SRM) due to a factor different from actual increase of neutron fluxes. Namely, judgement based on measured values obtained by a pulse counting method and a judgment based on measured values obtained by a Cambel method are combined. A logic of switching neutron flux measuring method to be used for monitoring, namely, switching to an intermediate region when both of the judgements are valid is adopted. Then, even if the indication value is elevated based on the Cambel method with no increase of the counter rate in a neutron source region, the switching to the intermediate region is not conducted. As a result, erroneous reactor scram such as 'shorter reactor period' can be avoided. (I.S.)

Ultracold neutron detectors based on {sup 10}B converters used in the qBounce experiments

Energy Technology Data Exchange (ETDEWEB)

Jenke, Tobias, E-mail: tjenke@ati.ac.at [Atominstitut TU Wien, Stadionallee 2, 1020 Wien (Austria); Cronenberg, Gunther; Filter, Hanno [Atominstitut TU Wien, Stadionallee 2, 1020 Wien (Austria); Geltenbort, Peter [Institut Laue-Langevin, 6 rue Jules Horowitz, 38042 Grenoble Cedex 9 (France); Klein, Martin [Physikalisches Institut Universität Heidelberg, Im Neuenheimer Feld 226, 69120 Heidelberg (Germany); Lauer, Thorsten [FRM II, TU München, Lichtenbergstraße 1, 85748 Garching (Germany); Mitsch, Kevin [Atominstitut TU Wien, Stadionallee 2, 1020 Wien (Austria); Saul, Heiko [Atominstitut TU Wien, Stadionallee 2, 1020 Wien (Austria); FRM II, TU München, Lichtenbergstraße 1, 85748 Garching (Germany); Seiler, Dominik [Physik Department, TU München, James-Franck-Straße, 85748 Garching (Germany); Stadler, David [Physikalisches Institut Universität Heidelberg, Im Neuenheimer Feld 226, 69120 Heidelberg (Germany); Thalhammer, Martin [Atominstitut TU Wien, Stadionallee 2, 1020 Wien (Austria); Abele, Hartmut, E-mail: abele@ati.ac.at [Atominstitut TU Wien, Stadionallee 2, 1020 Wien (Austria); Physikalisches Institut Universität Heidelberg, Im Neuenheimer Feld 226, 69120 Heidelberg (Germany); Physik Department, TU München, James-Franck-Straße, 85748 Garching (Germany)

2013-12-21

Gravity experiments with very slow, so-called ultracold neutrons connect quantum mechanics with tests of Newton's inverse square law at short distances. These experiments face a low count rate and hence need highly optimized detector concepts. In the frame of this paper, we present low-background ultracold neutron counters and track detectors with micron resolution based on a {sup 10}B converter. We discuss the optimization of {sup 10}B converter layers, detector design and concepts for read-out electronics focusing on high-efficiency and low-background. We describe modifications of the counters that allow one to detect ultracold neutrons selectively on their spin-orientation. This is required for searches of hypothetical forces with spin–mass couplings. The mentioned experiments utilize a beam-monitoring concept which accounts for variations in the neutron flux that are typical for nuclear research facilities. The converter can also be used for detectors, which feature high efficiencies paired with high spatial resolution of 1–2μm. They allow one to resolve the quantum mechanical wave function of an ultracold neutron bound in the gravity potential above a neutron mirror.

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

Science.gov (United States)

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

2013-10-01

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

The energy spectrum of delayed neutrons from thermal neutron induced fission of 235U and its analytical approximation

International Nuclear Information System (INIS)

Doroshenko, A.Yu.; Tarasko, M.Z.; Piksaikin, V.M.

2002-01-01

The energy spectrum of the delayed neutrons is the poorest known of all input data required in the calculation of the effective delayed neutron fractions. In addition to delayed neutron spectra based on the aggregate spectrum measurements there are two different approaches for deriving the delayed neutron energy spectra. Both of them are based on the data related to the delayed neutron spectra from individual precursors of delayed neutrons. In present work these two different data sets were compared with the help of an approximation by gamma-function. The choice of this approximation function instead of the Maxwellian or evaporation type of distribution is substantiated. (author)

SiC-based neutron detector in quasi-realistic working conditions: efficiency and stability at room and high temperature under fast neutron irradiations

Energy Technology Data Exchange (ETDEWEB)

Ferone, Raffaello; Issa, Fatima; Ottaviani, Laurent; Biondo, Stephane; Vervisch, Vanessa [IM2NP, UMR CNRS 7334, Aix-Marseille University, Case 231,13397 Marseille Cedex 20, (France); Szalkai, Dora; Klix, Axel [KIT- Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology Karlsruhe 76344, (Germany); Vermeeren, Ludo [SCK-CEN, Boeretang 200, B-2400 Mol, (Belgium); Saenger, Richard [Schlumberger, Clamart, (France); Lyoussi, Abadallah [CEA, DEN, Departement d' Etudes des Reacteurs, Service de Physique Experimentale, Laboratoire Dosimetrie Capteurs Instrumentation, 13108 Saint-Paul-lez-Durance, (France)

2015-07-01

In the framework of the European I SMART project, we have designed and made new SiC-based nuclear radiation detectors able to operate in harsh environments and to detect both fast and thermal neutrons. In this paper, we report experimental results of fast neutron irradiation campaign at high temperature (106 deg. C) in quasi-realistic working conditions. Our device does not suffer from high temperature, and spectra do show strong stability, preserving features. These experiments, as well as others in progress, show the I SMART SiC-based device skills to operate in harsh environments, whereas other materials would strongly suffer from degradation. Work is still demanded to test our device at higher temperatures and to enhance efficiency in order to make our device fully exploitable from an industrial point of view. (authors)

Spectral correction factors for conventional neutron dosemeters used in high-energy neutron environments

International Nuclear Information System (INIS)

Lee, K.W.; Sheu, R.J.

2015-01-01

High-energy neutrons (>10 MeV) contribute substantially to the dose fraction but result in only a small or negligible response in most conventional moderated-type neutron detectors. Neutron dosemeters used for radiation protection purpose are commonly calibrated with 252 Cf neutron sources and are used in various workplace. A workplace-specific correction factor is suggested. In this study, the effect of the neutron spectrum on the accuracy of dose measurements was investigated. A set of neutron spectra representing various neutron environments was selected to study the dose responses of a series of Bonner spheres, including standard and extended-range spheres. By comparing 252 Cf-calibrated dose responses with reference values based on fluence-to-dose conversion coefficients, this paper presents recommendations for neutron field characterisation and appropriate correction factors for responses of conventional neutron dosemeters used in environments with high-energy neutrons. The correction depends on the estimated percentage of high-energy neutrons in the spectrum or the ratio between the measured responses of two Bonner spheres (the 4P6-8 extended-range sphere versus the 6'' standard sphere). (authors)

Neutron activation analysis at the Californium User Facility for Neutron Science

International Nuclear Information System (INIS)

Martin, R.C.; Smith, E.H.; Glasgow, D.C.; Jerde, E.A.; Marsh, D.L.; Zhao, L.

1997-12-01

The Californium User Facility (CUF) for Neutron Science has been established to provide 252 Cf-based neutron irradiation services and research capabilities including neutron activation analysis (NAA). A major advantage of the CUF is its accessibility and controlled experimental conditions compared with those of a reactor environment The CUF maintains the world's largest inventory of compact 252 Cf neutron sources. Neutron source intensities of ≤ 10 11 neutrons/s are available for irradiations within a contamination-free hot cell, capable of providing thermal and fast neutron fluxes exceeding 10 8 cm -2 s -1 at the sample. Total flux of ≥10 9 cm -2 s -1 is feasible for large-volume irradiation rabbits within the 252 Cf storage pool. Neutron and gamma transport calculations have been performed using the Monte Carlo transport code MCNP to estimate irradiation fluxes available for sample activation within the hot cell and storage pool and to design and optimize a prompt gamma NAA (PGNAA) configuration for large sample volumes. Confirmatory NAA irradiations have been performed within the pool. Gamma spectroscopy capabilities including PGNAA are being established within the CUF for sample analysis

The criteria for selecting a method for unfolding neutron spectra based on the information entropy theory

International Nuclear Information System (INIS)

Zhu, Qingjun; Song, Fengquan; Ren, Jie; Chen, Xueyong; Zhou, Bin

2014-01-01

To further expand the application of an artificial neural network in the field of neutron spectrometry, the criteria for choosing between an artificial neural network and the maximum entropy method for the purpose of unfolding neutron spectra was presented. The counts of the Bonner spheres for IAEA neutron spectra were used as a database, and the artificial neural network and the maximum entropy method were used to unfold neutron spectra; the mean squares of the spectra were defined as the differences between the desired and unfolded spectra. After the information entropy of each spectrum was calculated using information entropy theory, the relationship between the mean squares of the spectra and the information entropy was acquired. Useful information from the information entropy guided the selection of unfolding methods. Due to the importance of the information entropy, the method for predicting the information entropy using the Bonner spheres' counts was established. The criteria based on the information entropy theory can be used to choose between the artificial neural network and the maximum entropy method unfolding methods. The application of an artificial neural network to unfold neutron spectra was expanded. - Highlights: • Two neutron spectra unfolding methods, ANN and MEM, were compared. • The spectrum's entropy offers useful information for selecting unfolding methods. • For the spectrum with low entropy, the ANN was generally better than MEM. • The spectrum's entropy was predicted based on the Bonner spheres' counts

A wide dynamic range BF{sub 3} neutron monitor with front-end electronics based on a logarithmic amplifier

Energy Technology Data Exchange (ETDEWEB)

Ferrarini, M., E-mail: michele.ferrarini@polimi.i [Politecnico di Milano, Dipartimento Energia, via G. Ponzio 34/3, I-20133 Milano (Italy); Fondazione CNAO, via Caminadella 16, 20123 Milano (Italy); Varoli, V. [Politecnico di Milano, Dipartimento Energia, via G. Ponzio 34/3, I-20133 Milano (Italy); Favalli, A. [European Commission, Joint Research Centre, Institute for the Protection and Security of Citizen, TP 800, Via E. Fermi, 21027 Ispra (Vatican City State, Holy See) (Italy); Caresana, M. [Politecnico di Milano, Dipartimento Energia, via G. Ponzio 34/3, I-20133 Milano (Italy); Pedersen, B. [European Commission, Joint Research Centre, Institute for the Protection and Security of Citizen, TP 800, Via E. Fermi, 21027 Ispra (Italy)

2010-02-01

This paper describes a wide dynamic range neutron monitor based on a BF{sub 3} neutron detector. The detector is used in current mode, and front-end electronics based on a logarithmic amplifier are used in order to have a measurement capability ranging over many orders of magnitude. The system has been calibrated at the Polytechnic of Milan, CESNEF, with an AmBe neutron source, and has been tested in a pulsed field at the PUNITA facility at JRC, Ispra. The detector has achieved a dynamic range of over 6 orders of magnitude, being able to measure single neutron pulses and showing saturation-free response for a reaction rate up to 10{sup 6} s{sup -1}. It has also proved effective in measuring the PUNITA facility pulse integral fluence.

New Neutron Dosimeter

CERN Multimedia

2001-01-01

CERN has been operating an Individual Dosimetry Service for neutrons for about 35 years. The service was based on nuclear emulsions in the form of film packages which were developed and scanned in the Service. In 1999, the supplier of theses packages informed CERN that they will discontinue production of this material. TIS-RP decided to look for an external service provider for individual neutron dosimetry. After an extensive market survey and an invitation for tender, a supplier that met the stringent technical requirements set up by CERN's host states for personal dosimeters was identified. The new dosimeter is based on a track-etching technique. Neutrons have the capability of damaging plastic material. The microscopic damage centres are revealed by etching them in a strong acid. The resulting etch pits can be automatically counted and their density is proportional to dose equivalent from neutrons. On the technical side, the new dosimeter provides an improved independence of its response from energy and th...

Influence of Neutron Spectra Unfolding Method on Fast Neutron Dose Determination

International Nuclear Information System (INIS)

Marinkovic, P.

1991-01-01

Full text: Accuracy of knowing the fast neutron spectra has great influence on equivalent dose determination. In usual fast neutron spectrum measurements with scintillation detectors based on proton recoil, the main difficulty is confidence of unfolding method. In former ones variance of obtained result is usually great and negative values are possible too, which does means that we don't now exactly is obtained neutron spectrum real one. The new unfolding method based on Shanon's information theory, which gives non-negative spectrum and relative low variance, is obtained and appropriate numerical code for application in fast neutron spectrometry based on proton recoil is realized. In this method principle of maximum entropy and maximum likelihood are used together. Unknown group density distribution functions, which are considered as desired normalized mean neutron group flux, are constl u cted using only constrain of knowing mean value. Obtained distributions are consistent to available information (counts in NCA from proton recoil), while being maximally noncommittal with respect to all other unknown circumstances. For maximum likelihood principle, distribution functions around mean value of counts in the channels of MCA are taken to be Gauss function shape. Optimal non-negative solution is searched by means of Lagrange parameter method. Nonlinear system of equations, is solved using gradient and Newton iterative algorithm. Error covariance matrix is obtained too. (author)

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

Science.gov (United States)

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

2018-01-01

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

A GDT-based fusion neutron source for academic and industrial applications

Science.gov (United States)

Anderson, J. K.; Forest, C. B.; Mirnov, V. V.; Peterson, E. E.; Waleffe, R.; Wallace, J.; Harvey, R. W.

2017-10-01

The design of a fusion neutron source based on the gas dynamic trap (GDT) configuration is underway. The motivation is both the ends and the means. There are immediate applications for neutrons including medical isotope production and actinide burners. Taking the next step in the magnetic mirror path will leverage advances in high-temperature superconducting magnets and additive manufacturing in confining a fusion plasma, and both the technological and physics bases exist. Recent breakthrough results at the GDT facility in Russia demonstrate stable confinement of a beta 60% mirror plasma at high Te ( 1 keV). These scale readily to a fusion neutron source with an increase in magnetic field, mirror ratio, and ion energy. Studies of a next-step compact device focus on calculations of MHD equilibrium and stability, and Fokker-Planck modeling to optimize the heating scenario. The conceptualized device uses off-the-shelf MRI magnets for a 1 T central field, REBCO superconducting mirror coils (which can currently produce fields in excess of 30T), and existing 75 keV NBI and 140 GHz ECRH. High harmonic fast wave injection is damped on beam ions, dramatically increasing the fusion reactivity for an incremental bump in input power. MHD stability is achieved with the vortex confinement scheme, where a biasing profile imposes optimal ExB rotation of the plasma. Liquid metal divertors are being considered in the end cells. Work supported by the Wisconsin Alumni Research Foundation.

High-power liquid-lithium target prototype for accelerator-based boron neutron capture therapy.

Science.gov (United States)

Halfon, S; Paul, M; Arenshtam, A; Berkovits, D; Bisyakoev, M; Eliyahu, I; Feinberg, G; Hazenshprung, N; Kijel, D; Nagler, A; Silverman, I

2011-12-01

A prototype of a compact Liquid-Lithium Target (LiLiT), which will possibly constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals, was built. The LiLiT setup is presently being commissioned at Soreq Nuclear Research Center (SNRC). The liquid-lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power generated using a high-intensity proton beam (>10 kW), necessary for sufficient neutron flux. In off-line circulation tests, the liquid-lithium loop generated a stable lithium jet at high velocity, on a concave supporting wall; the concept will first be tested using a high-power electron beam impinging on the lithium jet. High intensity proton beam irradiation (1.91-2.5 MeV, 2-4 mA) will take place at Soreq Applied Research Accelerator Facility (SARAF) superconducting linear accelerator currently in construction at SNRC. Radiological risks due to the (7)Be produced in the reaction were studied and will be handled through a proper design, including a cold trap and appropriate shielding. A moderator/reflector assembly is planned according to a Monte Carlo simulation, to create a neutron spectrum and intensity maximally effective to the treatment and to reduce prompt gamma radiation dose risks. Copyright © 2011 Elsevier Ltd. All rights reserved.

Study of a spherical torus based volumetric neutron source for nuclear technology testing and development

International Nuclear Information System (INIS)

Cheng, E.T.; Cerbone, R.J.; Sviatoslavsky, I.N.; Galambos, L.D.; Peng, Y.-K.M.

2000-01-01

A plasma based, deuterium and tritium (DT) fueled, volumetric 14 MeV neutron source (VNS) has been considered as a possible facility to support the development of the demonstration fusion power reactor (DEMO). It can be used to test and develop necessary fusion blanket and divertor components and provide sufficient database, particularly on the reliability of nuclear components necessary for DEMO. The VNS device can be complement to ITER by reducing the cost and risk in the development of DEMO. A low cost, scientifically attractive, and technologically feasible volumetric neutron source based on the spherical torus (ST) concept has been conceived. The ST-VNS, which has a major radius of 1.07 m, aspect ratio 1.4, and plasma elongation three, can produce a neutron wall loading from 0.5 to 5 MW m -2 at the outboard test section with a modest fusion power level from 38 to 380 MW. It can be used to test necessary nuclear technologies for fusion power reactor and develop fusion core components include divertor, first wall, and power blanket. Using staged operation leading to high neutron wall loading and optimistic availability, a neutron fluence of more than 30 MW year m -2 is obtainable within 20 years of operation. This will permit the assessments of lifetime and reliability of promising fusion core components in a reactor relevant environment. A full scale demonstration of power reactor fusion core components is also made possible because of the high neutron wall loading capability. Tritium breeding in such a full scale demonstration can be very useful to ensure the self-sufficiency of fuel cycle for a candidate power blanket concept

Pulsed neutron sources at Dubna

International Nuclear Information System (INIS)

Shabalin, E.P.

1991-01-01

In 1960 the first world repetitively pulsed reactor IBR was put into operation. It was the beginning of the story how fission based pulsed neutron sources at Dubna have survived. The engineers involved have experienced many successes and failures in the course of new sources upgrading to finally come to possess the world's brightest neutron source - IBR-2. The details are being reviewed through the paper. The fission based pulsed neutron sources did not reach their final state as yet- the conceptual views of IBR prospects are being discussed with the goal to double the thermal neutron peak flux (up to 2x10 16 ) and to enhance the cold neutron flux by 10 times (with the present one being as high that of the ISIS cold moderator). (author)

New techniques in neutron scattering

International Nuclear Information System (INIS)

Hayter, J.B.

1993-01-01

New neutron sources being planned, such as the Advanced Neutron Source (ANS) or the European Spallation Source (ESS), will provide an order of magnitude flux increase over what is available today, but neutron scattering will still remain a signal-limited technique. At the same time, the development of new materials, such as polymer and ceramic composites or a variety of complex fluids, will increasingly require neutron-based research. This paper will discuss some of the new techniques which will allow us to make better use of the available neutrons, either through improved instrumentation or through sample manipulation. Discussion will center primarily on unpolarized neutron techniques since polarized neutrons will be the subject of the next paper. (author)

Neutron reflectometry

DEFF Research Database (Denmark)

Klösgen-Buchkremer, Beate Maria

2014-01-01

of desired information. In the course, an introduction into the method and an overview on selected instruments at large scale facilities will be presented. Examples will be given that illustrate the potential of the method, mostly based on organic films. Results from the investigation of layered films......Neutron (and X-ray) reflectometry constitute complementary interfacially sensitive techniques that open access to studying the structure within thin films of both soft and hard condensed matter. Film thickness starts oxide surfaces on bulk substrates, proceeding to (pauci-)molecular layers and up...... films or films with magnetic properties. The reason is the peculiar property of neutron light since the mass of a neutron is close to the one of a proton, and since it bears a magnetic moment. The optical properties of matter, when interacting with neutrons, are described by a refractive index...

Fission-neutrons source with fast neutron-emission timing

Energy Technology Data Exchange (ETDEWEB)

Rusev, G., E-mail: rusev@lanl.gov; Baramsai, B.; Bond, E.M.; Jandel, M.

2016-05-01

A neutron source with fast timing has been built to help with detector-response measurements. The source is based on the neutron emission from the spontaneous fission of {sup 252}Cf. The time is provided by registering the fission fragments in a layer of a thin scintillation film with a signal rise time of 1 ns. The scintillation light output is measured by two silicon photomultipliers with rise time of 0.5 ns. Overall time resolution of the source is 0.3 ns. Design of the source and test measurements using it are described. An example application of the source for determining the neutron/gamma pulse-shape discrimination by a stilbene crystal is given.

Neutron applications in materials for energy

CERN Document Server

Kearley, Gordon J

2015-01-01

Neutron Applications in Materials for Energy collects results and conclusions of recent neutron-based investigations of materials that are important in the development of sustainable energy. Chapters are authored by leading scientists with hands-on experience in the field, providing overviews, recent highlights, and case-studies to illustrate the applicability of one or more neutron-based techniques of analysis. The theme follows energy production, storage, and use, but each chapter, or section, can also be read independently, with basic theory and instrumentation for neutron scattering being

Prompt fission neutron spectra and average prompt neutron multiplicities

International Nuclear Information System (INIS)

Madland, D.G.; Nix, J.R.

1983-01-01

We present a new method for calculating the prompt fission neutron spectrum N(E) and average prompt neutron multiplicity anti nu/sub p/ as functions of the fissioning nucleus and its excitation energy. The method is based on standard nuclear evaporation theory and takes into account (1) the motion of the fission fragments, (2) the distribution of fission-fragment residual nuclear temperature, (3) the energy dependence of the cross section sigma/sub c/ for the inverse process of compound-nucleus formation, and (4) the possibility of multiple-chance fission. We use a triangular distribution in residual nuclear temperature based on the Fermi-gas model. This leads to closed expressions for N(E) and anti nu/sub p/ when sigma/sub c/ is assumed constant and readily computed quadratures when the energy dependence of sigma/sub c/ is determined from an optical model. Neutron spectra and average multiplicities calculated with an energy-dependent cross section agree well with experimental data for the neutron-induced fission of 235 U and the spontaneous fission of 252 Cf. For the latter case, there are some significant inconsistencies between the experimental spectra that need to be resolved. 29 references

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

Neutron generator for BNCT based on high current ECR ion source with gyrotron plasma heating.

Science.gov (United States)

Skalyga, V; Izotov, I; Golubev, S; Razin, S; Sidorov, A; Maslennikova, A; Volovecky, A; Kalvas, T; Koivisto, H; Tarvainen, O

2015-12-01

BNCT development nowadays is constrained by a progress in neutron sources design. Creation of a cheap and compact intense neutron source would significantly simplify trial treatments avoiding use of expensive and complicated nuclear reactors and accelerators. D-D or D-T neutron generator is one of alternative types of such sources for. A so-called high current quasi-gasdynamic ECR ion source with plasma heating by millimeter wave gyrotron radiation is suggested to be used in a scheme of D-D neutron generator in the present work. Ion source of that type was developed in the Institute of Applied Physics of Russian Academy of Sciences (Nizhny Novgorod, Russia). It can produce deuteron ion beams with current density up to 700-800 mA/cm(2). Generation of the neutron flux with density at the level of 7-8·10(10) s(-1) cm(-2) at the target surface could be obtained in case of TiD2 target bombardment with deuteron beam accelerated to 100 keV. Estimations show that it is enough for formation of epithermal neutron flux with density higher than 10(9) s(-1) cm(-2) suitable for BNCT. Important advantage of described approach is absence of Tritium in the scheme. First experiments performed in pulsed regime with 300 mA, 45 kV deuteron beam directed to D2O target demonstrated 10(9) s(-1) neutron flux. This value corresponds to theoretical estimations and proofs prospects of neutron generator development based on high current quasi-gasdynamic ECR ion source. Copyright © 2015 Elsevier Ltd. All rights reserved.

The advanced neutron source

International Nuclear Information System (INIS)

Raman, S.; Hayter, J.B.

1990-01-01

The Advanced Neutron Source (ANS) is a new user experimental facility planned to be operational at Oak Ridge in the late 1990's. The centerpiece of the ANS will be a steady-state research reactor of unprecedented thermal neutron flux (φ th ∼ 8 x 10 19 m -2 ·s -1 ) accompanied by extensive and comprehensive equipment and facilities for neutron-based research

High-power electron beam tests of a liquid-lithium target and characterization study of (7)Li(p,n) near-threshold neutrons for accelerator-based boron neutron capture therapy.

Science.gov (United States)

Halfon, S; Paul, M; Arenshtam, A; Berkovits, D; Cohen, D; Eliyahu, I; Kijel, D; Mardor, I; Silverman, I

2014-06-01

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center (SNRC). The target is intended to demonstrate liquid-lithium target capabilities to constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals. The lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power >5kW generated by high-intensity proton beams, necessary for sufficient therapeutic neutron flux. In preliminary experiments liquid lithium was flown through the target loop and generated a stable jet on the concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power densities of more than 4kW/cm(2) and volumetric power density around 2MW/cm(3) at a lithium flow of ~4m/s, while maintaining stable temperature and vacuum conditions. These power densities correspond to a narrow (σ=~2mm) 1.91MeV, 3mA proton beam. A high-intensity proton beam irradiation (1.91-2.5MeV, 2mA) is being commissioned at the SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator. In order to determine the conditions of LiLiT proton irradiation for BNCT and to tailor the neutron energy spectrum, a characterization of near threshold (~1.91MeV) (7)Li(p,n) neutrons is in progress based on Monte-Carlo (MCNP and Geant4) simulation and on low-intensity experiments with solid LiF targets. In-phantom dosimetry measurements are performed using special designed dosimeters based on CR-39 track detectors. © 2013 Elsevier Ltd. All rights reserved.

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.

Discussions in symposium 'neutron dosimetry in neutron fields - from detection techniques to medical applications'

International Nuclear Information System (INIS)

Tanimura, Y.; Sato, T.; Kumada, H.; Terunuma, T.; Sakae, T.; Harano, H.; Matsumoto, T.; Suzuki, T.; Matsufuji, N.

2008-01-01

Recently the traceability system (JCSS) of neutron standard based on the Japanese law 'Measurement Act' has been instituted. In addition, importance of the neutron dose evaluation has been increasing in not only the neutron capture medical treatment but also the proton or heavy particle therapy. Against such a background, a symposium 'Neutron dosimetry in neutron fields - From detection techniques to medical applications-' was held on March 29, 2008 and recent topics on the measuring instruments and their calibration, the traceability system, the simulation technique and the medical applications were introduced. This article summarizes the key points in the discussion at the symposium. (author)

Stilbene crystalline powder in polymer base as a new fast neutron detector

International Nuclear Information System (INIS)

Budakovsky, S.V.; Galunov, N.Z.; Grinyov, B.V.; Karavaeva, N.L.; Kyung Kim, Jong; Kim, Yong-Kyun; Pogorelova, N.V.; Tarasenko, O.A.

2007-01-01

A new organic scintillation material consisting of stilbene grains in a polymer glue base is presented. The crystalline grains of stilbene are obtained by mechanical grinding of stilbene single crystals. The resulting composite scintillators have been studied as detectors for fast neutrons

ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

1987-01-01

Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory

Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

Science.gov (United States)

Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

2017-04-01

The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

International Nuclear Information System (INIS)

Simakov, S.P.; Fischer, U.; Moellendorff, U. von; Schmuck, I.; Konobeev, A.Yu.; Korovin, Yu.A.; Pereslavtsev, P.

2002-01-01

A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ 6,7 Li cross section data. A new code M c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M c DeLicious code was checked against available experimental data and calculation results of M c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M c DeLicious along with newly evaluated d+ 6,7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data

Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

CERN Document Server

Simakov, S P; Moellendorff, U V; Schmuck, I; Konobeev, A Y; Korovin, Y A; Pereslavtsev, P

2002-01-01

A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ sup 6 sup , sup 7 Li cross section data. A new code M sup c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M sup c DeLicious code was checked against available experimental data and calculation results of M sup c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M sup c DeLicious along with newly evaluated d+ sup 6 sup , sup 7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data.

CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident

International Nuclear Information System (INIS)

Bessis, J.

1976-01-01

Nature of physical problem solved: Calculation of flux and neutron spectra in the case of a criticality accident. The method is unsophisticated but fast. The program is divided into two parts: (1) The code CRITIC is based on the Fermi age equation and evaluates the neutron number per fission emitted from a moderate critical system and its energy spectrum. (2) The code NARCISSE uses concrete current albedo, evaluates the product of neutron reflection on walls of the source containment and calculates the resulting flux at any point, and its energy distribution into 21 groups. The results obtained seem satisfactory, if compared with a Monte Carlo program

Improved Neutron Scintillators Based on Nanomaterials

International Nuclear Information System (INIS)

Friesel, Dennis

2008-01-01

The development work conducted in this SBIR has so far not supported the premise that using nano-particles in LiFZnS:Ag foils improves their transparency to 420 (or other frequency) light. This conclusion is based solely on the light absorption properties of LiFZnS foils fabricated from nano- and from micro-particles. Furthermore, even for the case of the Gd 2 O 3 foils, the transmission of 420 nm light gained by using nano-particles all but disappears as the foil thickness is increased beyond about 0.2 mm, a practical scintillator thickness. This was not immediately apparent from the preliminary study since no foils thicker than about 0.04 mm were produced. Initially it was believed that the failure to see an improvement by using nano-particles for the LiFZnS foils was caused by the clumping of the particles in Toluene due to the polarity of the ZnS particles. However, we found, much to our surprise, that nano-particle ZnS alone in polystyrene, and in Epoxy, had worse light transmission properties than the micro-particle foils for equivalent thickness and density foils. The neutron detection measurements, while disappointing, are attributable to our inability to procure or fabricate Bulk Doped ZnS nanoparticles. The cause for the failure of nano-particles to improve the scintillation light, and hence improved neutron detection efficiency, is a fundamental one of light scattering within the scintillator. A consequence of PartTec's documentation of this is that several concepts for the fabrication of improved 6 LiFZnS scintillators were formulated that will be the subject of a future SBIR submission.

An advanced fusion neutron source facility

International Nuclear Information System (INIS)

Smith, D.L.

1992-01-01

Accelerator-based 14-MeV-neutron sources based on modifications of the original Fusion Materials Irradiation Facility are currently under consideration for investigating the effects of high-fluence high-energy neutron irradiation on fusion-reactor materials. One such concept for a D-Li neutron source is based on recent advances in accelerator technology associated with the Continuous Wave Deuterium Demonstrator accelerator under construction at Argonne National Laboratory, associated superconducting technology, and advances in liquid-metal technology. In this paper a summary of conceptual design aspects based on improvements in technologies is presented

Lunar neutron source function

International Nuclear Information System (INIS)

Kornblum, J.J.

1974-01-01

The search for a quantitative neutron source function for the lunar surface region is justified because it contributes to our understanding of the history of the lunar surface and of nuclear process occurring on the moon since its formation. A knowledge of the neutron source function and neutron flux distribution is important for the interpretation of many experimental measurements. This dissertation uses the available pertinent experimental measurements together with theoretical calculations to obtain an estimate of the lunar neutron source function below 15 MeV. Based upon reasonable assumptions a lunar neutron source function having adjustable parameters is assumed for neutrons below 15 MeV. The lunar neutron source function is composed of several components resulting from the action of cosmic rays with lunar material. A comparison with previous neutron calculations is made and significant differences are discussed. Application of the results to the problem of lunar soil histories is examined using the statistical model for soil development proposed by Fireman. The conclusion is drawn that the moon is losing mass

Neutronics of pulsed spallation neutron sources

CERN Document Server

Watanabe, N

2003-01-01

Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...

The Advanced Neutron Source

International Nuclear Information System (INIS)

Hayter, J.B.

1989-01-01

The Advanced Neutron Source (ANS) is a new user experimental facility planned to be operational at Oak Ridge in the late 1990's. The centerpiece of the ANS will be a steady-state research reactor of unprecedented thermal neutron flux (φ th ∼ 9·10 19 m -2 ·s -1 ) accompanied by extensive and comprehensive equipment and facilities for neutron-based research. 5 refs., 5 figs

Neutron fluence measurement in nuclear facilities.; Medicion de flujos de neutrones en instalaciones nucleares.

Energy Technology Data Exchange (ETDEWEB)

Camacho L, M E

1997-12-01

The objective of present work is to determine the fluence of neutrons in nuclear facilities using two neutron detectors designed and built at Instituto Nacional de Investigaciones Nucleares (ININ), Mexico. The two neutron detectors are of the passive type, based on solid state nuclear tracks detectors (SSNTD). One of the two neutron detectors was used to determine the fluence distribution of the ports at the nuclear research reactor TRIGA Mark III, which belongs to ININ. In these facilities is important to know the neutron fluence distribution characteristic to carried out diverse kind of research activities. The second neutron detector was employed in order to carry out environmental neutron surveillance. The detector has the property to separate the thermal, intermediate and fast components of the neutron fluence. This detector was used to measure the neutron fluence at hundred points around the primary container of the first Mexican Nuclear Power plant `Laguna Verde`. This last detector was also used to determine the neutron fluence in some points of interest, around and inside a low scattering neutron room at the `Centro de Metrologia de Radiaciones Ionizantes` of the ININ, to know the background neutron field produced by the neutron sources used there. The design of the two neutron detector and the results obtained for each of the surveying facilities, are described in this work. (Author).

Mean energy polarized neutron source

International Nuclear Information System (INIS)

Aleshin, V.A.; Zaika, N.I.; Kolotyj, V.V.; Prokopenko, V.S.; Semenov, V.S.

1988-01-01

Physical bases and realization scheme of a pulsed source of polarized neutrons with the energy of up to 75 MeV are described. The source comprises polarized deuteron source, transport line, low-energy ion and axial injector to the accelerator, U-240 isochronous cyclotron, targets for polarized neutron production, accelerated deuteron transport line and flight bases. The pulsed source of fast neutrons with the energy of up to 75 MeV can provide for highly polarized neutron beams with the intensity by 2-3 orders higher than in the most perfect source of this range which allows one to perform various experiments with high efficiency and energy resolution. 9 refs.; 1 fig

Moving converter as the possible tool for producing ultra-cold neutrons on pulsed neutron sources

International Nuclear Information System (INIS)

Pokotilovskij, Yu.N.

1991-01-01

A method is proposed for producing ultra-cold neutrons (UCN) at aperiodic pulse neutron sources. It is based on the use of the fast moving cooled converter of UCN in the time of the neutron pulse and includes the trapping of generated UCN's in a moving trap. 6 refs.; 2 figs

Preliminary neutronic assessment for ATF (Accident Tolerant Fuel) based on iron alloy

International Nuclear Information System (INIS)

Abe, Alfredo; Carluccio, Thiago; Piovezan, Pamela; Giovedi, Claudia; Martins, Marcelo R.

2015-01-01

After Fukushima Daiichi nuclear accident in 2011, the nuclear fuel performance under accident condition became a very important issue and currently different research and development program are in progress toward to reliability and withstand under accident condition. These initiatives are known as ATF (Accident Tolerant Fuel) R and D program, which many countries with different research institutes, fuel vendors and others are nowadays involved. Accident Tolerant Fuel (ATF) can be defined as enhanced fuel which can tolerate loss of active cooling system capability for a considerably longer time period and the fuel/cladding system can be maintained without significant degradation and can also improve the fuel performance during normal operations and transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) accident. Different materials have being proposed as fuel cladding candidates considering thermo-mechanical properties and lower reaction kinetic with steam and slower hydrogen production. The aim of this work is to perform a neutronic assessment for several cladding candidates based on iron alloy considering a standard PWR fuel rod (fuel pellet and dimension). The purpose of the assessment is to address different parameters that might contribute for possible neutronic reactivity gain in order to overcome the penalty due to increase of neutron absorption in the cladding materials. All the neutronic assessment is performed using MCNP, Monte Carlo code. (author)

A study on the characteristics of modified and novolac type epoxy resin based neutron shielding material

Energy Technology Data Exchange (ETDEWEB)

Cho, Soo Haeng; Hong, Sun Seok; Oh, Seung Chul; Do, Jae Bum [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-10-01

Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. In this study, we developed modified and novolac type epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, prolonged time heat resistance, thermal and mechanical properties were evaluated experimently. (author). 31 refs., 27 figs., 16 tabs.

A compact DD neutron generator-based NAA system to quantify manganese (Mn) in bone in vivo.

Science.gov (United States)

Liu, Yingzi; Byrne, Patrick; Wang, Haoyu; Koltick, David; Zheng, Wei; Nie, Linda H

2014-09-01

A deuterium-deuterium (DD) neutron generator-based neutron activation analysis (NAA) system has been developed to quantify metals, including manganese (Mn), in bone in vivo. A DD neutron generator with a flux of up to 3*10(9) neutrons s(-1) was set up in our lab for this purpose. Optimized settings, including moderator, reflector, and shielding material and thickness, were selected based on Monte Carlo (MC) simulations conducted in our previous work. Hand phantoms doped with different Mn concentrations were irradiated using the optimized DD neutron generator irradiation system. The Mn characteristic γ-rays were collected by an HPGe detector system with 100% relative efficiency. The calibration line of the Mn/calcium (Ca) count ratio versus bone Mn concentration was obtained (R(2) = 0.99) using the hand phantoms. The detection limit (DL) was calculated to be about 1.05 μg g(-1) dry bone (ppm) with an equivalent dose of 85.4 mSv to the hand. The DL can be reduced to 0.74 ppm by using two 100% HPGe detectors. The whole body effective dose delivered to the irradiated subject was calculated to be about 17 μSv. Given the average normal bone Mn concentration of 1 ppm in the general population, this system is promising for in vivo bone Mn quantification in humans.

A compact DD neutron generator–based NAA system to quantify manganese (Mn) in bone in vivo

Science.gov (United States)

Liu, Yingzi; Byrne, Patrick; Wang, Haoyu; Koltick, David; Zheng, Wei; Nie, Linda H.

2015-01-01

A deuterium-deuterium (DD) neutron generator–based neutron activation analysis (NAA) system has been developed to quantify metals, including manganese (Mn), in bone in vivo. A DD neutron generator with a flux of up to 3*109 neutrons/second was set up in our lab for this purpose. Optimized settings, including moderator, reflector, and shielding material and thickness, were selected based on Monte Carlo (MC) simulations conducted in our previous work. Hand phantoms doped with different Mn concentrations were irradiated using the optimized DD neutron generator irradiation system. The Mn characteristic γ-rays were collected by an HPGe detector system with 100% relative efficiency. The calibration line of the Mn/calcium (Ca) count ratio versus bone Mn concentration was obtained (R2 = 0.99) using the hand phantoms. The detection limit (DL) was calculated to be about 1.05 μg/g dry bone (ppm) with an equivalent dose of 85.4 mSv to the hand. The DL can be reduced to 0.74 ppm by using two 100% HPGe detectors. The whole body effective dose delivered to the irradiated subject was calculated to be about 17 μSv. Given the average normal bone Mn concentration of 1 ppm in the general population, this system is promising for in vivo bone Mn quantification in humans. PMID:25154883

Cf-252 based neutron radiography using real-time image processing system

International Nuclear Information System (INIS)

Mochiki, Koh-ichi; Koiso, Manabu; Yamaji, Akihiro; Iwata, Hideki; Kihara, Yoshitaka; Sano, Shigeru; Murata, Yutaka

2001-01-01

For compact Cf-252 based neutron radiography, a real-time image processing system by particle counting technique has been developed. The electronic imaging system consists of a supersensitive imaging camera, a real-time corrector, a real-time binary converter, a real-time calculator for centroid, a display monitor and a computer. Three types of accumulated NR image; ordinary, binary and centroid images, can be observed during a measurement. Accumulated NR images were taken by the centroid mode, the binary mode and ordinary mode using of Cf-252 neutron source and those images were compared. The centroid mode presented the sharpest image and its statistical characteristics followed the Poisson distribution, while the ordinary mode showed the smoothest image as the averaging effect by particle bright spots with distributed brightness was most dominant. (author)

[International Panel on 14 MeV Intense Neutron Source Based on Accelerators for Fusion Materials Study

International Nuclear Information System (INIS)

Thoms, K.R.; Wiffen, F.W.

1991-01-01

Both travelers were members of a nine-person US delegation that participated in an international workshop on accelerator-based 14 MeV neutron sources for fusion materials research hosted by the University of Tokyo. Presentations made at the workshop reviewed the technology developed by the FMIT Project, advances in accelerator technology, and proposed concepts for neutron sources. One traveler then participated in the initial meeting of the IEA Working Group on High Energy, High Flux Neutron Sources in which efforts were begun to evaluate and compare proposed neutron sources; the Fourth FFTF/MOTA Experimenters' Workshop which covered planning and coordination of the US-Japan collaboration using the FFTF reactor to irradiate fusion reactor materials; and held discussions with several JAERI personnel on the US-Japan collaboration on fusion reactor materials

Monte Carlo calculations and neutron spectrometry in quantitative prompt gamma neutron activation analysis (PGNAA) of bulk samples using an isotopic neutron source

International Nuclear Information System (INIS)

Spyrou, N.M.; Awotwi-Pratt, J.B.; Williams, A.M.

2004-01-01

An activation analysis facility based on an isotopic neutron source (185 GBq 241 Am/Be) which can perform both prompt and cyclic activation analysis on bulk samples, has been used for more than 20 years in many applications including 'in vivo' activation analysis and the determination of the composition of bio-environmental samples, such as, landfill waste and coal. Although the comparator method is often employed, because of the variety in shape, size and elemental composition of these bulk samples, it is often difficult and time consuming to construct appropriate comparator samples for reference. One of the obvious problems is the distribution and energy of the neutron flux in these bulk and comparator samples. In recent years, it was attempted to adopt the absolute method based on a monostandard and to make calculations using a Monte Carlo code (MCNP4C2) to explore this further. In particular, a model of the irradiation facility has been made using the MCNP4C2 code in order to investigate the factors contributing to the quantitative determination of the elemental concentrations through prompt gamma neutron activation analysis (PGNAA) and most importantly, to estimate how the neutron energy spectrum and neutron dose vary with penetration depth into the sample. This simulation is compared against the scattered and transmitted neutron energy spectra that are experimentally and empirically determined using a portable neutron spectrometry system. (author)

Total variation-based neutron computed tomography

Science.gov (United States)

Barnard, Richard C.; Bilheux, Hassina; Toops, Todd; Nafziger, Eric; Finney, Charles; Splitter, Derek; Archibald, Rick

2018-05-01

We perform the neutron computed tomography reconstruction problem via an inverse problem formulation with a total variation penalty. In the case of highly under-resolved angular measurements, the total variation penalty suppresses high-frequency artifacts which appear in filtered back projections. In order to efficiently compute solutions for this problem, we implement a variation of the split Bregman algorithm; due to the error-forgetting nature of the algorithm, the computational cost of updating can be significantly reduced via very inexact approximate linear solvers. We present the effectiveness of the algorithm in the significantly low-angular sampling case using synthetic test problems as well as data obtained from a high flux neutron source. The algorithm removes artifacts and can even roughly capture small features when an extremely low number of angles are used.

Neutronics of pulsed spallation neutron sources

International Nuclear Information System (INIS)

Watanabe, Noboru

2003-01-01

Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, poisoning, etc are discussed, aiming at a high performance pulsed spallation source

Prospects for a new cold neutron beam measurement of the neutron lifetime

Energy Technology Data Exchange (ETDEWEB)

Dewey, M., E-mail: mdewey@nist.go [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); Coakley, K., E-mail: kevin.coakley@nist.go [National Institute of Standards and Technology, Boulder, CO 80305 (United States); Gilliam, D., E-mail: david.gilliam@nist.go [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); Greene, G., E-mail: greenegl@ornl.go [Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Physics Division, Oak Ridge National Lab, Building 6010, Oak Ridge, TN 37831 (United States); Laptev, A., E-mail: alaptev@nist.go [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Nico, J., E-mail: jnico@nist.go [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); Snow, W., E-mail: wsnow@indiana.ed [Indiana University/IUCF, Bloomington, IN 47408 (United States); Wietfeldt, F., E-mail: few@tulane.ed [Tulane University, New Orleans, LA 70118 (United States); Yue, A., E-mail: ayue@nist.go [Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States)

2009-12-11

In the most accurate cold neutron beam determination of the neutron lifetime based on the absolute counting of decay protons, the largest uncertainty was attributed to the absolute determination of the capture flux of the cold neutron beam. Currently an experimental effort is underway at the National Institute of Standards and Technology (NIST) that will significantly reduce this contribution to the uncertainty in the lifetime determination. The next largest source of uncertainty is the determination of the absolute count rate of decay protons, which contributes to the experimental uncertainty approximately at the 1 s level. Experience with the recent neutron radiative decay experiment, which used the neutron lifetime apparatus, has provided valuable insights into ways to reduce other uncertainties. In addition, the cold neutron fluence rate at NIST is presently 1.5 times greater than in the 2003 measurement, and there is the prospect for a significantly higher rate with the new guide hall expansion. This paper discusses an approach for achieving a determination of the neutron lifetime with an accuracy of approximately 1 s.

Assessment of NJOY generated neutron heating factors based on JEF/EFF-1

International Nuclear Information System (INIS)

Vontobel, P.

1990-01-01

Using the NJOY nuclear data processing system, a coupled neutron-photon multigroup MATXS-formatted nuclear data library was generated based on the files JEF/EFF-1. The neutron heating factors contained in this VITAMIN-J structured library are compared with those of MACLIB-IV. The main differences are due to the included decay heat of shortlived reaction products in MACKLIB-IV and/or due to too high/low photon production data of some JEF/EFF-1 isotopes. It is recommended to check carefully the energy balance of new evaluations containing photon production data. How this can be done with the help of the NJOY HEATR module is shown in an example. (author) 35 figs., 9 refs