Nuclear data measurements in 3x592 GBq 241Am-Be neutron cell

International Nuclear Information System (INIS)

2010-01-01

The aim of this study is to present the results of the activities carried out within the scope of the Nuclear Data Measurements in 3x592 GBq Am-Be Neutron Cell project. The study covers the establishment of neutron irradiation systems, neutron and gamma dose rate evaluations in and around the laboratory, performance measurements of neutron irradiation systems, measurements of thermal, epithermal and fast neutron flux, gamma spectrometer efficiency calibrations, fast neutron fission product yield measurements for fertile nuclides ( 2 32Th and 2 38U), cross section measurements for fast neutron threshold detectors, gamma ray intensity measurements of the nuclides in uranium decay chain, elemental detection limit measurements and the half life measurement of short-lived isotopes. First of all, an irradiation geometry, which enables optimum irradiation, was designed for an irradiation system of 3 2 41Am-Be sources with 592 GBq activity each. Paraffin was chosen in order to slow down the source neutrons. An equilateral quadrangle with 70 cm side length and 60 cm height was used as paraffin moderator. Experimentally, it was determined that paraffin with approximately 3.5 cm thickness slows down to maximum thermal neutron flux of 2 41Am-Be neutrons. Paraffin block was placed on the base of the source room. In order to determine the positions of thermal and fast neutron irradiations, indium wires were irradiated with 5 mm intervals vertically parallel to the neutron sources in thermal and fast neutron irradiation cells. The position of maximum thermal and fast neutron fluxes is 61.5 cm for the thermal neutron irradiation cells and 69 cm for the fast neutron irradiation cell, from the top of the irradiation pipes down. One of the most important parameters of nuclear data measurements is the counting efficiency of the gamma spectrometer used for each counting geometry. For this reason, the detector efficiencies for the related counting geometries need to be measured

Characterization of the neutron field of the 241AmBe in a calibration room

International Nuclear Information System (INIS)

Vega C, H.R.; Gallego, E.; Lorente, A.

2003-01-01

The field of neutrons produced by an isotopic source of neutrons of 241 Am Be had been characterized. The characterization was carried out modeling those relevant details of the calibration room and simulating the neutron transport at different distances of the source. The calculated spectra were used to determine the equivalent environmental dose rate. A series of experiments were carried out with the Bonner sphere spectrometric system to measure the spectra in the same points where the calculations were carried out and with these spectra the rates of environmental dose were calculated. By means of a one sphere dosemeter type Berthold the rates of environmental dose were measured. To the one to compare the calculated spectra and measured its were found small differences in the group of the thermal neutrons due to the elementary composition used during the simulation. When comparing the derived rates starting from the calculated spectra with those measured it was found a maxim difference smaller to 13%. (Author)

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.

Neutron capture cross section of ^243Am

Science.gov (United States)

Jandel, M.

2009-10-01

The Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos National Laboratory (LANL) was used for neutron capture cross section measurement on ^243Am. The high granularity of DANCE (160 BaF2 detectors in a 4π geometry) enables the efficient detection of prompt gamma-rays following neutron capture. DANCE is located on the 20.26 m neutron flight path 14 (FP14) at the Manuel Lujan Jr. Neutron Scattering Center at the Los Alamos Neutron Science Center (LANSCE). The methods and techniques established in [1] were used for the determination of the ^243Am neutron capture cross section. The cross sections were obtained in the range of neutron energies from 0.02 eV to 400 keV. The resonance region was analyzed using SAMMY7 and resonance parameters were extracted. The results will be compared to existing evaluations and calculations. Work was performed under the auspices of the U.S. Department of Energy at Los Alamos National Laboratory by the Los Alamos National Security, LLC under Contract No. DE-AC52-06NA25396 and at Lawrence Livermore National Laboratory by the Lawrence Livermore National Security, LLC under Contract No. DE-AC52-07NA27344. [4pt] [1] M. Jandel et al., Phys. Rev. C78, 034609 (2008)

Neutron energy spectra of sup 2 sup 5 sup 2 Cf, Am-Be source and of the D(d,n) sup 3 He reaction

CERN Document Server

Sang Tae Park

2003-01-01

The neutron energy spectrum of the following sources were measured using a fast neutron spectrometer with the NE-213 liquid scintillator: sup 2 sup 5 sup 2 Cf, Am-Be and D(d,n) sup 3 He reaction from a 3 MeV Pelletron accelerator in Tokyo Institute of Technology. The measured proton recoil pulse height data of sup 2 sup 5 sup 2 Cf, Am-Be and D(d,n) sup 3 He were unfolded using the mathematical program to obtain the neutron energy spectrum. The sup 2 sup 5 sup 2 Cf and Am-Be neutron energy spectra were measured and the results obtained showed a good agreement with the spectra usually published in the literature. The neutron energy spectrum from D(d,n) sup 3 He was measured and the results obtained also showed a good agreement with the calculation by time of flight (TOF) methods. (author)

Measurements of neutron cross section of the {sup 243}Am(n,{gamma}){sup 244}Am reaction

Energy Technology Data Exchange (ETDEWEB)

Hatsukawa, Yuichi; Shinohara, Nobuo; Hata, Kentaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

The effective thermal neutron cross section of {sup 243}Am(n,{gamma}){sup 244}Am reaction was measured by the activation method. Highly-purified {sup 243}Am target was irradiated in an aluminum capsule by using a research reactor JRR-3M. The tentative effective thermal neutron cross sections are 3.92 b, and 84.44 b for the production of {sup 244g}Am and {sup 244m}Am, respectively. (author)

Characteristics of the quarry as shielding for {sup 241}AmBe neutrons and monoenergetic photons; Caracteristicas de la cantera como blindaje para los neutrones del {sup 241}AmBe y fotones monoenergeticos

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C.; Salas L, M. A. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Rodriguez R, J. A.; Juarez A, C. A., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Nuevo Leon, Facultad de Ingenieria Civil, Pedro de Alba s/n, San Nicolas de los Garza, Nuevo Leon (Mexico)

2016-09-15

Shielding is an important element in radiation protection since allows the management of radiation sources. Currently there are different materials of natural or anthropogenic origin that are used as shielding for both photons and neutrons. The quarry is a material of natural origin and abundant in our country, which is used in construction or for the manufacture of sculptures, however its characteristics as shielding have not been reported. In this paper we report some of the properties of the quarry as shielding for monoenergetic photons and for neutrons produced by an isotopic neutron source of {sup 241}AmBe. A quarry piece was used to determine its density and its chemical composition, with the XCOM code the elemental composition was determined and the mass interaction and total attenuation coefficients of the quarry were determined with photons of 10{sup -3} to 10{sup -5} MeV; the interaction coefficients included coherent dispersion, photoelectric absorption, Compton dispersion and the production of pairs in the nuclear and electronic field. Using the MCNP5 code, a narrow geometry attenuation experiment was modeled and the photon fluence was estimated that reaches a point detector at a distance of 42 cm from a point source, isotropic and monoenergetic photon when the source and the point detector were added quarry pieces of different thicknesses. The reduction of the number of photons as a function of the thickness of the quarry was used to determine the coefficient of linear attenuation of the quarry before photons of 0.03, 0.07, 0.1, 0.3, 1, 2 and 3 MeV that were the same as those calculated with the XCOM code. With the MCNP, the K a and H(10) transmission curves were also calculated. This same model was used to determined the variation of the {sup 241}AmBe neutron spectrum as a function of quarry thickness, as well as the E{sub ROT} and H(10) transmission curves. (Author)

Neutron beam facilities at the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Kim, S.

2003-01-01

The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

Determination of the hydrogen content of oil samples from Nigeria using an Am-Be neutron source

International Nuclear Information System (INIS)

Jonah, S.A.; Elegba, S.B.; Zakari, I.I.

1998-01-01

A 5 Ci Am-Be neutron source-based facility, which utilises the principles of thermal neutron reflection technique in combination with foil activation method, has been used to determine the total hydrogen content of commercial oil samples from Nigeria. With an established detection limit of 0.25 H w% for oil matrix of volume 600-ml, the total hydrogen contents of the samples were found to be in the range of 11.11-14.22 H w%. The facility is economical and suitable for the determination of moisture in solid samples. A brief description of the ongoing projects and future plans concerning the CRP are enumerated. (author)

Neutron beam facilities at the Australian Replacement Research Reactor

International Nuclear Information System (INIS)

Kennedy, Shane; Robinson, Robert; Hunter, Brett

2001-01-01

Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10 14 n/cm 2 /sec and a liquid D 2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

50 curie Am-Be neutron source in determining impurities in various materials

International Nuclear Information System (INIS)

Rastikerdar, S.

1998-01-01

The neutrons from a 50 Curies Am-Be neutron source after being thermalized have been used to study the impurities in various materials by measuring the gamma rays emitted from the activated samples. To get good resolution two HPGe detectors, one of them suitable for low energy gamma rays as well as X-rays and the other suitable for measuring the gamma-ray energies up to 10 MeV have been used. The resolution of the detectors were measured and proved to be better than 1.8 keV for 60 Co gamma rays. During the measurements the detectors were placed in thick lead chambers. In these chambers the background was reduced dramatically. To make the whole system safe and also for saving time in activation analysis a fully computerized control rabbit device has been coupled to the system. Our main purpose is to set up a portable, cheap and reliable system for activation analysis for research institutions that are not able to have reactors due to various reasons. Although our tests and analysis is still in progress we think that the system is very promising. In this paper we will discuss about the details and the future prospects. (author)

A cytogenetic study on persons exposed to 241Am-Be neutron source

International Nuclear Information System (INIS)

Bai Yushu; Zhang Xiuxia; Guan Shurong; Xie Feng; Lu Meiying

1988-01-01

An 241 Am-Be neutron source of 120 millicrocurie was stolen by a boy on April the 3rd, 1982. He brought it to his home. The neutron source was broken and the radioactive substance contaminated the whole family. He and his other family members did not leave the contaminated environment for 70 days. There were altogether 14 persons exposed to the radioactive substances. Chromosome aberrations of peripheral blood lymphocytes of six exposed cases were analysed on an early day of June 1982. The exposure doses were estimated by the frequency of chromosome aberration. Biological dose absorbed by the boy who stole the source was about 96 ∼ 128 mGy (physical dose about 120 mGy), the others were about 10 ∼ 30 mGy (physical dose about 10 ∼ 30 mGy). The results indicated that biological dose were quite approximate to the physical doses. One and half years after the accident, the analysis of stable chromosome aberrations for 14 exposed persons showed that the frequencies of chromosome aberrations of the irradiated individuals were still higher than those of the controls. The differences between them are very significant

Application of the neutron irradiator with AmBe sources for inorganic elements in commercial fertilizers determination; Aplicacao do irradiador de neutrons com fontes de AmBe para determinacao de elementos inorganicos em fertilizantes comerciais

Energy Technology Data Exchange (ETDEWEB)

Madi Filho, Tufic; Armelin, Maria Jose Aguirre; Fulas, Paulo Marcelo Marangon [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: tmfilho@ipen.br; Figueira, Rubens Cesar Lopes [Universidade Cruzeiro do Sul (UNICSUL), SP (Brazil); Trevizam, Anderson Ricardo [Centro de Energia Nuclear na Agricultura (CENA), Piracicaba, SP (Brazil)

2005-07-01

The rational use of fertilizers, for the soil fertility correction, contributes to the increase of agricultural production, using the same areas previously available. The quality of products could be improved with reduced costs. Therefore, knowledge of the chemical characteristics of the correctives used is required to streamline the application and avoid excesses or deficiencies. The studied characteristics are generally limited to the essential nutrients for the nutrition of plants and animals, e.g.: Mn, Zn, P, K, Cu and those known toxic, such as: As, Cd, Hg and Pb. Neutron activation analysis (NAA) is a highly sensitive non destructive technique, for the determination of the elemental composition in samples. It has been particularly useful in the simultaneous determination of inorganic elements in complex samples of several kinds. Several analysis methods for activation are used, such as: comparative and absolute. Commercial fertilizers were analyzed applying the absolute and comparative methods. Using the absolute method, samples were submitted to neutron flux generated by an Irradiator with two AmBe sources. The obtained results were compared with those obtained by the comparative method using neutrons generated in the IEA-R1 Reactor. (author)

Application of AmBe source neutron irradiator for determination of inorganic elements in commercial fertilizers; Aplicacao do irradiador de neutrons com fonte de AmBe para determinacao de elementos inorganicos em fertilizantes comerciais

Energy Technology Data Exchange (ETDEWEB)

Madi Filho, Tufic; Armelim, Maria Jose Aguirre; Fulas, Paulo Marcelo Marangon [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: tmfilho@ipen.br; Trevizam, Anderson Ricardo [Universidade Cruzeiro do Sul (UNICSUL), Sao Paulo, SP (Brazil); Figueira, Rubens Cesar Lopes [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil)

2005-11-15

The rational use of fertilizers , for the soil fertility correction, contributes to the increase of agricultural production, using the same areas previously available. The quality of products could ne improved with reduced costs. Therefore, knowledge of the chemical characteristics of the correctives used is required to streamline the application and avoid excesses or deficiencies. The studied characteristics are generally limited to the essential nutrients for the nutrition of plants and animals, e.g.: Mn, Zn, P, K, Cu and those known toxic, such as: As, Cd, Hg and Pb. Neutron activation analysis (NAA) is a highly sensitive non destructive technique, for the determination of the elemental composition in samples. It has been particularly useful in the simultaneous determination of inorganic elements in complex samples of several kinds. Several analysis methods for activation are used, such as: comparative and absolute. Commercial fertilizers were analyzed applying the absolute and comparative methods. Using the absolute method, samples were submitted to neutron flux generated by an irradiator with two Am Be sources. The obtained results were compared with those obtained by the comparative method using neutrons generated in the IEA-R1 Reactor. (author)

Response of LR-115 type II and CR-39 plastic track detectors to Am-Be and 14.1-MeV neutrons

International Nuclear Information System (INIS)

Bradley, D.A.; Chong, C.S.; Saat, Ahmat; Sidik, A.G.; Ghose, A.M.

1987-01-01

The fast-neutron response of the plastic LR-115 type II and CR-39 track detectors have been compared, using a 14.1-MeV neutron generator and a radionuclide Am-Be neutron source (effective primary neutron energy 4.5-MeV). The distribution of track diameters for a range of etching times has been evaluated, taking into account track registration efficiency and the relevant fast neutron scattering cross-sections. The efficiency of etched-track formation in LR-115 type II due to neutron irradiation is approximately double that in CR-39. The 14.1-MeV neutrons also tend to produce tracks in both materials with somewhat greater efficiency than do the lower energy neutrons from the radionuclide source, for a given etching time. (author)

The neutron beam facility at the Australian replacement research reactor

International Nuclear Information System (INIS)

Hunter, B.; Kennedy, S.

1999-01-01

Full text: The Australian federal government gave ANSTO final approval to build a research reactor to replace HIFAR on August 25th 1999. The replacement reactor is to be a multipurpose reactor with a thermal neutron flux of 3 x 10 14 n.cm -2 .s -1 and having improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The replacement reactor will commence operation in 2005 and will cater for Australian scientific, industrial and medical needs well into the 21st century. The scientific capabilities of the neutron beams at the replacement reactor are being developed in consultation with representatives from academia, industry and government research laboratories to provide a facility for condensed matter research in physics, chemistry, materials science, life sciences, engineering and earth sciences. Cold, thermal and hot neutron sources are to be installed, and neutron guides will be used to position most of the neutron beam instruments in a neutron guide hall outside the reactor confinement building. Eight instruments are planned for 2005, with a further three to be developed by 2010. A conceptual layout for the neutron beam facility is presented including the location of the planned suite of neutron beam instruments. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by an accredited reactor builder in a turnkey contract. Tenders have been called for December 1999, with selection of contractor planned by June 2000. The neutron beam instruments will be developed by ANSTO and other contracted organisations in consultation with the user community and interested overseas scientists. The facility will be based, as far as possible, around a neutron guide hall that is be served by three thermal and three cold neutron guides. Efficient transportation of thermal and cold neutrons to the guide hall requires the use of modern super

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.

Characteristics of the quarry as shielding for "2"4"1AmBe neutrons and monoenergetic photons

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C.; Salas L, M. A.; Rodriguez R, J. A.; Juarez A, C. A.

2016-09-01

Shielding is an important element in radiation protection since allows the management of radiation sources. Currently there are different materials of natural or anthropogenic origin that are used as shielding for both photons and neutrons. The quarry is a material of natural origin and abundant in our country, which is used in construction or for the manufacture of sculptures, however its characteristics as shielding have not been reported. In this paper we report some of the properties of the quarry as shielding for monoenergetic photons and for neutrons produced by an isotopic neutron source of "2"4"1AmBe. A quarry piece was used to determine its density and its chemical composition, with the XCOM code the elemental composition was determined and the mass interaction and total attenuation coefficients of the quarry were determined with photons of 10"-"3 to 10"-"5 MeV; the interaction coefficients included coherent dispersion, photoelectric absorption, Compton dispersion and the production of pairs in the nuclear and electronic field. Using the MCNP5 code, a narrow geometry attenuation experiment was modeled and the photon fluence was estimated that reaches a point detector at a distance of 42 cm from a point source, isotropic and monoenergetic photon when the source and the point detector were added quarry pieces of different thicknesses. The reduction of the number of photons as a function of the thickness of the quarry was used to determine the coefficient of linear attenuation of the quarry before photons of 0.03, 0.07, 0.1, 0.3, 1, 2 and 3 MeV that were the same as those calculated with the XCOM code. With the MCNP, the K a and H(10) transmission curves were also calculated. This same model was used to determined the variation of the "2"4"1AmBe neutron spectrum as a function of quarry thickness, as well as the E_R_O_T and H(10) transmission curves. (Author)

Application of AmBe source neutron irradiator for determination of inorganic elements in commercial fertilizers

International Nuclear Information System (INIS)

Madi Filho, Tufic; Armelim, Maria Jose Aguirre; Fulas, Paulo Marcelo Marangon; Trevizam, Anderson Ricardo; Figueira, Rubens Cesar Lopes

2005-01-01

The rational use of fertilizers , for the soil fertility correction, contributes to the increase of agricultural production, using the same areas previously available. The quality of products could ne improved with reduced costs. Therefore, knowledge of the chemical characteristics of the correctives used is required to streamline the application and avoid excesses or deficiencies. The studied characteristics are generally limited to the essential nutrients for the nutrition of plants and animals, e.g.: Mn, Zn, P, K, Cu and those known toxic, such as: As, Cd, Hg and Pb. Neutron activation analysis (NAA) is a highly sensitive non destructive technique, for the determination of the elemental composition in samples. It has been particularly useful in the simultaneous determination of inorganic elements in complex samples of several kinds. Several analysis methods for activation are used, such as: comparative and absolute. Commercial fertilizers were analyzed applying the absolute and comparative methods. Using the absolute method, samples were submitted to neutron flux generated by an irradiator with two Am Be sources. The obtained results were compared with those obtained by the comparative method using neutrons generated in the IEA-R1 Reactor. (author)

Methodology for Quantitative Analysis of Large Liquid Samples with Prompt Gamma Neutron Activation Analysis using Am-Be Source

International Nuclear Information System (INIS)

Idiri, Z.; Mazrou, H.; Beddek, S.; Amokrane, A.

2009-01-01

An optimized set-up for prompt gamma neutron activation analysis (PGNAA) with Am-Be source is described and used for large liquid samples analysis. A methodology for quantitative analysis is proposed: it consists on normalizing the prompt gamma count rates with thermal neutron flux measurements carried out with He-3 detector and gamma attenuation factors calculated using MCNP-5. The relative and absolute methods are considered. This methodology is then applied to the determination of cadmium in industrial phosphoric acid. The same sample is then analyzed by inductively coupled plasma (ICP) method. Our results are in good agreement with those obtained with ICP method.

Monte Carlo Simulation on Compensated Neutron Porosity Logging in LWD With D-T Pulsed Neutron Generator

International Nuclear Information System (INIS)

Zhang Feng; Hou Shuang; Jin Xiuyun

2010-01-01

The process of neutron interaction induced by D-T pulsed neutron generator and 241 Am-Be source was simulated by using Monte Carlo method. It is concluded that the thermal neutron count descend exponentially as the spacing increasing. The smaller porosity was, the smaller the differences between the two sources were. When the porosity reached 40%, the ratio of thermal neutron count generated by D-T pulsed neutron source was much larger than that generated by 241 Am-Be neutron source, and its distribution range was wider. The near spacing selected was 20-30 cm, and that of far spacing was about 60-70 cm. The detection depth by using D-T pulsed neutron source was almost unchanged under condition of the same sapcing, and the sensitivity of measurement to the formation porosity decreases. The results showed that it can not only guarantee the statistic of count, but also improve detection sensitivity and depth at the same time of increasing spacing. Therefore, 241 Am-Be neutron source can be replaced by D-T neutron tube in LWD tool. (authors)

Application of the neutron irradiator with AmBe sources for inorganic elements in commercial fertilizers determination

International Nuclear Information System (INIS)

Madi Filho, Tufic; Armelin, Maria Jose Aguirre; Fulas, Paulo Marcelo Marangon; Figueira, Rubens Cesar Lopes; Trevizam, Anderson Ricardo

2005-01-01

The rational use of fertilizers, for the soil fertility correction, contributes to the increase of agricultural production, using the same areas previously available. The quality of products could be improved with reduced costs. Therefore, knowledge of the chemical characteristics of the correctives used is required to streamline the application and avoid excesses or deficiencies. The studied characteristics are generally limited to the essential nutrients for the nutrition of plants and animals, e.g.: Mn, Zn, P, K, Cu and those known toxic, such as: As, Cd, Hg and Pb. Neutron activation analysis (NAA) is a highly sensitive non destructive technique, for the determination of the elemental composition in samples. It has been particularly useful in the simultaneous determination of inorganic elements in complex samples of several kinds. Several analysis methods for activation are used, such as: comparative and absolute. Commercial fertilizers were analyzed applying the absolute and comparative methods. Using the absolute method, samples were submitted to neutron flux generated by an Irradiator with two AmBe sources. The obtained results were compared with those obtained by the comparative method using neutrons generated in the IEA-R1 Reactor. (author)

International key comparison of measurements of neutron source emission rate (1999-2005): CCRI(III)-K9.AmBe

Energy Technology Data Exchange (ETDEWEB)

Roberts, N.J.; Jones, L.N. [National Physical Laboratory (NPL), Teddington, (United Kingdom); Wang, Z.; Liu, Y.; Wang, Q.; Chen, X.; Luo, H.; Rong, C. [China Institute of Atomic Energy (CIAE), Beijing (China); Kralik, M. [Czech Metrology Institute (CMI), Praha, (Czech Republic); Park, H.; Choi, K.O. [Korea Research Institute of Standards and Science (KRISS), Daejeon, (Korea, Republic of); Pereira, W.W.; Da Fonseca, E.S. [National Laboratory of Metrology of Ionizing Radiation (LNMRI), Rio de Janeiro, (Brazil); Cassette, P. [Laboratoire National Henri Becquerel (LNE-LNHB), Paris, (France); Dewey, M.S. [National Institute of Standards and Technology (NIST), Gaithersburg, MD, (United States); Moiseev, N.N.; Kharitonov, I.A. [D I Mendeleyev Institute for Metrology (VNIIM), St Petersburg, (Russian Federation)

2011-07-01

Section III (neutron measurements) of the Comite Consultatif des Rayonnements Ionisants, CCRI, conducted a key comparison of primary measurements of the neutron emission rate of an {sup 241}Am-Be({alpha},n) radionuclide source. A single {sup 241}Am-Be({alpha},n) source was circulated to all the participants between 1999 and 2005. Eight laboratories participated - the CIAE (China), CMI (Czech Republic), KRISS (Republic of Korea), LNMRI (Brazil), LNE-LNHB (France), NIST (USA), NPL (UK) and the VNIIM (Russian Federation) - with the NPL making their measurements at the start and repeating them near the end of the exercise to verify the stability of the source. Each laboratory reported the emission rate into 4{pi} sr together with a detailed uncertainty budget. All participants used the manganese bath technique, with the VNIIM also making measurements using an associated particle technique. The CMI, KRISS, VNIIM, and later the NPL, also measured the anisotropy of the source although this was not a formal part of the comparison. The first draft report was released in May 2006 and having been discussed and modified by the participants and subsequently reviewed by the CCRI(III), the present paper is now the final report of the comparison. (authors)

Optimization of shadow cone length and mass for determination the amount of scattered radiation dose in the calibration laboratory of Am/Be neutron source

International Nuclear Information System (INIS)

Raisali, G.; Hamidi, S.; Hallajfard, E.; Shahvar, A.; Hajiloo, N.

2007-01-01

The shadow cone technique is one of the methods which is used for determining the contribution of scattered particles on the response of neutron detectors. This technique is used for neutron field calibration in Agriculture, Medicine and Industry Research School. In this investigation, we have designed and constructed an optimized shadow cone. According to the calculated neutron dose equivalent attenuation factors, a cone with 20 cm of iron and 30 cm of polyethylene has been found as optimum. For this cone, the neutron dose equivalent attenuation factor for 241 Am/Be neutron source, is 0.00035 for which the contribution of scattered neutrons in Agriculture, Medicine and Industry Research School neutron calibration laboratory according to the calculation and measurement results, can be evaluated with less than 0.5% of error

Characterization of neutron flux spectra in the irradiation sites of a 37 GBq {sup 241}Am-Be isotopic source

Energy Technology Data Exchange (ETDEWEB)

Yücel, Haluk [Ankara University, Institute of Nuclear Sciences, 06100 Tandogan, Ankara (Turkey); Budak, Mustafa Guray, E-mail: mbudak@gazi.edu.tr [Gazi University, Gazi Education Faculty, 06500 Teknikokullar, Ankara (Turkey); Karadag, Mustafa [Gazi University, Gazi Education Faculty, 06500 Teknikokullar, Ankara (Turkey); Yüksel, Alptuğ Özer [Ankara University, Institute of Nuclear Sciences, 06100 Tandogan, Ankara (Turkey)

2014-11-01

Highlights: • An irradiation unit was installed using a 37 GBq {sup 241}Am-Be neutron source. • The source neutrons moderated by using both water and paraffin. • Irradiation unit was shielded by boron oxide and lead against neutrons and gammas. • There are two sites for irradiations, one of them has a pneumatic transfer system. • Cadmium ratio method was used for irradiation site characterization. - Abstract: For the applicability of instrumental neutron activation analysis (NAA) technique, an irradiation unit with a 37 GBq {sup 241}Am-Be neutron source was installed at Institute of Nuclear Sciences of Ankara University. Design and configuration properties of the irradiation unit are described. It has two different sample irradiation positions, one is called site #1 having a pneumatic sample transfer system and the other is site #2 having a location for manual use. In order to characterize neutron flux spectra in the irradiation sites, the measurement results were obtained for thermal (φ{sub th}) and epithermal neutron fluxes (φ{sub epi}), thermal to epithermal flux ratio (f) and epithermal spectrum shaping factors (α) by employing cadmium ratios of gold (Au) and molybdenum (Mo) monitors. The activities produced in these foils were measured by using a p-type, 44.8% relative efficiency HPGe well detector. For the measured γ-rays, self-absorption and true coincidence summing effects were taken into account. Additionally, thermal neutron self-shielding and resonance neutron self-shielding effects were taken into account in the measured results. For characterization of site #1, the required parameters were found to be φ{sub th} = (2.11 ± 0.05) × 10{sup 3} n cm{sup −2} s{sup −1}, φ{sub epi} = (3.32 ± 0.17) × 10{sup 1} n cm{sup −2} s{sup −1}, f = 63.6 ± 1.5, α = 0.045 ± 0.009, respectively. Similarly, those parameters were measured in site #2 as φ{sub th} = (1.49 ± 0.04) × 10{sup 3} n cm{sup −2} s{sup −1}, φ{sub epi} = (2.93 ± 0

Report on neutron reflectometry for the Australian Replacement Reactor

International Nuclear Information System (INIS)

James, M.

2001-01-01

There is a clear need for at least one neutron reflectometer at the Australian Replacement Research Reactor when it commences operation in 2005. The participants at the reflectometry workshop have identified that the neutron reflectometer to be built at the Australian Replacement Research Reactor must be capable of the study of: 1. Specular scattering from air/solid, solid/liquid and in particular 'free liquid' samples; and 2. Off-specular' scattering from the above sample types. 3. Kinetics phenomena on a minute or slower time scale; 4. A range of samples of differing thicknesses, ranging from ultra-thin films to thousand angstrom thick films. In order to achieve this the reflectometer should have the capacity to vary its resolution. Interest was also expressed at the ability to conduct glancing-angle and wide-angle scattering studies for the investigation of short length scale, in-plane structures. There was little interest expressed by the workshop participants for polarised neutron reflectometry. This report contains a scientific case for a neutron reflectometer to be built at the Australian Replacement Research Reactor on a cold neutron guide, which is based on the areas of scientific research expressed by the workshop participants. In addition, trends in neutron reflectometry research conducted at major overseas neutron facilities are noted. The new neutron Reflectometer should: 1. Be based on the Time-of-Flight method; 2. Have a vertical scattering plane (i.e. operate for horizontal samples); 3. Be located on the end of a cold neutron guide, or be built off the guide axis using a bender, 4. Have a position sensitive area detector, 5. Be similar in spirit to the new D17 reflectometer at the ILL. Basic aspects of a reflectometer design are discussed which meet the above-stated scientific criteria and include a preliminary list of instrument specifications, capabilities and ancillary equipment requested by the workshop participants. A preliminary instrument

Determination of vanadium, manganese and tungsten in steels with an 241 Am-Be isotopic neutron source

International Nuclear Information System (INIS)

Galdino, S.M.L.

1985-09-01

A non-destructive neutron activation method was developed for determination of vanadium, manganese, and tungsten in alloy-steel, with the aid of an Am-Be 1,85x10 11 Bq(5Ci) isotopic neutron source, employing NaI (T1) detector well type 2x2 in. The 51 V (n,γ) 52 V, 55 Mn (n,γ) 56 Mn, and 186 W (n,γ) 187 W nuclear reactions are induced in steel samples subject to activation by thermal neutron. After irradiation, the activity of the samples was measured by γ-spectrometry under the 1434 KeV 52 V, 847KeV 56 Mn, and 686 KeV 187 W photopeaks. Possible interferences due to other radionuclides activity were investigated by determining the 52 V, 56 Mn, and 187 W half-lifes. The time of analysis for vanadium determination was 11 min, with 1,5% of precision and 3,4% of average absolute deviation. The time of analysis for manganese determination was 22,8 min with 4,0% of precision and 3,4% of average absolute deviation. The time of analysis for tungsten determination was 44,62 min with 3,8% of precision and 3,1% of average absolute deviation. The activation analysis method is adequated for steel quality control in industry. (Author) [pt

Determination of boron in water solution by an indirect neutron activation technique from a 241Am/Be source

International Nuclear Information System (INIS)

Sales, H.B.

1981-08-01

Boron content in water solutions has been analysed by Indirect Activation Technique a twin 241 Am/Be neutron source with a source strength of 9x10 6 n/seg. The boron concentration was inferred from the measurement of the activity induced in a vanadium flux monitor. The vanadium rod was located inside the boron solution in a standart geometrical set up with respect to the neutron source. Boron concentrations in the range of 100 to 1000 ppm were determined with an overall accuracy of about 2% during a total analysis time of about 20 minutes. Eventhough the analysis is not selective for boron yet due the rapid, simple and precise nature, it is proposed for the analysis of boron in the primary coolant circuit of Nuclear Power Plants of PWR type. (Author) [pt

Neutron beam facilities at the replacement research reactor

International Nuclear Information System (INIS)

Kennedy, S.

1999-01-01

Full text: On September 3rd 1997 the Australian Federal Government announced their decision to replace the HIFAR research reactor by 2005. The proposed reactor will be a multipurpose reactor with improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The neutron beam facilities are intended to cater for Australian scientific needs well into the 21st century. In the first stage of planning the neutron Beam Facilities at the replacement reactor, a Consultative Group was formed (BFCG) to determine the scientific capabilities of the new facility. Members of the group were drawn from academia, industry and government research laboratories. The BFCG submitted their report in April 1998, outlining the scientific priorities to be addressed. Cold and hot neutron sources are to be included, and cold and thermal neutron guides will be used to position most of the instruments in a neutron guide hall outside the reactor confinement building. In 2005 it is planned to have eight instruments installed with a further three to be developed by 2010, and seven spare instrument positions for development of new instruments over the life of the reactor. A beam facilities technical group (BFTG) was then formed to prepare the engineering specifications for the tendering process. The group consisted of some members of the BFCG, several scientists and engineers from ANSTO, and scientists from leading neutron scattering centres in Europe, USA and Japan. The BFTG looked in detail at the key components of the facility such as the thermal, cold and hot neutron sources, neutron collimators, neutron beam guides and overall requirements for the neutron guide hall. The report of the BFTG, completed in August 1998, was incorporated into the draft specifications for the reactor project, which were distributed to potential reactor vendors. An assessment of the first stage of reactor vendor submissions was completed in

Neutron capture and (n,2n) measurements on {sup 241}Am

Energy Technology Data Exchange (ETDEWEB)

Vieira, D.J.; Jandel, M.; Bredeweg, T.A.; Bond, E.M.; Clement, R.R.; Couture, A.; Haight, R.C.; O' Donnell, J.M.; Reifarth, R.; Rundberg, R.S.; Ullmann, J.L.; Wilhelmy, J.B.; Wouters, J.M. [Los Alamos National Laboratory, NM (United States); Tonchev, A.P.; Hutcheson, A.; Angell, C.T.; Crowell, A.S.; Fallin, B.; Hammond, S.; Howell, C.R.; Karowowski, H.J.; Kelley, J.H.; Pedroni, R.; Tornow, W. [Triangle Univ. Nuclear Laboratory, Durham, NC (United States); Macri, R.A.; Agvaanluvsan, U.; Becker, J.A.; Dashdorj, D.; Stoyer, M.A.; Wu, C.Y. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

2008-07-01

We report on a set of neutron-induced reaction measurements on {sup 241}Am which are important for nuclear forensics and advanced nuclear reactor design. Neutron capture measurements have been performed on the DANCE detector array at the Los Alamos Neutron Scattering Center (LANSCE). In general, good agreement is found with the most recent data evaluations up to an incident neutron energy of about 300 keV where background limits the measurement. Using mono-energetic neutrons produced in the {sup 2}H(d,n){sup 3}He reaction at the Triangle Universities Nuclear Laboratory (TUNL), we have measured the {sup 241}Am(n,2n) excitation function from 7.6 to 14.5 MeV using the activation method. Good agreement is found with previous measurements, with the exception of the three data points reported by Perdikakis et al. around 11 MeV, where we obtain a much lower cross section that is more consistent with theoretical estimates. (authors)

Summary of alpha-neutron sources in GADRAS

International Nuclear Information System (INIS)

Mitchell, Dean James; Thoreson, Gregory G.; Harding, Lee T.

2012-01-01

A common source of neutrons for calibration and testing is alpha-neutron material, named for the alpha-neutron nuclear reaction that occurs within. This material contains a long-lived alpha-emitter and a lighter target element. When the alpha particle from the emitter is absorbed by the target, neutrons and gamma rays are released. Gamma Detector Response and Analysis Software (GADRAS) includes built-in alpha-neutron source definitions for AcC, AmB, AmBe, AmF, AmLi, CmC, and PuC. In addition, GADRAS users may create their own alpha-neutron sources by placing valid alpha-emitters and target elements in materials within their one-dimensional models (1DModel). GADRAS has the ability to use pre-built alpha-neutron sources for plotting or as trace-sources in 1D models. In addition, if any material (existing or user-defined) specified in a 1D model contains both an alpha emitter in conjunction with a target nuclide, or there is an interface between such materials, then the appropriate neutron-emission rate from the alpha-neutron reaction will be computed. The gamma-emissions from these sources are also computed, but are limited to a subset of nine target nuclides. If a user has experimental data to contribute to the alpha-neutron gamma emission database, it may be added directly or submitted to the GADRAS developers for inclusion. The gadras.exe.config file will be replaced when GADRAS updates are installed, so sending the information to the GADRAS developers is the preferred method for updating the database. This is also preferable because it enables other users to benefit from your efforts.

Use of CITATION code for flux calculation in neutron activation analysis with voluminous sample using an Am-Be source

International Nuclear Information System (INIS)

Khelifi, R.; Idiri, Z.; Bode, P.

2002-01-01

The CITATION code based on neutron diffusion theory was used for flux calculations inside voluminous samples in prompt gamma activation analysis with an isotopic neutron source (Am-Be). The code uses specific parameters related to the energy spectrum source and irradiation system materials (shielding, reflector). The flux distribution (thermal and fast) was calculated in the three-dimensional geometry for the system: air, polyethylene and water cuboidal sample (50x50x50 cm). Thermal flux was calculated in a series of points inside the sample. The results agreed reasonably well with observed values. The maximum thermal flux was observed at a distance of 3.2 cm while CITATION gave 3.7 cm. Beyond a depth of 7.2 cm, the thermal flux to fast flux ratio increases up to twice and allows us to optimise the detection system position in the scope of in-situ PGAA

Opportunities for research using neutron beams at Australia's replacement research reactor

International Nuclear Information System (INIS)

Robinson, R.A.

2000-01-01

Full text: On July 13th 2000, a contract was signed for construction of Australia's Replacement Research Reactor at Lucas Heights just outside Sydney. This may represent Australia's largest single investment in scientific infrastructure, and it provides researchers in condensed matter physics, chemistry, materials science, and some aspects of engineering, the earth sciences and biology with the 'opportunity of a generation' The replacement reactor, which will commence operation in 2005, will be comparable with the national neutron sources of Japan, France and the U.S.A. Cold and thermal neutron sources are to be installed and supermirror guides will transport cold and thermal neutron beams into a large modern guide hall. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by the Argentinian company INVAP S.E., in collaboration with two Australian firms, in a turnkey contract. The instruments will be developed by ANSTO and other contracted organisations, in consultation with the Australian user community and interested overseas parties. This presentation reviews the planned scientific capabilities and opportunities, gives a description of the facility and a status report on the activities so far

Neutron spectra and dosimetric features of isotopic neutron sources: a review

International Nuclear Information System (INIS)

Vega C, H. R.; Martinez O, S. A.

2015-10-01

A convenient way to produce neutrons is the isotopic neutron source, where the production is through (α, n), (γ, n), and spontaneous fission reactions. Isotopic neutron sources are small, easy to handle, and have a relative low cost. On the other hand the neutron yield is small and mostly of them produces neutrons with a wide energy distribution. In this work, a review is carried out about the the main features of 24 NaBe, 24 NaD 2 O, 116 InBe, 140 LaBe, 238 PuLi, 239 PuBe, 241 AmB, 241 AmBe, 241 AmF, 241 AmLi, 242 CmBe, 210 PoBe, 226 RaBe, 252 Cf and 252 Cf/D 2 O isotopic neutron source. Also, using Monte Carlo methods, the neutron spectra in 31 energy groups, the neutron mean energy; the Ambient dose equivalent, the Personal dose equivalent and the Effective dose were calculated for these isotopic neutron sources. (Author)

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)

Detection limits of pollutants in water for PGNAA using Am-Be source

International Nuclear Information System (INIS)

Khelifi, R.; Amokrane, A.; Bode, P.

2007-01-01

A basic PGNAA facility with an Am-Be neutron source is described to analyze the pollutants in water. The properties of neutron flux were determined by MCNP calculations. In order to determine the efficiency curve of a HPGe detector, the prompt-gamma rays from chlorine were used and an exponential curve was fitted. The detection limits for typical water sample are also estimated using the statistical fluctuations of the background level in the areas of recorded the prompt-gamma spectrum

$^{11}$Be($\\beta$p), a quasi-free neutron decay?

CERN Document Server

Riisager, K.; Borge, M.J.G.; Briz, J.A.; Carmona-Gallardo, M.; Fraile, L.M.; Fynbo, H.O.U.; Giles, T.; Gottberg, A.; Heinz, A.; Johansen, J.G.; Jonson, B.; Kurcewicz, J.; Lund, M.V.; Nilsson, T.; Nyman, G.; Rapisarda, E.; Steier, P.; Tengblad, O.; Thies, R.; Winkler, S.R.

2014-01-01

We have observed $\\beta$-delayed proton emission from the neutron-rich nucleus $^{11}$Be by analysing a sample collected at the ISOLDE facility at CERN with accelerator mass spectrometry (AMS). With a branching ratio of (8.4 $\\pm$ 0.6)$\\times$ 10$^{-6}$ the strength of this decay mode, as measured by the B$_\\mathrm{GT}$-value, is unexpectedly high. The result is discussed within a simple single-particle model and could be interpreted as a quasi-free decay of the $^{11}$Be halo neutron into a single-proton state.

Total and (n, 2n) neutron cross section measurements on 241Am

International Nuclear Information System (INIS)

Sage, C.

2009-01-01

Neutron induced reaction cross sections on 241 Am have been measured at the IRMM in Geel, Belgium, in the frame of a collaboration between the EC Joint Research Centres IRMM and ITU and French laboratories from CNRS and CEA. Raw material coming from the Atalante facility of CEA Marcoule has been transformed into suitable AmO 2 samples embedded in Al 2 O 3 and Y 2 O 3 matrices. The irradiations for the 241 Am(n, 2n) 240 Am reaction cross section measurement were carried out at the 7 MV Van de Graaff accelerator using the activation technique with quasi mono-energetic neutrons from 8 to 21 MeV produced via the D(d, n) 3 He and the T(d, n) 4 He reactions. The cross section was determined relative to the 27 Al(n, α) 24 Na standard cross section and was investigated for the first time above 15 MeV. The induced activity was measured off-line by standard γ-ray spectrometry using a high purity Ge detector. A special effort was made for the estimation of the uncertainties and the correlations between our experimental points. A different sample of the same isotope 241 Am has been measured in transmission and capture experiments in the resolved resonance region at the neutron ToF facility GELINA. The transmission measurement was performed in two campaigns, with an upgrade of the whole data acquisition system in between, followed by an investigation of its new performances. A preliminary analysis of the resonance parameters tends to confirm the recent evaluation to a higher value for the cross section at the bottom of the first resonances. A new design of C 6 D 6 detectors for capture measurements has been studied, but the data reduction and analysis of the measurement are not part of this work. (author) [fr

Neutron spectra and dosimetric features of isotopic neutron sources: a review

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas, Zac. (Mexico); Martinez O, S. A., E-mail: fermineutron@yahoo.com [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Av. Central del Norte 39-115, 150003 Tunja, Boyaca (Colombia)

2015-10-15

A convenient way to produce neutrons is the isotopic neutron source, where the production is through (α, n), (γ, n), and spontaneous fission reactions. Isotopic neutron sources are small, easy to handle, and have a relative low cost. On the other hand the neutron yield is small and mostly of them produces neutrons with a wide energy distribution. In this work, a review is carried out about the the main features of {sup 24}NaBe, {sup 24}NaD{sub 2}O, {sup 116}InBe, {sup 140}LaBe, {sup 238}PuLi, {sup 239}PuBe, {sup 241}AmB, {sup 241}AmBe, {sup 241}AmF, {sup 241}AmLi, {sup 242}CmBe, {sup 210}PoBe, {sup 226}RaBe, {sup 252}Cf and {sup 252}Cf/D{sub 2}O isotopic neutron source. Also, using Monte Carlo methods, the neutron spectra in 31 energy groups, the neutron mean energy; the Ambient dose equivalent, the Personal dose equivalent and the Effective dose were calculated for these isotopic neutron sources. (Author)

Aluminium and copper analysis in metallic alloys by neutron activation analysis from an 241 Am-Be source

International Nuclear Information System (INIS)

Carvalho, J. de.

1980-01-01

Aluminium and copper have been determined in aluminium alloys by the method of activation with neutrons from an 241 Am-Be source of intensity 9,8 x 10 6 n/s. The activity induced due to reactions 27 Al (n, γ) 28 Al and 63 Cu (n, γ) 64 Cu have been measured with a NaI (Tl) detector coupled to a single channel system. In order to obtain the samples and standards of about the same composition, the material to be irradiated was powdered. In view of low intensity of neutron source it was necessary to use samples of up to 50 g. A series of preliminary irradiations were carried out to ensure that the geometry for the irradiation and for the counting are reproducible. The results have been compared with those obtained by chemical methods. Assuming that the results obtained by chemical method is exact, a maximum relative error of 3,6% is obtained by this method. The method has a good reproducibility. The time needed for analysis of aluminium and copper are 18 min and 2 hours 40 minutes respectively. Four different samples were analysed. The average of five measurements for one of the samples was: 88.0% for aluminium and 10.0% for copper. The standard deviation and coefficient of variation were 0,8 and 1.0% for aluminium and 0,2 and 2.0% for copper. (author)

Comparison between {sup 241}AmBe (α, n) sources in the calibration of survey monitors for neutrons in magnitude H⁎(10); Comparação entre fontes de {sup 241}AmBe (α,n) na calibração de monitores de área para nêutrons na grandeza H⁎(10)

Energy Technology Data Exchange (ETDEWEB)

Pereira, W.W.; Filho, C.A.L.S.; Silva, F.S.; Patrão, K.C.S., E-mail: walsanwagner@gmail.com [Instituto de Radioproteção e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Estrada, J.J.S. [Fundação Técnico Educacional Souza Marques, Rio de Janeiro, RJ (Brazil); Alves, C.F.E.; Magalhães, L.A.G.; Leite, S.P. [Universidade do Estado do Rio de Janeiro (UERJ), RJ (Brazil)

2017-07-01

Recently the Laboratório de Ciências Radiológicas (LCR / UERJ) characterized a {sup 241}AmBe (α, n) neutron source for metrological use. This source comes from the Oil Company and it was originally used in well logging tools. In this work, this source was used in the calibration of an survey meter using procedures and facilities of the Laboratório Nacional de Metrologia das Radiações Ionizantes (LN/LNMRI). The results of the calibration with the LCR neutron source agreed with those obtained when the LN/LNMRI neutron source was used. (author)

Measurement of thermal neutron cross section for {sup 241}Am(n,f) reaction

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Katsuhei; Yamamoto, Shuji; Fujita, Yoshiaki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.; Miyoshi, Mitsuharu; Kimura, Itsuro; Kanno, Ikuo; Shinohara, Nobuo

1997-03-01

Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the {sup 241}Am(n,f) reaction has been measured relative to the reference value of 586.2b for the {sup 235U}(n,f) reaction. For the present measurement, electrodeposited layers of {sup 241}Am and {sup 235}U have been employed as back-to-back type double fission chambers. The present result at neutron energy of 0.0253 eV is 3.15 {+-} 0.097b. The ENDF/B-VI data is in good agreement with the present value, while the JENDL-3.2 data is lower by 4.2%. The evaluated data in JEF-2.2 and by Mughabghab are higher by 0.9% and 1.6%, respectively than the present result. The ratios of the earlier experimental data to the present value are distributed between 0.89 and 1.02. (author)

Replacement of the moderator cell unit of JRR-3's cold neutron source facility

International Nuclear Information System (INIS)

Hazawa, Tomoya; Nagahori, Kazuhisa; Kusunoki, Tsuyoshi

2006-10-01

The moderator cell of the JRR-3's cold neutron source (CNS) facility, converts thermal neutrons into cold neutrons by passing through liquid cold hydrogen. The cold neutrons are used for material and life science research such as the neutron scattering. The CNS has been operated since the start of JRR-3's in 1990. The moderator cell containing liquid hydrogen is made of stainless steel. The material irradiation lifetime is limited to 7 years due to irradiation brittleness. The first replacement was done by using a spare part made in France. This replacement work of 2006 was carried out by using the domestic moderator cell unit. The following technologies were developed for the moderator cell unit production. 1) Technical development of black treatment on moderator cell surface to increase radiation heat. 2) Development of bending technology of concentric triple tubes consisting from inside tube, Outside tube and Vacuum insulation tube. 3) Development of manufacturing technique of the moderator cell with complicated shapes. According to detail planed work procedures, replacement work was carried out. As results, the working days were reduced to 80% of old ones. The radiation dose was also reduced due to reduction of working days. It was verified by measurement of neutrons characteristics that the replaced moderator cell has the same performance as that of the old moderator cell. The domestic manufacturing of the moderator cell was succeeded. As results, the replacement cost was reduced by development of domestic production technology. (author)

Neutron capture cross section measurements of $^{238}$U, $^{241}$Am and $^{243}$Am at n_TOF

CERN Multimedia

Koehler, P E; Plag, R

The increase of the world energy demand and the need of low carbon energy sources have triggered the renaissance and/or enhancement of nuclear energy in many countries. Fundamental nuclear physics can contribute in a practical way to the sustainability and safety of the nuclear energy production and the management of the nuclear waste. There exists a series of recent studies which address the most relevant isotopes, decay data, nuclear reaction channels and energy ranges which have to be investigated in more detail for improving the design of different advanced nuclear systems [1] and nuclear fuel cycles [2]. In this proposal, we aim at the measurement of the neutron capture cross sections of $^{238}$U, $^{241}$Am and $^{243}$Am. All three isotopes are listed in the NEA High Priority Request List [37], are recommended for measurements [1] and play an important role in the nuclear energy production and fuel cycle scenarios. The measurements will provide as well valuable nuclear structure data necessary for the...

Maslov. Evaluated neutron reaction data for Am and Cm isotopes. Summary documentation

International Nuclear Information System (INIS)

Lemmel, H.D.

1996-01-01

The nuclear data file by V.M. Maslov et al includes evaluated neutron reaction data for Am-241,243 Cm-243-245,246 in EDNF-6 format. The data are available from the IAEA Nuclear Data Section, costfree upon request. (author)

Neutron-induced reactions on U and Th - A new approach via AMS

International Nuclear Information System (INIS)

Wallner, A.; Capote, R.; Christl, M.; Fifield, L.K.; Srncik, M.; Tims, S.; Hotchkis, M.; Krasa, A.; Lachner, J.; Lippold, J.; Plompen, A.; Semkova, V.; Steier, P.; Winkler, S.

2014-01-01

Recent studies exhibit discrepancies at keV and MeV energies between major nuclear data libraries for 238 U(n,γ), 232 Th(n,γ) and also for (n,xn) reactions. We have extended our initial (n,γ) measurements on 235,238 U to higher neutron energies and to additional reaction channels. Neutron-induced reactions on 232 Th and 238 U were measured by a combination of the activation technique and atom counting of the reaction products using accelerator mass spectrometry (AMS). Natural thorium and uranium samples were activated with quasi-monoenergetic neutrons at IRMM. Neutron capture data were produced for neutron energies between 0.5 and 5 MeV. Fast neutron-induced reactions were studied in the energy range from 17 to 22 MeV. Preliminary data indicate a fair agreement with data libraries; however at the lower band of existing data. This approach represents a complementary method to on-line particle detection techniques and also to conventional decay counting. (authors)

Radiochemical determination of the neutron capture cross sections of {sup 241}Am irradiated in the JMTR reactor

Energy Technology Data Exchange (ETDEWEB)

Shinohara, N.; Hatsukawa, Y.; Hata, K.; Kohno, N. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

The thermal neutron capture cross section {sigma}{sub 0} and Resonance integral I{sub 0} of {sup 241}Am leading to the production of {sup 242m}Am and {sup 242g}Am were measured by radiochemical method. The cross sections obtained in this study are {sigma}{sub 0}=60.9 {+-} 2.6 barn, I{sub 0}=213 {+-} 13 barn for {sup 241}Am(n,{gamma}){sup 242m}Am and {sigma}{sub 0}=736 {+-} 31 barn, I{sub 0}=1684 {+-} 92 barn for {sup 241}Am(n,{gamma}){sup 242g}Am. (author)

Neutron-induced fission cross-section of 233U, 241Am and 243Am in the energy range 0.5 MeV ≤ En ≤ 20 MeV

International Nuclear Information System (INIS)

Belloni, F.; Milazzo, P.M.; Calviani, M.

2011-01-01

Neutron-induced fission cross-sections of 233 U, 241 Am and 243 Am relative to 235 U have been measured in a wide energy range at the neutron time of flight facility n-TOF in Geneva to address the present discrepancies in evaluated and experimental databases for reactions and isotopes relevant for transmutation and new generation fast reactors. A dedicated fast ionization chamber was used. Each isotope was mounted in a different cell of the modular detector. The measurements took advantage of the characteristics of the n-TOF installation. Its intrinsically low background, coupled to its high instantaneous neutron flux, results in high accuracy data. Its wide energy neutron spectrum helps to reduce systematic uncertainties due to energy-domain matching problems while the 185 m flight path and a 6 ns pulse width assure an excellent energy resolution. This paper presents results obtained between 500 keV and 20 MeV neutron energy. (authors)

Report on inelastic neutron scattering for the Australian Replacement Research Reactor

International Nuclear Information System (INIS)

Cussen, L.

2001-01-01

The Replacement Reactor Project includes a sub project, the Neutron Beam Instrument project, to construct a suite of eight leading edge neutron scattering instruments to be ready at reactor start up in 2005. Amongst these will be an inelastic scattering instrument and this workshop explored the options that would best serve the future needs of the Australian neutron scattering community.Considering the dynamic range of each of these instruments and the ability to address the needs of the widest group of problems, the workshop endorsed the recommendation of the Beam Facilities Consultative Group to build a thermal neutron three-axis spectrometer, which offers the project the best opportunity and flexibility in terms of the available wave-vector and energy range. It was also recommend that priority consideration be given to developing cold neutron three-axis and time-of-flight spectrometers, perhaps in collaboration with outside organisations, after completion of the initial project. thermal neutron three axis spectrometer. The workshop recommended that such an instrument should be built at the reactor face on a thermal beam, using a double-focussing monochromator and analyser. Polarisation analysis should be available as an option and that the instrument perform at or close to world's best level. Recommendations were received on sample environment requirements and low temperature, high temperature and strong magnetic fields are seen as necessary

Application Research of Quasi-monochromatic X-ray Machine Replacing 241Am Radioactive Source in Thickness Measurement

Directory of Open Access Journals (Sweden)

LIN Hui1;XIAO Xue-fu2;HOU Yue-xin3;ZHAO Jing1;JIAN Li-min1

2014-02-01

Full Text Available In the research of BS-03 thickness detector, a self-designed quasi-monochromatic (50-60 keV X-ray machine was studied as a substitution to Am-241 low energy photon source（1.11×109 Bq）The range of output current, the attenuation of X rays through the steel plate, the aluminum plate, and the organic glass plate and stability of quasi-monochromatic X-ray machine was tested. The result showed that quasi-monochromatic X-ray machine could be an applicable replacement of Am-241 radioactive source.

Medical observation and evaluation of persons accidentally exposed to 241Am-Be neutron source

International Nuclear Information System (INIS)

Li Jinbang; Bai Yushu; Zhang Jiayu

1986-01-01

A radiation accident caused by an 241 Am-Be source is reported. A boy, 17 years old, brought an 241 Am-Be source of 120 mCi to home and drilled a hole in it. In all 10 persons were irradiated with varying doses and were contaminated to different extent. The maximum dose of external exposure was 121 x 10 -3 Gy and the maximum concentration of urinary 241 Am was 2.8 dpm. The exposed persons felt dizziness, palpitation and myalgia at the early stage. At that time there were slight increase in leucocyte counts, presence of heterogeneous lymphocytes and increase in frequency of leucocyte spinous process. Follow-up study for one and a half years on general health condition, hematological changes, immune function (T-and B-lymphocytes) and crystalline lens alterations revealed no significant differences between the irradiated and the non-irradiated persons, but the frequencies of all types of chromosome aberrations in the irradiated group were significantly higher than those in the control group

Review of fission product yields and delayed neutron data for the actinides NP-237, PU-242, AM-242M, AM-243, CM-243 and CM-245

International Nuclear Information System (INIS)

Mills, R.W.

1990-07-01

A review of fission product yields and delayed neutron data for Np-237, Pu-242, Am-242m, Am-243, Cm-243 and Cm-245 has been undertaken. Gaps in understanding and inconsistencies in existing data were identified and priority areas for further experimental, theoretical and evaluation investigation detailed

Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, {sup 235}U)O{sub 2} fuel matrixes

Energy Technology Data Exchange (ETDEWEB)

Gholamzadeh, Z. [Talca Univ. (Chile). Dept. of Physics; Feghhi, S.A.H. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Dept. of Radiation Application

2014-11-15

Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th-{sup 235}U)O{sub 2} on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of {sup 241}Am, {sup 243}Am, {sup 235}U and {sup 232}Th respectively. 132.8 kg of {sup 233}U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of {sup 241/243}Am, {sup 235}U and {sup 232}Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional {sup 233}U production.

Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.

Science.gov (United States)

Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y

2003-10-01

A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

Precise measurement of the neutron capture reaction 54Fe(n,γ)55Fe via AMS

International Nuclear Information System (INIS)

Wallner, A; Buczak, K; Forstner, O; Golser, R; Kutschera, W; Lederer, C; Priller, A; Steier, P; Belgya, T; Szentmiklosi, L; Bichler, M; Coquard, L; Dillmann, I; Kaeppeler, F; Mengoni, A; Reifarth, R

2010-01-01

The measurement of cross sections relevant to nuclear astrophysics has become one main research topic at the VERA (Vienna Environmental Research Accelerator) facility. The technique applied, accelerator mass spectrometry (AMS), offers excellent sensitivity for the detection of long-lived radionuclides through ultra-low isotope ratio measurements. We discuss the potential and preliminary results of ongoing precision measurements of neutron-capture cross sections of 54 Fe. Such measurements might help to clarify the recently found discrepancy of s-process nucleosynthesis at lower-mass nuclei (A 55 Fe (t 1/2 = 2.72 yr) was analyzed using AMS. At VERA, detection of 55 Fe was developed with a reproducibility of about 1%, which makes the 54 Fe(n,γ) 55 Fe reaction a precise and unique laboratory measurement, which can serve as a reference to complementary techniques. In this regard a new 55 Fe standard for AMS measurements was produced. The final cross-section data are expected to be accurate to better than 3%. We report a preliminary, however, already significantly improved thermal neutron cross section value of (2.32 ± 0.10) barn, and a value of (6.3 ± 0.6) mbarn for E n = (520 ± 50) keV.

Monte Carlo study of a flexible device for in situ PGNAA using 241Am-Be source: application to total chlorine determination

International Nuclear Information System (INIS)

Khelifi, R.; Bode, P.

2016-01-01

MCNP5 has been used to optimize the design of a Prompt gamma ray neutron activation analysis (PGNAA) facility, which was subsequently constructed for quantification of total chlorine in water to simulate neutron transport from an 241 AmBe source into a PGNAA set-up. Modeling calculations were performed to optimize the experimental set-up for Cl measurements in water. The optimization with MCNP5 was focused on maximizing the thermal neutrons flux which leads to improving the gamma prompt production after neutron capture in a water sample. The influence of dimensions and materials for the neutron collimation as well as the dimensions of the sample together were studied. A PGNAA facility with an 241 AmBe neutron source was built based on the optimized configuration and used to determine chlorine concentration. Measured values of the chlorine count rate were plotted versus the NaCl in water. The count rate versus amount of chlorine show a good coefficient of correlation of the linear fit. The result permits PGNAA to be a valuable diagnostic tool for getting an indication of the salinity contamination of water. (author)

Studies of Be migration in the JET tokamak using AMS with 10Be marker

Science.gov (United States)

Bykov, I.; Bergsåker, H.; Possnert, G.; Zhou, Y.; Heinola, K.; Pettersson, J.; Conroy, S.; Likonen, J.; Petersson, P.; Widdowson, A.

2016-03-01

The JET tokamak is operated with beryllium limiter tiles in the main chamber and tungsten coated carbon fiber composite tiles and solid W tiles in the divertor. One important issue is how wall materials are migrating during plasma operation. To study beryllium redistribution in the main chamber and in the divertor, a 10Be enriched limiter tile was installed prior to plasma operations in 2011-2012. Methods to take surface samples have been developed, an abrasive method for bulk Be tiles in the main chamber, which permits reuse of the tiles, and leaching with hot HCl to remove all Be deposited at W coated surfaces in the divertor. Quantitative analysis of the total amount of Be in cm2 sized samples was made with inductively coupled plasma atomic emission spectroscopy (ICP-AES). The 10Be/9Be ratio in the samples was measured with accelerator mass spectrometry (AMS). The experimental setup and methods are described in detail, including sample preparation, measures to eliminate contributions in AMS from the 10B isobar, possible activation due to plasma generated neutrons and effects of diffusive isotope mixing. For the first time marker concentrations are measured in the divertor deposits. They are in the range 0.4-1.2% of the source concentration, with moderate poloidal variation.

Boron-Coated Straw Collar for Uranium Neutron Coincidence Collar Replacement

International Nuclear Information System (INIS)

Hu, Jianwei; Croft, Stephen; McElroy, Robert Dennis

2017-01-01

The objective of this project was to design and optimize, in simulation space, an active neutron coincidence counter (or collar) using boron-coated straws (BCSs) as a non- 3 He replacement to the Uranium Neutron Coincidence Collar (UNCL). UNCL has been used by the International Atomic Energy Agency (IAEA) and European Atomic Energy Community (Euratom) since the 1980s to verify the 235 U content in fresh light water reactor fuel assemblies for safeguards purposes. This report documents the design and optimization of the BCS collar.

Boron-Coated Straw Collar for Uranium Neutron Coincidence Collar Replacement

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Croft, Stephen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-01-01

The objective of this project was to design and optimize, in simulation space, an active neutron coincidence counter (or collar) using boron-coated straws (BCSs) as a non-³He replacement to the Uranium Neutron Coincidence Collar (UNCL). UNCL has been used by the International Atomic Energy Agency (IAEA) and European Atomic Energy Community (Euratom) since the 1980s to verify the ²³⁵U content in fresh light water reactor fuel assemblies for safeguards purposes. This report documents the design and optimization of the BCS collar.

Simultaneous measurement of neutron-induced fission and capture cross sections for {sup 241}Am at neutron energies below fission threshold

Energy Technology Data Exchange (ETDEWEB)

Hirose, K., E-mail: hirose.kentaro@jaea.go.jp [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nishio, K.; Makii, H.; Nishinaka, I.; Ota, S. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nagayama, T. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Engineering, Ibaraki University, Mito 310-0056 (Japan); Tamura, N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Goto, S. [Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Andreyev, A.N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Vermeulen, M.J. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Gillespie, S.; Barton, C. [Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Kimura, A.; Harada, H. [Nuclear Science and Engineering Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Meigo, S. [J-PARC Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Chiba, S. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo 152-8550 (Japan); Ohtsuki, T. [Research Reactor Institute, Kyoto University, Kumatori-cho S' ennangun,Osaka 590-0494 (Japan)

2017-06-01

Fission and capture reactions were simultaneously measured in the neutron-induced reactions of {sup 241}Am at the spallation neutron facility of the Japan Proton Accelerator Research Complex (J-PARC). Data for the neutron energy range of E{sub n}=0.1–20 eV were taken with the TOF method. The fission events were observed by detecting prompt neutrons accompanied by fission using liquid organic scintillators. The capture reaction was measured by detecting γ rays emitted in the deexcitation of the compound nuclei using the same detectors, where the prompt fission neutrons and capture γ rays were separated by a pulse shape analysis. The cross sections were obtained by normalizing the relative yields at the first resonance to evaluations or other experimental data. The ratio of the fission to capture cross sections at each resonance is compared with those from an evaluated nuclear data library and other experimental data. Some differences were found between the present values and the library/literature values at several resonances.

Improvement of evaluated neutron nuclear data for {sup 237}Np and {sup 241}Am

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Tsuneo; Iwamoto, Osamu; Hasegawa, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-01-01

The nuclear data of {sup 237}Np and {sup 241}Am that are particularly important among the minor actinides were investigated by comparing JENDL-3.2 with the recent evaluated data and available experimental data. As a result of the study, several defects of JENDL-3.2 data were revealed. They were improved on the basis of experimental data or recent evaluated data. For the both nuclides, main quantities revised in the present work were the resonance parameters, cross sections, angular and energy distributions of secondary neutrons, number of neutrons per fission. The data were given in the neutron energy range from 10{sup -5} eV to 20 MeV, and compiled in the ENDF-6 format. (author)

Neutron beam design for low intensity neutron and gamma-ray radioscopy using small neutron sources

CERN Document Server

Matsumoto, T

2003-01-01

Two small neutron sources of sup 2 sup 5 sup 2 Cf and sup 2 sup 4 sup 1 Am-Be radioisotopes were used for design of neutron beams applicable to low intensity neutron and gamma ray radioscopy (LINGR). In the design, Monte Carlo code (MCNP) was employed to generate neutron and gamma ray beams suited to LINGR. With a view to variable neutron spectrum and neutron intensity, various arrangements were first examined, and neutron-filter, gamma-ray shield and beam collimator were verified. Monte Carlo calculations indicated that with a suitable filter-shield-collimator arrangement, thermal neutron beam of 3,900 ncm sup - sup 2 s sup - sup 1 with neutron/gamma ratio of 7x10 sup 7 , and 25 ncm sup - sup 2 s sup - sup 1 with very large neutron/gamma ratio, respectively, could be produced by using sup 2 sup 5 sup 2 Cf(122 mu g) and a sup 2 sup 4 sup 1 Am-Be(37GBq)radioisotopes at the irradiation port of 35 cm from the neutron sources.

Evaluation of neutron nuclear data for 243Am

International Nuclear Information System (INIS)

Igarasi, Sin-iti; Nakagawa, Tsuneo

1977-06-01

Evaluation of neutron nuclear data for 243 Am was performed below 16 MeV. The energy region above 250 eV was separated from the lower region where the resonance parameters were given. Evaluation was made to select suitable resonance parameters, and thermal values of the capture and fission cross sections were obtained with the adopted resonance parameters. An average fission width was assumed to bridge the cross sections at 0.0253 eV and above 250 eV. Using a semi-empirical formula, the fission cross section was reproduced above 250 eV. Optical and statistical model calculations were made in order to obtain the total, capture, inelastic and elastic scattering, and (n,2n) reaction cross sections. (auth.)

Comparison Of ²⁵²Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Joshi, Jay Prakash [Los Alamos National Laboratory

2017-03-30

The effective application of international safeguards to research reactors requires verification of spent fuel as well as fresh fuel. To accomplish this goal various nondestructive and destructive assay techniques have been developed in the US and around the world. The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) system developed at Los Alamos National Laboratory (LANL) combining both neutron and gamma measurement capabilities. Since spent fuel assemblies are stored in water, the system was designed to be watertight to facilitate underwater measurements by inspectors. The AEFC is comprised of six ³He detectors as well as a shielded and collimated ion chamber. The ³He detectors are used for active and passive neutron coincidence counting while the ion chamber is used for gross gamma counting. Active coincidence measurement data is used to measure residual fissile mass, whereas the passive coincidence measurement data along with passive gamma measurement can provide information about burnup, cooling time, and initial enrichment. In the past, most of the active interrogation systems along with the AEFC used an AmLi neutron interrogation source. Owing to the difficulty in obtaining an AmLi source, a ²⁵²Cf spontaneous fission (SF) source was used during a 2014 field trail in Uzbekistan as an alternative. In this study, experiments were performed to calibrate the AEFC instrument and compare use of the ²⁵²Cf spontaneous fission source and the AmLi (α,n) neutron emission source. The ²⁵²Cf source spontaneously emits bursts of time-correlated prompt fission neutrons that thermalize in the water and induce fission in the fuel assembly. The induced fission (IF) neutrons are also time correlated resulting in more correlated neutron detections inside the ³He detector, which helps reduce the statistical errors in doubles when using the ²⁵²Cf interrogation source instead of

Neutron activation analysis: Modelling studies to improve the neutron flux of Americium-Beryllium source

Energy Technology Data Exchange (ETDEWEB)

Didi, Abdessamad; Dadouch, Ahmed; Tajmouati, Jaouad; Bekkouri, Hassane [Advanced Technology and Integration System, Dept. of Physics, Faculty of Science Dhar Mehraz, University Sidi Mohamed Ben Abdellah, Fez (Morocco); Jai, Otman [Laboratory of Radiation and Nuclear Systems, Dept. of Physics, Faculty of Sciences, Tetouan (Morocco)

2017-06-15

Americium–beryllium (Am-Be; n, γ) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

Detection and dosimetry studies on the response of silicon diodes to an 241Am-Be source

International Nuclear Information System (INIS)

Lotfi, Y; Dizaji, H Zaki; Davani, F Abbasi

2014-01-01

Silicon diode detectors show potential for the development of an active personal dosimeter for neutron and photon radiation. Photons interact with the constituents of the diode detector and produce electrons. Fast neutrons interact with the constituents of the diode detector and converter, producing recoil nuclei and causing (n,α) and (n,p) reactions. These photon- and neutron-induced charged particles contribute to the response of diode detectors. In this work, a silicon pin diode was used as a detector to produce pulses created by photon and neutron. A polyethylene fast neutron converter was used as a recoil proton source in front of the detector. The total registered photon and neutron efficiency and the partial contributions of the efficiency, due to interactions with the diode and converter, were calculated. The results show that the efficiency of the converter-diode is a function of the incident photon and neutron energy. The optimized thicknesses of the converter for neutron detection and neutron dosimetry were found to be 1 mm and 0.1 mm respectively. The neutron records caused by the (n,α) and (n,p) reactions were negligible. The photon records were strongly dependent upon the energy and the depletion layer of the diode. The photons and neutrons efficiency of the diode-based dosimeter was calculated by the MCNPX code, and the results were in good agreement with experimental results for photons and neutrons from an 241 Am-Be source

A survey of neutron energy spectra and angular distributions of the 9Be(p,n)9B reaction for fast neutron radiotherapy

International Nuclear Information System (INIS)

Allab, M.

1984-03-01

Encouraging findings in radiobiology have stimulated a renewed use of fast neutrons in radiotherapy. The physical characteristics required for neutron beams to be suitable for radiotherapy are well established. As a result, the tendency is to replace the previous machines which generated the neutron beams from deuteron bombardment of thick targets (T, Li, Be) by hospital based cyclotrons which accelerate protons on thick Beryllium targets. This report surveys the available experimental data of the 9 Be(p,n) reaction (cross sections, neutron spectra, yields, mean neutron energies) from the threshold to the proton energy Esub(p)=120 MeV and the works using this reaction in dosimetry measurements, with an emphasis on the data since 1977

Effect of sample moisture and bulk density on performance of the 241Am-Be source based prompt gamma rays neutron activation analysis setup. A Monte Carlo study

International Nuclear Information System (INIS)

Almisned, Ghada

2010-01-01

Monte Carlo simulations were carried out using the dependence of gamma ray yield on the bulk density and moisture content for five different lengths of Portland cement samples in a thermal neutron capture based Prompt Gamma ray Neutron Activation Analysis (PGNAA) setup for source inside moderator geometry using an 241 Am-Be neutron source. In this study, yields of 1.94 and 6.42 MeV prompt gamma rays from calcium in the five Portland cement samples were calculated as a function of sample bulk density and moisture content. The study showed a strong dependence of the 1.94 and 6.42 MeV gamma ray yield upon the sample bulk density but a weaker dependence upon sample moisture content. For an order of magnitude increase in the sample bulk density, an order of magnitude increase in the gamma rays yield was observed, i.e., a one-to-one correspondence. In case of gamma ray yield dependence upon sample moisture content, an order of magnitude increase in the moisture content of the sample resulted in about 16-17% increase in the yield of 1.94 and 6.42 MeV gamma rays from calcium. (author)

Neutron activation analysis: Modelling studies to improve the neutron flux of Americium–Beryllium source

Directory of Open Access Journals (Sweden)

Abdessamad Didi

2017-06-01

Full Text Available Americium–beryllium (Am-Be; n, γ is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci, yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

Response of CsI:Pb Scintillator Crystal to Neutron Radiation

Science.gov (United States)

Costa Pereira, Maria da Conceição; Filho, Tufic Madi; Berretta, José Roberto; Náhuel Cárdenas, José Patrício; Iglesias Rodrigues, Antonio Carlos

2018-01-01

The helium-3 world crisis requires a development of new methods of neutron detection to replace commonly used 3He proportional counters. In the past decades, great effort was made to developed efficient and fast scintillators to detect radiation. The inorganic scintillator may be an alternative. Inorganic scintillators with much higher density should be selected for optimal neutron detection efficiency taking into consideration the relevant reactions leading to light emission. These detectors should, then, be carefully characterized both experimentally and by means of advanced simulation code. Ideally, the detector should have the capability to separate neutron and gamma induced events either by amplitude or through pulse shape differences. As neutron sources also generate gamma radiation, which can interfere with the measurement, it is necessary that the detector be able to discriminate the presence of such radiation. Considerable progress has been achieved to develop new inorganic scintillators, in particular increasing the light output and decreasing the decay time by optimized doping. Crystals may be found to suit neutron detection. In this report, we will present the results of the study of lead doped cesium iodide crystals (CsI:Pb) grown in our laboratory, using the vertical Bridgman technique. The concentration of the lead doping element (Pb) was studied in the range 5x10-4 M to 10-2 M . The crystals grown were subjected to annealing (heat treatment). In this procedure, vacuum of 10-6 mbar and continuous temperature of 350°C, for 24 hours, were employed. In response to neutron radiation, an AmBe source with energy range of 1 MeV to 12 MeV was used. The activity of the AmBe source was 1Ci Am. The fluency was 2.6 x 106 neutrons/second. The operating voltage of the photomultiplier tube was 1700 V; the accumulation time in the counting process was 600 s and 1800 s. The scintillator crystals used were cut with dimensions of 20 mm diameter and 10 mm height.

Characterization of a sealed Americium-Beryllium (AmBe) source by inductively coupled plasma mass spectrometry

International Nuclear Information System (INIS)

Sommers, J.; Jimenez, M.; Adamic, M.; Giglio, J.; Carney, K.

2009-01-01

Two Americium-Beryllium neutron sources were dismantled, sampled (sub-sampled) and analyzed via inductively coupled plasma mass spectrometry (ICP-MS). Characteristics such as 'age' since purification, actinide content, trace metal content and inter and intra source composition were determined. The 'age' since purification of the two sources was determined to be 25.0 and 25.4 years, respectively. The systematic uncertainties in the 'age' determination were ±4% 2σ. The amount and isotopic composition of U and Pu varied substantially between the sub-samples of Source 2 (n = 8). This may be due to the physical means of sub-sampling or the way the source was manufactured. Source 1 was much more consistent in terms of content and isotopic composition (n = 3 sub-samples). The Be-Am ratio varied greatly between the two sources. Source 1 had an Am-Be ratio of 6.3 ± 52% (1σ). Source 2 had an Am-Be ratio of 9.81 ± 3.5% (1σ). In addition, the trace element content between the samples varied greatly. Significant differences were determined between Sources 1 and 2 for Sc, Sr, Y, Zr, Mo, Ba and W. (author)

Neutron dosimetry; Dosimetria de neutrons

Energy Technology Data Exchange (ETDEWEB)

Fratin, Luciano

1993-12-31

A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

Neutron dosimetry; Dosimetria de neutrons

Energy Technology Data Exchange (ETDEWEB)

Fratin, Luciano

1994-12-31

A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

Bubble detectors as a tool of the dosimetry and microdosimetry in neutron fields

International Nuclear Information System (INIS)

Spurny, F.; Vlcek, B.; Rannou, A.

1998-01-01

Two types of bubble detector were studied: the Bubble Damage Neutron Detector (BDND) and the Superheated Drop Detector (SDD). The detectors were tested in neutron beams and fields. The relative response of the detectors varied with the average neutron energy. The response of SDD 100 started to decrease at higher energies than for BDND's, at 100 keV it was only about 1/4 of the response to AmBe neutrons. The responses of SDD 1000 and SDD 6000 decreased with the average neutron energy in a rather similar way. Starting from the AmLi source they represented less than 0.1 of the response to AmBe neutrons. Their response to high energy neutrons was practically the same as to AmBe neutrons. This is important for individual air crew dosimetry on board aircraft. (M.D.)

Neutron radiative capture by the 241Am nucleus in the energy range 1 keV-20 MeV

International Nuclear Information System (INIS)

Zolotarev, K.I.; Ignatyuk, A.V.; Tolstikov, V.A.; Tertychnyj, G.Ya.

1998-01-01

Production of high actinides leads to many technological problems in the nuclear power. The 241 Am(n,γ) 242 Am reaction is one of the sources of high actinide buildup. So a knowledge of the radiative capture cross-section of 241 Am for neutron energies up to 20 MeV is of considerable important for present day fission reactors and future advanced reactors. The main goal of this paper is the evaluation of the excitation function for the reaction 241 Am(n,γ) 242 Am in the energy range 1 keV-20 MeV. The evaluation was done on the basis of analysed experimental data, data from theoretical model calculations and systematic predictions for 14.5 MeV and 20 MeV. Data from the present evaluation are compared with the cross-section values given in the evaluations carried out earlier. (author)

Activation measurements for fast neutrons. Part B. 63Ni measurements by accelerator mass spectrometry (AMS)

International Nuclear Information System (INIS)

Straume, Tore; Rugel, Georg; Marchetti, Alfredo A.

2005-01-01

The AMS-measurements of 63 Ni in copper samples obtained from different distances in Hiroshima represent the first detection of fast neutrons after more than 50 years, and the first reliable bomb-induced fast neutron measurements beyond 700 m from the hypocenter. A major significance of these 63 Ni results is that they provide, for the first time, fast neutron measurements at the distances most relevant to atomic-bomb survivor locations (900-1,500 m). The measurement of 63 Ni in copper samples using AMS provides a substantial improvement in fast neutron detection compared with the 32 P measurements made in 1945. For example, the measurement-to-background ratio at about 400 m from the hypocenter is about 55 for 63 Ni and only about 2 for the 32 P measurements. Also, the 63 Ni measurements reach background levels at about 1,800 m from the hypocenter compared with only about 700 m for 32 P. At distances of ∼1,800 m from the hypocenter, out to at least 5,000 meters, the measurements appear to level off at a value on the order of 7x10 4 atoms of 63 Ni/g of Cu, suggesting a background of approximately that magnitude. When this background is subtracted and the resulting data corrected to 1945, the measured 63 Ni in copper samples from Hiroshima are in good agreement with DS02 sample-specific calculations. Comparisons with DS86 calculations also show good agreement except at the Bank of Japan. The difference at that distance is significant when compared with DS86 calculations. (J.P.N.)

Thermal neutron capture cross-section measurements of 243Am and 242Pu using the new mini-INCA α- and γ-spectroscopy station

International Nuclear Information System (INIS)

Marie, F.; Letourneau, A.; Fioni, G.; Deruelle, O.; Veyssiere, Ch.; Faust, H.; Mutti, P.; AlMahamid, I.; Muhammad, B.

2006-01-01

In the framework of the Mini-INCA project, dedicated to the study of Minor Actinide transmutation process in high neutron fluxes, an α- and γ-spectroscopy station has been developed and installed at the High Flux Reactor of the Laue-Langevin Institut. This set-up allows short irradiations as well as long irradiations in a high quasi-thermal neutron flux and post-irradiation spectroscopy analysis. It is well suited to measure precisely, in reference to 59 Co cross-section, neutron capture cross-sections, for all the actinides, in the thermal energy region. The first measurements using this set-up were done on 243 Am and 242 Pu isotopes. Cross-section values, at E n =0.025eV, were found to be (81.8+/-3.6)b for 243 Am and (22.5+/-1.1)b for 242 Pu. These values differ from evaluated data libraries by a factor of 9% and 17%, respectively, but are compatible with the most recent measurements, validating by the way the experimental apparatus

Beam plug replacement and alignment under high radiation conditions for cold neutron facilities at Hanaro

International Nuclear Information System (INIS)

Yeong-Garp, Cho; Jin-Won, Shin; Jung-Hee, Lee; Jeong-Soo, Ryu

2010-01-01

Full text : The HANARO, an open-tank-in-pool type research reactor of a 30 MWth power in Korea, has been operating for 15 years since its initial criticality in February 1995. The beam port assigned for the cold neutron at HANARO had been used for an 8-m SANS without neutron guides until it was replaced by a cold neutron guide system in 2008. It was developed a cold neutron guide system for the delivery of cold neutrons from the cold neutron source in the reactor to the neutron scattering instruments in the guide hall. Since the HANARO has been operated from 1995, it was a big challenge to replace the existing plug and shutter with the new facilities under high radiation conditions. When the old plug was removed from the beam port in 2008, the radiation level was 230 mSv/hr at the end of beam port. In addition to that, there were more difficult situations such as the poor as-built dimensions of the beam port, limited work space and time constraint due to other constructions in parallel in the reactor hall. Before the removal of the old plug the level of the radiation was measured coming out through a small hole of the plug to estimate the radiation level during the removal of the old plug and installation of a new plug. Based on the measurement and analysis results, special tools and various shielding facilities were developed for the removal of old in-pile plug and the installation of the new in-pile plug assembly safely. In 2008, the old plug and shutter were successfully replaced by the new plug and shutter as shown in this article with a minimum exposure to the workers. A laser tracker system was also one of the main factors in our successful installation and alignment under high radiation conditions and limited work space. The laser tracker was used to measure and align all the mechanical facilities and the neutron guides with a minimum radiation exposure to workers. The alignment of all the guides and accessories were possible during reactor operation because

Thermal neutron capture cross-section and resonance integral measurements of {sup 139}La(n, γ){sup 140}La and {sup 140}Ce(n, γ){sup 141}Ce using a Am-Be neutron source

Energy Technology Data Exchange (ETDEWEB)

Panikkath, Priyada; Mohanakrishnan, P. [Manipal University, Manipal Centre for Natural Sciences, Karnataka (India)

2017-03-15

Thermal neutron capture cross-sections and resonance integrals of {sup 139}La(n, γ){sup 140}La and {sup 140}Ce (n, γ){sup 141}Ce are measured with respect to reference reactions {sup 197}Au(n, γ){sup 198}Au and {sup 55}Mn(n, γ){sup 56}Mn using the neutron activation technique. Measurements are carried out using neutrons from an Am-Be source located inside a concrete bunker. Two different methods are used for determining self-shielding factors of activation foils as well as for finding the epithermal neutron spectrum shape factor. For {sup 139}La with reference to {sup 197}Au and {sup 55}Mn the measured thermal cross sections are 9.24 ± 0.25 b and 9.28 ± 0.37 b, respectively, while the measured resonance integrals are 12.18 ± 0.67 b and 11.81 ± 0.94 b, respectively. For {sup 140}Ce with reference to {sup 197}Au and {sup 55}Mn the measured thermal cross sections are 0.44 ± 0.01 b and 0.44 ± 0.02 b, respectively, while the measured resonance integrals are 0.55 ± 0.03 b and 0.54 ± 0.04 b, respectively. The present measurements are compared with earlier measurements and evaluations. Presently estimated values confirm the established {sup 139}La(n, γ){sup 140}La cross-sections. The presently measured thermal capture cross-section {sup 140}Ce(n, γ){sup 141}Ce, though lower than the evaluated data, is having higher accuracy compared to previous measurements with large uncertainties. The resonance integral measured is higher (like most previous measurements) than most evaluations requiring a revision of the evaluated data. (orig.)

Research Regarding the Manufacturing through AM Technologies of an Implant for Cervical Disc Replacement

Directory of Open Access Journals (Sweden)

Miron-Borzan Cristina Stefana

2017-01-01

Full Text Available Worldwide, accidents are one of the main causes of illness in developed and undeveloped socio-economic countries. Additive Manufacturing (AM technologies bring extremely useful and advantageous applications for the new neurosurgical procedures. Because the surgical insertion of devices for cervical disc replacement is very difficult, the development of new devices that can minimize these disadvantages, are needed. The aim of this paper was to improve the characteristics of an implant for cervical intervertebral disc replacement, a model based on an existing implant. A cervical cage designed for stabilization and arthrodesis between the cervical vertebrae was analyzed. A new design of a cage that have some improvements, useful for patient safety, as well as for facilitating the surgery was developed. The new proposed design was verified through Fine Element Analysis.

Neutron Characterization of Additively Manufactured Components. Workshop Report

Energy Technology Data Exchange (ETDEWEB)

Watkins, Thomas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Payzant, E. Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Babu, Sudarsanam Suresh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

Additive manufacturing (AM) is a collection of promising manufacturing methods that industry is beginning to explore and adopt. Macroscopically complicated and near net shape components are being built using AM, but how the material behaves in service is a big question for industry. Consequently, AM components/materials need further research into exactly what is made and how it will behave in service. This one and a half day workshop included a series of invited presentations from academia, industry and national laboratories (see Appendix A for the workshop agenda and list of talks). The workshop was welcomed by Alan Tennant, Chief Scientist, Neutron Sciences Directorate, ORNL, and opened remotely by Rob Ivestor, Deputy Director, Advanced Manufacturing Office-DOE, who declared AM adoptees as titans who will be able to create customized 3-D structures with 1 million to 1 billion micro welds with locally tailored microstructures. Further he stated that characterization with neutrons is key to be able to bring critical insight/information into the AM process/property/behavior relationship. Subsequently, the presentations spanned a slice of the current state of the art AM techniques and many of the most relevant characterization techniques using neutrons. After the talks, a panel discussion was held; workshop participants (see Appendix B for a list of attendees) providing questions and the panel answers. The main purpose of the panel discussion was to build consensus regarding the critical research needs in AM that can be addressed with neutrons. These needs were placed into three categories: modes of access for neutrons, new capabilities needed, new AM material issues and neutrons. Recommendations from the workshop were determined based on the panel discussion.

Measurement of thermal neutron cross-sections and resonance integrals for sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As by using sup 2 sup 4 sup 1 Am-Be isotopic neutron source

CERN Document Server

Karadag, M; Tan, M; Oezmen, A

2003-01-01

Thermal neutron cross-sections and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions were measured by the activation method. The experimental samples with and without a cylindrical Cd shield case in 1 mm wall thickness were irradiated in an isotropic neutron field of the sup 2 sup 4 sup 1 Am-Be neutron source. The induced activities in the samples were measured by high-resolution gamma-ray spectrometry with a calibrated reverse-electrode germanium detector. Thermal neutron cross-sections for 2200 m/s neutrons and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions have been obtained relative to the reference values, sigma sub 0 =13.3+-0.1 b and I sub 0 =14.0+-0.3 b for the sup 5 sup 5 Mn(n,gamma) sup 5 sup 6 Mn reaction as a single comparator. The necessary correction factors for gamma attenuation, thermal neutron and resonance neutron self-shielding effects were taken into...

Measurement of 241Am Ground State Radiative Neutron Capture Cross Section with Cold Neutron Beam. Progress Report on Research Contract HUN14318 for the CRP on Minor Actinide Neutron Reaction Data (MANREAD)

International Nuclear Information System (INIS)

Belgya, T.; Szentmiklosi, L.; Kis, Z.; Nagy, N.M.; Konya, J.

2012-01-01

The ground state cross section of 242 Am has been measured with beams of cold neutrons at the Budapest Research Reactor using the X-ray emission of the decay product of 242 Pu. This methodology avoids the uncertainty caused by resonance neutrons in the pile activations. The target was characterized with gamma and X-ray spectrometry. The obtained ground state cross section is 540 ± 32 b, which is at the low end of the most recent literature values, but agrees with most of them within their uncertainty. (author)

Measurement of the removal sechtion of biological protection concretes in the spectrum of an AmBe source

International Nuclear Information System (INIS)

Sandin, C.; Tenrreiro, J.; Desdin, L.; Sarria, P.; Monne, G.; Tellez, E.; Csikai, J

1992-01-01

A method to measure the neutron removal cross sections in cuban concrete for biological protection using the average spectrum of a radioisotopic source was developed. The method was verified by the measurements of different materials with know removal cross sections like Pb, paraffin and water. These spetial cuban concrete are elaborated on the basis of arids of limestone, limonite, magnetite, baryte and serpentine. The aim of this work is to contribute to the characterization of those concretes for its use in cuban nuclear installations. In the experiments were used an AmBe source with an strength of 10 6 n/s. and a long counter detector with associated electronics. It can be seen in the table that the concretes of magnetite and limonite have the best shielding properties for neutrons

Study for correction of neutron scattering in the calibration of the albedo individual monitor from the Neutron Laboratory (LN), IRD/CNEN-RJ, Brazil

International Nuclear Information System (INIS)

Freitas, B.M.; Silva, A.X. da

2014-01-01

The Instituto de Radioprotecao e Dosimetria (IRD) runs a neutron individual monitoring service with albedo type monitor and thermoluminescent detectors (TLD). Moreover the largest number of workers exposed to neutrons in Brazil is exposed to 241 Am-Be fields. Therefore a study of the response of albedo dosemeter due to neutron scattering from 241 Am-Be source is important for a proper calibration. In this work, it has been evaluated the influence of the scattering correction in two distances at the Low Scattering Laboratory of the Neutron Laboratory of the Brazilian National Laboratory (Lab. Nacional de Metrologia Brasileira de Radiacoes Ionizantes) in the calibration of that albedo dosemeter for a 241 Am-Be source. (author)

Radiation Source Replacement Workshop

Energy Technology Data Exchange (ETDEWEB)

Griffin, Jeffrey W.; Moran, Traci L.; Bond, Leonard J.

2010-12-01

This report summarizes a Radiation Source Replacement Workshop in Houston Texas on October 27-28, 2010, which provided a forum for industry and researchers to exchange information and to discuss the issues relating to replacement of AmBe, and potentially other isotope sources used in well logging.

The comparison of four neutron sources for Prompt Gamma Neutron Activation Analysis (PGNAA) in vivo detections of boron.

Science.gov (United States)

Fantidis, J G; Nicolaou, G E; Potolias, C; Vordos, N; Bandekas, D V

A Prompt Gamma Ray Neutron Activation Analysis (PGNAA) system, incorporating an isotopic neutron source has been simulated using the MCNPX Monte Carlo code. In order to improve the signal to noise ratio different collimators and a filter were placed between the neutron source and the object. The effect of the positioning of the neutron beam and the detector relative to the object has been studied. In this work the optimisation procedure is demonstrated for boron. Monte Carlo calculations were carried out to compare the performance of the proposed PGNAA system using four different neutron sources ( 241 Am/Be, 252 Cf, 241 Am/B, and DT neutron generator). Among the different systems the 252 Cf neutron based PGNAA system has the best performance.

Measurement of fast neutron induced fission cross sections of 232Th, 238U, 237Np and 243Am

International Nuclear Information System (INIS)

Kanda, Kazutaka; Sato, Osamu; Yoshida, Kazuo; Imaruoka, Hiromitsu; Terayama, Hiromichi; Yoshida, Masashi; Hirakawa, Naohiro

1984-01-01

Neutron induced fission cross sections of 232 Th, 238 U, 237 Np and 243 Am relative to 235 U were measured in the energy range from 1.5 to 6.6 MeV. The present results are compared with experimental results of others and evaluated data in JENDL-2 and ENDF/B-IV. (author)

Tl response of LiF:Mg, Cu, P + PTFE to Am-Be neutrons

International Nuclear Information System (INIS)

Gonzalez M, P.R.

2000-01-01

In different laboratories of the world it is followed the research about development of new Tl materials, whose main characteristics should be their equivalence with the tissue and their high sensibility to any type of radiation. The study consists in to measure the Tl peak intensity which TLD-100 presents at being irradiated with neutrons and that appears over 250 Centigrade, for compare it with the Tl intensity of the LiF: Mg, Cu, P + PTFE dosemeters. However, not all dosemeters of the same group show the interesting peak, by this only can be the total Tl intensity of dosemeters studied. In the ININ dosemeters development laboratory, we have developed a Tl material of lithium fluoride activated with magnesium, copper and phosphorus (LiF: Mg, Cu, P) that in polycrystalline powder form is almost 35 times more sensitive than the TLD-100 commercial dosemeter of Harshaw/Filtrol, USA. With the use of polytetrafluorethylene (PTFE) and with the above described Tl material, it has been possible to obtain dosemeters in pellet form of LiF: Mg, Cu, P + PTFE. (Author)

Partial neutron capture cross sections of actinides using cold neutron prompt gamma activation analysis

International Nuclear Information System (INIS)

Genreith, Christoph

2015-01-01

Nuclear waste needs to be characterized for its safe handling and storage. In particular long-lived actinides render the waste characterization challenging. The results described in this thesis demonstrate that Prompt Gamma Neutron Activation Analysis (PGAA) with cold neutrons is a reliable tool for the non-destructive analysis of actinides. Nuclear data required for an accurate identification and quantification of actinides was acquired. Therefore, a sample design suitable for accurate and precise measurements of prompt γ-ray energies and partial cross sections of long-lived actinides at existing PGAA facilities was presented. Using the developed sample design the fundamental prompt γ-ray data on 237 Np, 241 Am and 242 Pu were measured. The data were validated by repetitive analysis of different samples at two individual irradiation and counting facilities - the BRR in Budapest and the FRM II in Garching near Munich. Employing cold neutrons, resonance neutron capture by low energetic resonances was avoided during the experiments. This is an improvement over older neutron activation based works at thermal reactor neutron energies. 152 prompt γ-rays of 237 Np were identified, as well as 19 of 241 Am, and 127 prompt γ-rays of 242 Pu. In all cases, both high and lower energetic prompt γ-rays were identified. The most intense line of 237 Np was observed at an energy of E γ =182.82(10) keV associated with a partial capture cross section of σ γ =22.06(39) b. The most intense prompt γ-ray lines of 241 Am and of 242 Pu were observed at E γ =154.72(7) keV with σ γ =72.80(252) b and E γ =287.69(8) keV with σ γ =7.07(12) b, respectively. The measurements described in this thesis provide the first reported quantifications on partial radiative capture cross sections for 237 Np, 241 Am and 242 Pu measured simultaneously over the large energy range from 45 keV to 12 MeV. Detailed uncertainty assessments were performed and the validity of the given uncertainties was

A simple neutron-gamma discriminating system

International Nuclear Information System (INIS)

Liu Zhongming; Xing Shilin; Wang Zhongmin

1986-01-01

A simple neutron-gamma discriminating system is described. A detector and a pulse shape discriminator are suitable for the neutron-gamma discriminating system. The influence of the constant fraction discriminator threshold energy on the neutron-gamma resolution properties is shown. The neutron-gamma timing distributions from an 241 Am-Be source, 2.5 MeV neutron beam and 14 MeV neutron beam are presented

Measurement and analysis of the $^{243}$Am neutron capture cross section at the n_TOF facility at CERN

CERN Document Server

Mendoza, E; Guerrero, C; Berthoumieux, E; Abbondanno, U; Aerts, G; Alvarez-Velarde, F; Andriamonje, S; Andrzejewski, J; Assimakopoulos, P; Audouin, L; Badurek, G; Balibrea, J; Baumann, P; Becvar, F; Belloni, F; Calvino, F; Calviani, M; Capote, R; Carrapico, C; Carrillo de Albornoz, A; Cennini, P; Chepel, V; Chiaveri, E; Colonna, N; Cortes, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillmann, I; Dolfini, R; Domingo-Pardo, C; Dridi, W; Duran, I; Eleftheriadis, C; Ferrant†, L; Ferrari, A; Ferreira-Marques, R; Fitzpatrick, L; Frais-Koelbl, H; Fujii, K; Furman, W; Goncalves, I; Gonz alez-Romero, E; Goverdovski, A; Gramegna, F; Griesmayer, E; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martinez, A; Igashira, M; Isaev, S; Jericha, E; Kappeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Ketlerov, V; Kerveno, M; Koehler, P; Konovalov, V; Kossionides, E; Krticka, M; Lampoudis, C; Leeb, H; Lindote, A; Lopes, I; Lossito, R; Lozano, M; Lukic, S; Marganiec, J; Marques, L; Marrone, S; Martınez, T; Massimi, C; Mastinu, P; Mengoni, A; Milazzo, P M; Moreau, C; Mosconi, M; Neves, F; Oberhummer, H; O’Brien, S; Oshima, M; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M T; Plag, R; Plompen, A; Plukis, A; Poch, A; Praena, J; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, C; Rudolf, G; Rullhusen, P; Salgado, J; Santos, C; Sarchiapone, L; Savvidis, I; Stephan, C; Tagliente, G; Tain, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarin, D; Vicente, M C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescher, M; Wisshak, K

2014-01-01

Background:The design of new nuclear reactors and transmutation devices requires to reduce the present neutron cross section uncertainties of minor actinides. Purpose: Reduce the $^{243}$Am(n,$\\gamma$) cross section uncertainty. Method: The $^{243}$Am(n,$\\gamma$) cross section has been measured at the n_TOF facility at CERN with a BaF$_{2}$ Total Absorption Calorimeter, in the energy range between 0.7 eV and 2.5 keV. Results: The $^{243}$Am(n,$\\gamma$) cross section has been successfully measured in the mentioned energy range. The resolved resonance region has been extended from 250 eV up to 400 eV. In the unresolved resonance region our results are compatible with one of the two incompatible capture data sets available below 2.5 keV. The data available in EXFOR and in the literature has been used to perform a simple analysis above 2.5 keV. Conclusions: The results of this measurement contribute to reduce the $^{243}$Am(n,$\\gamma$) cross section uncertainty and suggest that this cross section is underestimate...

A comparison of the potential therapeutic gain of p(66)/Be neutrons and d(14)/Be neutrons

International Nuclear Information System (INIS)

Slabbert, Jacobus P.; Theron, Therina; Zoelzer, Friedo; Streffer, Christian; Boehm, Lothar

2000-01-01

Purpose: To determine the relationship between photon sensitivity and neutron sensitivity and between neutron RBE and photon resistance for two neutron modalities (with mean energies of 6 and 29 MeV) using human tumor cell lines spanning a wide range of radiosensitivities, the principal objective being whether or not a neutron advantage can be demonstrated. Methods and Materials: Eleven human tumor cell lines with mean photon inactivation doses of 1.65-4.35 Gy were irradiated with 0-5.0 Gy of p(66)/Be neutrons (mean energy of 29 MeV) at Faure, S.A. and the same plating was irradiated on the same day with 0-10.0 Gy of Cobalt-γ-rays . Twelve human tumor cell lines, many of which were identical with the above selection, and spanning mean photon inactivation doses of 1.75-4.08 Gy, were irradiated with 0-4 Gy of d(14)/Be neutrons (mean energy of 6 MeV) and with 0-10 Gy of 240 kVp X-rays at the Essen Klinikum. Cell survival was determined by the clonogenic assay, and data were fitted to the linear quadratic equation. Results: 1. Using the mean inactivation dose, a significant correlation was found to exist between neutron sensitivity and photon sensitivity. However, this correlation was more pronounced in the Faure beam (r 2 = 0.89 , p ≤ 0.0001) than in the Essen beam (r 2 = 0.65, p = 0.0027). 2. No significant relationship could be established between neutron RBE and photon resistance for both modalities (p = 0.69 and p = 0.07, respectively). 3. Using α-coefficients as a criterion, the neutron sensitivity for the Faure beam correlated with photon sensitivity (p = 0.001), but this did not apply to the Essen beam (p = 0.27). 4. The neutron RBE for the Essen beam derived from α-coefficients showed a steep increase with photon resistance (p = 0.003). In the Faure beam there was no increase of RBE with photon resistance (p = 0.494). Conclusion: Radiobiological differences between high-energy and low-energy neutrons are particularly apparent in the dependence of the

Personal fast neutrons dosimetry using radiophotoluminescent glass

International Nuclear Information System (INIS)

Salem, Y. O.; Nachab, A.; Nourreddine, A.; Roy, C.

2013-06-01

In a previous paper we described a new ambient RPL dosimeter that detects fast neutrons in a mixed n-γ field via (n, p) reactions in a polyethylene converter. In the present study, a personal dosimeter is introduced to enable evaluating the individual dose equivalent H p (10) taking into account the albedo. A calibration factor for estimating H p (10) has been determined from the diminishing angular response as the angle of neutron incidence increases to 60 deg from the normal. MCNPX simulations for 241 Am-Be and 252 Cf neutrons, together with a series of monoenergetic neutron beams from 0.144 to 5 MeV, have been used to characterize the dosimeter response, which agrees well with the experimental 241 Am-Be response. (authors)

Anisotropy of neutron sources of Neutron Metrology Laboratory, IRD, Brazil; Anisotropia de fontes de neutrons do Laboratorio de Metrologia de Neutrons

Energy Technology Data Exchange (ETDEWEB)

Silva, A.C.F.; Silva, F.S.; Leite, S.P.; Creazolla, P.G; Patrão, K.C.S.; Fonseca, E.S. da; Fernandes, S.S.; Pereira, W.W., E-mail: Alexander.camargo@oi.com.br, E-mail: s.felippesouza@gmail.com, E-mail: karla@ird.gov.br, E-mail: walsan@ird.gov.br, E-mail: evaldo@ird.gov.br, E-mail: simonesilvafernandes@gmail.com, E-mail: prycyllacreazolla@gmail.com, E-mail: leitesprk@gmail.com [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Laboratorio Nacional de Metrologia; Fundação Técnico Educacional Souza Marques (FTESM), Rio de Janeiro, RJ (Brazil)

2017-11-01

The anisotropy measurements have as main objective to define the emission of the radiation by different angles of an encapsulated neutron source. The measurements were performed using a Long Accuracy Counter (PLC) Detector in the Low Dispersion Room of the LNMRI / IRD with different neutron sources. Each measurement was made using a support for the source, emulated through an arduino system to rotate it. The carrier is marked with a variation of 5 °, ranging from 0 ° to 360 °, for the work in question only half, 0 ° to 180 ° is used for a total of nineteen steps. In this paper three sources of {sup 241}AmBe (α, n) 5.92 GBq (16 Ci) were used, neutron sources having the following dimensions: 105 mm in height and 31 mm in diameter. The PLC was positioned at a distance of 2 meters from the neutron source and has a radius of 15 cm for the detection area. The anisotropy factor of the {sup 241}AmBe source was 17%. The results in this work will focus mainly on the area of radioprotection and studies that will improve the process of routine measurements in laboratories and instrument calibrations. (author)

Investigation of Response of Several Neutron Surveymeters by a DT Neutron Generator

International Nuclear Information System (INIS)

Kim, Sang In; Jang, In Su; Kim, Jang Lyul; Lee, Jung IL; Kim, Bong Hwan

2012-01-01

Several neutron measuring devices were tested under the neutron fields characterized with two distinct kinds of thermal and fast neutron spectrum. These neutron fields were constructed by the mixing of both thermal neutron fields and fast neutron fields. The thermal neutron field was constructed using by a graphite pile with eight AmBe neutron sources. The fast neutron field of 14 MeV was made by a DT neutron generator. In order to change the fraction of fast neutron fluence rate in each neutron fields, a neutron generator was placed in the thermal neutron field at 50 cm and 150 cm from the reference position. The polyethylene neutron collimator was used to make moderated 14 MeV neutron field. These neutron spectra were measured by using a Bonner sphere system with an LiI scintillator, and dosimetric quantities delivered to neutron surveymeters were determined from these measurement results.

Kinetics of methane-ethane gas replacement in clathrate-hydrates studied by time-resolved neutron diffraction and Raman spectroscopy.

Science.gov (United States)

Murshed, M Mangir; Schmidt, Burkhard C; Kuhs, Werner F

2010-01-14

The kinetics of CH(4)-C(2)H(6) replacement in gas hydrates has been studied by in situ neutron diffraction and Raman spectroscopy. Deuterated ethane structure type I (C(2)H(6) sI) hydrates were transformed in a closed volume into methane-ethane mixed structure type II (CH(4)-C(2)H(6) sII) hydrates at 5 MPa and various temperatures in the vicinity of 0 degrees C while followed by time-resolved neutron powder diffraction on D20 at ILL, Grenoble. The role of available surface area of the sI starting material on the formation kinetics of sII hydrates was studied. Ex situ Raman spectroscopic investigations were carried out to crosscheck the gas composition and the distribution of the gas species over the cages as a function of structure type and compared to the in situ neutron results. Raman micromapping on single hydrate grains showed compositional and structural gradients between the surface and core of the transformed hydrates. Moreover, the observed methane-ethane ratio is very far from the one expected for a formation from a constantly equilibrated gas phase. The results also prove that gas replacement in CH(4)-C(2)H(6) hydrates is a regrowth process involving the nucleation of new crystallites commencing at the surface of the parent C(2)H(6) sI hydrate with a progressively shrinking core of unreacted material. The time-resolved neutron diffraction results clearly indicate an increasing diffusion limitation of the exchange process. This diffusion limitation leads to a progressive slowing down of the exchange reaction and is likely to be responsible for the incomplete exchange of the gases.

High energy neutron radiography

International Nuclear Information System (INIS)

Gavron, A.; Morley, K.; Morris, C.; Seestrom, S.; Ullmann, J.; Yates, G.; Zumbro, J.

1996-01-01

High-energy spallation neutron sources are now being considered in the US and elsewhere as a replacement for neutron beams produced by reactors. High-energy and high intensity neutron beams, produced by unmoderated spallation sources, open potential new vistas of neutron radiography. The authors discuss the basic advantages and disadvantages of high-energy neutron radiography, and consider some experimental results obtained at the Weapons Neutron Research (WNR) facility at Los Alamos

Measurement of the Neutron Capture Cross Sections of $^{233}$U, $^{237}$Np, $^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm with a Total Absorption Calorimeter at n_TOF

CERN Multimedia

Beer, H; Wiescher, M; Cox, J; Rapp, W; Embid, M; Dababneh, S

2002-01-01

Accurate and reliable neutron capture cross section data for actinides are necessary for the poper design, safety regulation and precise performance assessment of transmutation devices such as Fast Critical Reactors or Accelerator Driven Systems (ADS). The goal of this proposal is the measurement of the neutron capture cross sections of $^{233}$U, $^{237}$Np, $^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm at n_TOF with an accuracy of 5~\\%. $^{233}$U plays an essential role in the Th fuel cycle, which has been proposed as a safer and cleaner alternative to the U fuel cycle. The capture cross sections of $^{237}$Np,$^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm play a key role in the design and optimization of a strategy for the Nuclear Waste Transmutation. A high accuracy can be achieved at n_TOF in such measurements due to a combination of features unique in the world: high instantaneous neutron fluence and excellent energy resolution of the facility, innovative Data Acquisition System based on flash ADCs and t...

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.

Partial knee replacement - slideshow

Science.gov (United States)

... page: //medlineplus.gov/ency/presentations/100225.htm Partial knee replacement - series—Normal anatomy To use the sharing ... A.M. Editorial team. Related MedlinePlus Health Topics Knee Replacement A.D.A.M., Inc. is accredited ...

Replacement of core components in the Advanced Test Reactor

International Nuclear Information System (INIS)

Durney, J.L.; Croucher, D.W.

1990-01-01

The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

Replacement of core components in the Advanced Test Reactor

International Nuclear Information System (INIS)

Durney, J.L.; Croucher, D.W.

1989-01-01

The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

The 4π neutron detector CARMEN

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-01

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

A neutron irradiator to perform nuclear activation

International Nuclear Information System (INIS)

Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

2001-01-01

The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

Anisotropy of neutron sources of Neutron Metrology Laboratory, IRD, Brazil

International Nuclear Information System (INIS)

Silva, A.C.F.; Silva, F.S.; Leite, S.P.; Creazolla, P.G; Patrão, K.C.S.; Fonseca, E.S. da; Fernandes, S.S.; Pereira, W.W.

2017-01-01

The anisotropy measurements have as main objective to define the emission of the radiation by different angles of an encapsulated neutron source. The measurements were performed using a Long Accuracy Counter (PLC) Detector in the Low Dispersion Room of the LNMRI / IRD with different neutron sources. Each measurement was made using a support for the source, emulated through an arduino system to rotate it. The carrier is marked with a variation of 5 °, ranging from 0 ° to 360 °, for the work in question only half, 0 ° to 180 ° is used for a total of nineteen steps. In this paper three sources of "2"4"1AmBe (α, n) 5.92 GBq (16 Ci) were used, neutron sources having the following dimensions: 105 mm in height and 31 mm in diameter. The PLC was positioned at a distance of 2 meters from the neutron source and has a radius of 15 cm for the detection area. The anisotropy factor of the "2"4"1AmBe source was 17%. The results in this work will focus mainly on the area of radioprotection and studies that will improve the process of routine measurements in laboratories and instrument calibrations. (author)

Characterization of detectors of neutrons from B+ZnS (Ag) as an alternative to 3He detectors

International Nuclear Information System (INIS)

Gonzalez, Juan A.; Suarez, Maria J.; Pujol, Luis; Lorente, Alfredo; Gallego, Eduardo

2013-01-01

The objective of this paper is to present the progress made in the design of prototypes for dynamic detection of neutron detectors based on scintillation of B + ZnS (Ag), which can replace existing 3 He detectors for the detection of illicit traffic of radioactive material and special nuclear material. These detectors B + ZnS (Ag) can be used, together with gamma detectors, PVT and NaI (Tl) also developed in the UPM. Two neutron detectors of different shapes and sizes were characterized using two neutron sources of 241 Am + Be. Were determined depth, overall efficiency, intrinsic efficiency and limit of detection. The results of these tests allow to verify that: 1) two cylindrical detectors B + ZnS (Ag) of 5x68 cm, or 4x15x132 cm rectangular detector can replace the cylindrical detector of 5x180 cm 3 He currently employed in the arcades. 2) the dynamic detection limit obtained is less than 20000 neutrons per second, when the sample becomes 2 m to 2m/s, with a probability of having no false positive or negative of the 99.99% 3) digital electronics eliminates interference from gamma emissions samples when their dose rate in the neutron detector is 65 μSv/h in less than factor 10 - 8, and keeps its detection limit and 4) two cylindrical detectors with two moderators of different thickness, of 25 to 50 mm of high density polyethylene, allow to measure the average energy of the neutrons

An alternative method for the measurement of neutron flux

Indian Academy of Sciences (India)

dose rate. We have used a 241AmBe neutron source for neutron irradiation, and the neutron dose rate and count rate were ... neutron sources, e.g., for the characterization of superheated droplet detectors (SDD). [1–6]. The SDD is a .... Grants Commission (UGC) for the financial assistance provided for this work. References.

Using the nuclear activation AMS method for determining chlorine in solids at ppb-levels and below

International Nuclear Information System (INIS)

Winkler, Stephan R.; Eigl, Rosmarie; Forstner, Oliver; Martschini, Martin; Steier, Peter; Sterba, Johannes H.; Golser, Robin

2015-01-01

Neutron activation analysis using decay counting of the activated element is a well-established method in elemental analysis. However, for chlorine there is a better alternative to measuring decay of the short-lived activation product chlorine-38 (t 1/2 = 37.24 min) – accelerator mass spectrometry (AMS) of 36 Cl: the relatively high neutron capture cross section of chlorine-35 for thermal neutrons (43.7 b) and combined the AMS technique for chlorine-36 (t 1/2 = 301 ka) allow for determination of chlorine down to ppb-levels using practical sample sizes and common exposure durations. The combination of neutron activation and AMS can be employed for a few other elements (nitrogen, thorium, and uranium) as well. For bulk solid samples an advantage of the method is that lab contamination can be rendered irrelevant. The chlorine-35 in the sample is activated to chlorine-36, and surface chlorine can be removed after the irradiation. Subsequent laboratory contamination, however, will not carry a prominent chlorine-36 signature. After sample dissolution and addition of sufficient amounts of stable chlorine carrier the produced chlorine-36 and thus the original chlorine-35 of the sample can be determined using AMS. We have developed and applied the method for analysis of chlorine in steel samples. The chlorine content of steel is of interest to nuclear industry, precisely because of above mentioned high neutron capture cross section for chlorine-35, which leads to accumulation of chlorine-36 as long-term nuclear waste. The samples were irradiated at the TRIGA Mark II reactor of the Atominstitut in Vienna and the 36 Cl-AMS setup at the Vienna Environmental Research Accelerator (VERA) was used for 36 Cl/Cl analysis.

Using the nuclear activation AMS method for determining chlorine in solids at ppb-levels and below

Energy Technology Data Exchange (ETDEWEB)

Winkler, Stephan R., E-mail: srw@tlabs.ac.za [Universität Wien, Fakultät für Physik, Institut für Isotopenforschung und Kernphysik (Austria); Eigl, Rosmarie [Universität Wien, Fakultät für Chemie, Institut für Anorganische Chemie (Austria); Forstner, Oliver; Martschini, Martin; Steier, Peter [Universität Wien, Fakultät für Physik, Institut für Isotopenforschung und Kernphysik (Austria); Sterba, Johannes H. [Technische Universität Wien, Atominstitut (Austria); Golser, Robin [Universität Wien, Fakultät für Physik, Institut für Isotopenforschung und Kernphysik (Austria)

2015-10-15

Neutron activation analysis using decay counting of the activated element is a well-established method in elemental analysis. However, for chlorine there is a better alternative to measuring decay of the short-lived activation product chlorine-38 (t{sub 1/2} = 37.24 min) – accelerator mass spectrometry (AMS) of {sup 36}Cl: the relatively high neutron capture cross section of chlorine-35 for thermal neutrons (43.7 b) and combined the AMS technique for chlorine-36 (t{sub 1/2} = 301 ka) allow for determination of chlorine down to ppb-levels using practical sample sizes and common exposure durations. The combination of neutron activation and AMS can be employed for a few other elements (nitrogen, thorium, and uranium) as well. For bulk solid samples an advantage of the method is that lab contamination can be rendered irrelevant. The chlorine-35 in the sample is activated to chlorine-36, and surface chlorine can be removed after the irradiation. Subsequent laboratory contamination, however, will not carry a prominent chlorine-36 signature. After sample dissolution and addition of sufficient amounts of stable chlorine carrier the produced chlorine-36 and thus the original chlorine-35 of the sample can be determined using AMS. We have developed and applied the method for analysis of chlorine in steel samples. The chlorine content of steel is of interest to nuclear industry, precisely because of above mentioned high neutron capture cross section for chlorine-35, which leads to accumulation of chlorine-36 as long-term nuclear waste. The samples were irradiated at the TRIGA Mark II reactor of the Atominstitut in Vienna and the {sup 36}Cl-AMS setup at the Vienna Environmental Research Accelerator (VERA) was used for {sup 36}Cl/Cl analysis.

Using the nuclear activation AMS method for determining chlorine in solids at ppb-levels and below

Science.gov (United States)

Winkler, Stephan R.; Eigl, Rosmarie; Forstner, Oliver; Martschini, Martin; Steier, Peter; Sterba, Johannes H.; Golser, Robin

2015-10-01

Neutron activation analysis using decay counting of the activated element is a well-established method in elemental analysis. However, for chlorine there is a better alternative to measuring decay of the short-lived activation product chlorine-38 (t1/2 = 37.24 min) - accelerator mass spectrometry (AMS) of 36Cl: the relatively high neutron capture cross section of chlorine-35 for thermal neutrons (43.7 b) and combined the AMS technique for chlorine-36 (t1/2 = 301 ka) allow for determination of chlorine down to ppb-levels using practical sample sizes and common exposure durations. The combination of neutron activation and AMS can be employed for a few other elements (nitrogen, thorium, and uranium) as well. For bulk solid samples an advantage of the method is that lab contamination can be rendered irrelevant. The chlorine-35 in the sample is activated to chlorine-36, and surface chlorine can be removed after the irradiation. Subsequent laboratory contamination, however, will not carry a prominent chlorine-36 signature. After sample dissolution and addition of sufficient amounts of stable chlorine carrier the produced chlorine-36 and thus the original chlorine-35 of the sample can be determined using AMS. We have developed and applied the method for analysis of chlorine in steel samples. The chlorine content of steel is of interest to nuclear industry, precisely because of above mentioned high neutron capture cross section for chlorine-35, which leads to accumulation of chlorine-36 as long-term nuclear waste. The samples were irradiated at the TRIGA Mark II reactor of the Atominstitut in Vienna and the 36Cl-AMS setup at the Vienna Environmental Research Accelerator (VERA) was used for 36Cl/Cl analysis.

Determination of the silver content in some ancient coins using an Am-Be neutron source

International Nuclear Information System (INIS)

Cosma, C.; Fiat, T.; Znamizovschi, V.; Daraban, L.; Morariu, V.; Boros, D.; Alicu, D.

1985-01-01

The silver content of 40 Roman and Greek coins was determined using neutron activation analysis. with an M-Be thermal source. The pure silver standard had a weigth and shape similar to those of the analyzed coins. A monopchannel spectrometer with a NaI(Tl), 45x40 mm scintillator crystal was used. The coins were placed at the center of the entrance window of the crystal, the maximum measuring geometry being about 2π. A VA-5-968 scintillation probe was used, placed in a VA-H-161 lead shield in order to reduce the background of the detector. The irradiation time, cooling time and measuring time were 2 min, 20 sec and 1 min correspondingly. The ratio of counting rates for the standard material and the background was 2,5. The experimental errors were about 5% for concentrations higher than 50% and increasingly higher for lower silver concentrations

Response of electret dosemeter to slow neutrons

International Nuclear Information System (INIS)

Ghilardi, A.J.P.; Pela, C.A.; Zimmerman, R.L.

1987-01-01

The response of electret dosemeter to slow neutrons exposure is cited, mentioning the preparation and the irradiation of dosemeter with Am-Be source. Some theory considerations about the response of electret dosemeter to slow and fast neutrons are also presented. (C.G.C.) [pt

Brazilian two-component TLD albedo neutron individual monitoring system

Energy Technology Data Exchange (ETDEWEB)

Martins, M.M., E-mail: marcelo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD), Av. Salvador Allende, s/n, CEP: 22780-160, Rio de Janeiro, RJ (Brazil); Mauricio, C.L.P., E-mail: claudia@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD), Av. Salvador Allende, s/n, CEP: 22780-160, Rio de Janeiro, RJ (Brazil); Fonseca, E.S. da, E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD), Av. Salvador Allende, s/n, CEP: 22780-160, Rio de Janeiro, RJ (Brazil); Silva, A.X. da, E-mail: ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao em Engenharia, COPPE/PEN Caixa Postal 68509, CEP: 21941-972, Rio de Janeiro, RJ (Brazil)

2010-12-15

Since 1983, Instituto de Radioprotecao e Dosimetria, Brazil, uses a TLD one-component albedo neutron monitor, which has a single different calibration factor specifically for each installation type. In order to improve its energy response, a two-component albedo monitor was developed, which measure the thermal neutron component besides the albedo one. The two-component monitor has been calibrated in reference neutron fields: thermal, five accelerator-produced monoenergetic beams (70, 144, 565, 1200 and 5000 keV) and five radionuclide sources ({sup 252}Cf, {sup 252}Cf(D{sub 2}O), {sup 241}Am-Be, {sup 241}Am-B and {sup 238}Pu-Be) at several distances. Since January 2008, mainly Brazilian workers who handle neutron sources at different distances and moderation, such as in well logging and calibration facilities are using it routinely.

A large solid angle multiparameter neutron detector

International Nuclear Information System (INIS)

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

1991-01-01

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

Diamondlike carbon can replace beryllium in physics with ultracold neutrons

International Nuclear Information System (INIS)

Atchison, F.; Blau, B.; Daum, M.; Fierlinger, P.; Foelske, A.; Geltenbort, P.; Gupta, M.; Henneck, R.; Heule, S.; Kasprzak, M.; Kuzniak, M.; Kirch, K.; Meier, M.; Pichlmaier, A.; Plonka, Ch.; Reiser, R.; Theiler, B.; Zimmer, O.; Zsigmond, G.

2006-01-01

To complete our study of ultracold neutron (UCN) storage-vessel coatings, we have measured the Fermi potential for neutrons on diamondlike carbon coatings produced by laser induced vacuum arc deposition. A sample with an sp 3 content of 0.45, measured using, for the first time, neutron transmission had a Fermi potential of (249+/-14)neV. A second sample with an sp 3 fraction of 0.67, measured using cold neutron reflectometry, gave (271+/-13)neV. These values complete the demonstration that there is a viable alternative to Be in UCN physics

Measurement of the 241Am neutron capture cross section at the n_TOF facility at CERN

Directory of Open Access Journals (Sweden)

Mendoza E.

2017-01-01

Full Text Available New neutron cross section measurements of minor actinides have been performed recently in order to reduce the uncertainties in the evaluated data, which is important for the design of advanced nuclear reactors and, in particular, for determining their performance in the transmutation of nuclear waste. We have measured the 241Am(n,γ cross section at the n_TOF facility between 0.2 eV and 10 keV with a BaF2 Total Absorption Calorimeter, and the analysis of the measurement has been recently concluded. Our results are in reasonable agreement below 20 eV with the ones published by C. Lampoudis et al. in 2013, who reported a 22% larger capture cross section up to 110 eV compared to experimental and evaluated data published before. Our results also indicate that the 241Am(n,γ cross section is underestimated in the present evaluated libraries between 20 eV and 2 keV by 25%, on average, and up to 35% for certain evaluations and energy ranges.

Neutron spectra of /sup 239/Pu-Be neutron sources

Energy Technology Data Exchange (ETDEWEB)

Kumar, A; Nagarajan, P S [Bhabha Atomic Research Centre, Bombay (India). Div. of Radiation Protection

1977-01-01

Neutron spectra of /sup 239/Pu-Be(..cap alpha..,n) sources have been calculated by using the most recent data on the differential cross sections and angular distributions. The contribution from the multibody break-up reaction /sup 9/Be(..cap alpha..,..cap alpha..n)/sup 8/Be has also been incorporated. Modifications to the primary spectrum due to the secondary interactions in the source such as elastic scattering with beryllium, oxygen and plutonium and the /sup 9/Be(n,2n) and /sup 239/Pu(n,f) reaction have been calculated for different strengths and geometries. The present calculation has shown that the spectrum changes considerably because of these events within the source by way of smearing of peaks and filling up of valleys and raising the low energy part of the spectrum. Increase in H/D value leads to channeling of extra neutrons into the equatorial plane at the cost of the neutrons along the axial direction. The present calculations show that inclusion of secondary interactions to the extent considered in this work does not account completely for the increased intensity in the lower energy end of the measured spectrum.

Experimental evaluation of the neutrons flux of a irradiator with AmBe sources and its possibility of use in materials analysis

International Nuclear Information System (INIS)

Lima, Ruy Barros de

2003-01-01

This work had as a target to determine the irradiator thermal and over cadmium (epithermal and fast) neutrons flux , of the Nuclear Experimental Laboratory of the Nuclear Energy Center (CNEN) - IPEN, and the possibility of its use for Neutron Activation Analysis (NAA) by the absolute method. The neutrons flux quantification was performed indirectly by the gold naked and cadmium-covered foils activation technique. The neutrons flux was determined for two situations: with polyethylene block 5.0 cm thick and without the polyethylene block. The quantification of the elements present in the irradiated samples was obtained after the experimental determination of the incident neutrons flux in the irradiation position of the sample. Flux values along the irradiator axis were determined. Some materials were analyzed, presenting good agreement with reference values. (author)

Report on neutron powder diffraction for the Australian Replacement Research Reactor

International Nuclear Information System (INIS)

Hunter, B.A.

2000-01-01

There is a clear need for two neutron powder diffractometers at the Australian Replacement Research Reactor when it starts operation in 2005. The high-intensity instrument should be capable of measuring a 10mg sample of moderate complexity, or perform single-shot time-resolved experiments with 1-second time slices, or perform stroboscopic measurements with time slices of order 50 microseconds. The high-resolution instrument should have a target resolution of Δd/d∼6x10 -4 , and be capable of collecting data at this resolution within 1-48 hours depending on sample size and crystal complexity. Key questions that need to be answered in the next 9 months include: (1) a detailed study of monochromator options, (2) analysing the detector options for the high-intensity machine and exploring ways in which the solid angle can be maximised for both instruments, (3) whether the instruments are better situated at the reactor face or on super mirror guides, (4) how to integrate the two instruments (physically, if they are only the same guide), and scientifically as regards detailed performance specifications. The user community clearly wants a wide range of sample-environment options, and these are listed in the report. Combinations of these options will be important

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

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)

Fast neutrons dosimetry

International Nuclear Information System (INIS)

Rzyski, B.M.

1977-01-01

A proton recoil technique has been developed for inducing thermoluminescence with incident fast neutrons. CaF 2 was used as the TL phosphor, and cane sugar and polyethylene were used as proton radiators. The phosphor and the hydrogeneous material powders were well mixed, encapsulated in glass tubes and exposed to Am-Be sources, resulting in recoils from incident fast neutrons of energy between 0,25 and 11,25 MeV. The intrinsic response of pure CaF 2 to fast neutrons without a hydrogeneous radiator was checked by using LiF (TLD-700). Glow curves were recorded from room temperature up to 350 0 C after different doses of neutrons and gamma rays of 60 Co. First collision dose due to fast neutrons in tissue like materials such as cane sugar and polyethylene was also calculated [pt

Neutron producing reactions in PuBe neutron sources

Energy Technology Data Exchange (ETDEWEB)

Bagi, János [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU) (Germany); Lakosi, László; Nguyen, Cong Tam [Centre for Energy Research, Hungarian Academy of Sciences, Budapest (Hungary)

2016-01-01

There are a plenty of out-of-use plutonium–beryllium neutron sources in Eastern Europe presenting both nuclear safeguards and security issues. Typically, their actual Pu content is not known. In the last couple of years different non-destructive methods were developed for their characterization. For such methods detailed knowledge of the nuclear reactions taking place within the source is necessary. In this paper we investigate the role of the neutron producing reactions, their contribution to the neutron yield and their dependence on the properties of the source.

Neutron source strength associated with FTR fuel

International Nuclear Information System (INIS)

Boroughs, G.L.; Bunch, W.L.; Johnson, D.L.

1975-01-01

The study presented shows the important effect of shelf life on the neutron source strength anticipated from fuel irradiated in the FTR. The neutron source strength will be enhanced appreciably by extended shelf lives. High neutron source strengths will also be associated with reprocessed LWR plutonium, which is expected to contain a greater abundance of the higher isotopes. The branching ratio and cross section of 241 Am is an important parameter that needs to be defined more precisely to establish calculated values with greater precision

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

Science.gov (United States)

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

2013-10-01

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

Neutron-induced Fission Cross Sections of Am and Cm isotopes (Final Report of Research Contract 14485). Resonance and Fast Neutron Induced Fission Cross Sections of Americium and Curium Nuclides (Third-year Progress Report of Research Contract 14485)

International Nuclear Information System (INIS)

Alekseev, A.A.; Bergman, A.A.; Berlev, A.I.; Koptelov, E.A.; Egorov, A.S.; Samylin, B.F.; Trufanov, A.M.; Fursov, B.I.; Shorin, V.S.

2012-01-01

The neutron induced fission cross sections of Am and Cm isotopes were measured relative to 239 Pu in the neutron energy range from 1 eV to 20 keV at the INR RAS lead slowing down spectrometer LSDS-100. The fission resonance integrals were also estimated using the measured cross section data. The results have been compared with the available experimental and evaluated data. This analysis has shown the present status of the measured fission cross sections and the necessity to revise the evaluated cross sections libraries for the minor actinides. (author)

Response of CR-39 Detector Against Fast Neutron Using D-Polyethylene and H-Polyethylene Radiator

International Nuclear Information System (INIS)

Sofyan, Hasnel

1996-01-01

The research on the response of detector CR-39 by using D-Polyethylene and H-Polyethylene radiator has been carried out. The optimum number of nuclear tracks was found with the use of 30 % NaOH at 80 + 0,5oC for 80 minutes of etching time. The comparison of CR-39 detector response caused by D-Polyethylene radiator against H-Polyethylene radiator of irradiation in air, were found to be 1.18 and 0.84 for 241Am-Be neutron source and neutron source from reactor respectively. For phantom irradiation, the results were found to be 1.75 for 241Am-Be neutron source, and 0.77 for neutron source from reactor

Fast and epithermal neutron radiography using neutron irradiator

International Nuclear Information System (INIS)

Oliveira, Karol A.M. de; Crispim, Verginia R.; Ferreira, Francisco J.O.

2013-01-01

The neutron radiography technique (NR) with neutrons in the energy range fast to epithermal is a powerful tool used in no-destructive inspection of bulky objects of diverse materials, including those rich in hydrogen, oxygen, nitrogen ad carbon. Thus, it can be used to identify, inclusions, voids and thickness differences in materials such as explosive artifacts and narcotics. Aiming at using NR with fast and epithermal neutrons, an Irradiator was constructed by: a 241 Am-Be source, with 5 Ci activity, a collimator with adjustable collimation rate, L/D; and a shield device composed by plates of borated paraffin and iron. The test specimens chosen were a Beam Purity Indicator (BPI) and an Indicator of Visual Resolution (IVR). The neutron radiography images obtained had a resolution of 444.4 μm and 363.6 μm respectively when registered in: 1) the sheet of the nuclear track solid detector, CR-39 type, through X (n,p) Y nuclear reaction; and 2) Kodak Industrex M radiographic film plate in close contact with a boron converter screen, both stored in a Kodak radiographic cassette. (author)

Anisotropy of neutrons sources of the Neutron Metrology Laboratory

International Nuclear Information System (INIS)

Silva, A.C.F.; Silva, F.S.; Creazolla, P.G.; Patrão, K.C.S.; Fonseca, E.S. da; Pereira, W.W.

2017-01-01

The anisotropy measurements have as main objective to define the emission of the radiation by different angles of an encapsulated neutron source. Measurements were performed using a Precision Long Counter (PLC) detector in the Laboratório de Baixo Espalhamento of the LNMRI / IRD. In this study were used an 241 AmBe (α,n) 5.92 GBq and a 238 PuBe (α,n) 1.85 TBq. The anisotropy factor was 8.65% to 241 AmBe and 4.36% to 238 PuBe, due to variations in the source encapsulation. The results in this work will focus mainly on the area of radiation protection and studies that will improve the process of routine measurements in laboratories and instrument calibrations. (author)

Prototype Neutron Energy Spectrometer

International Nuclear Information System (INIS)

Mitchell, Stephen; Mukhopadhyay, Sanjoy; Maurer, Richard; Wolff, Ronald

2010-01-01

The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production (ship effect), (a, n) reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.

Prototype Neutron Energy Spectrometer

Energy Technology Data Exchange (ETDEWEB)

Stephen Mitchell, Sanjoy Mukhopadhyay, Richard Maurer, Ronald Wolff

2010-06-16

The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production [ship effect], [a, n] reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.

Measurements of fast neutrons by bubble detectors

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-03

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

Exposure potential to neutrons of the Brazilian workers

International Nuclear Information System (INIS)

Martins, Marcelo M.; Almeida, Carlos E. de

1995-01-01

Nowadays, there are 222 neutron radioactive sources registered in use in Brazil, in addition to several neutron fields, power and research reactors and neutron generators. Secondary neutron fields can also be generated in particle accelerators by nuclear reactions with its shielding, the experimental set up and the conversion target. These neutron fields are very different, not only in their spectra but also in their fluences. Around 200 radiation workers are monthly monitored since 1983 by an albedo system. Of the evaluated dosimeters 4% only have shown neutron doses, being 13 mSv the maximum measured value. Most of these doses were received by workers of the oil-well logging and research activities users of 241 Am-Be sources. (author). 7 refs., 3 tabs

Neutron spectra of /sup 242/Cm-Be and /sup 244/Cm-Be neutron sources

Energy Technology Data Exchange (ETDEWEB)

Kumar, A; Nagarajan, P S [Bhabha Atomic Research Centre, Bombay (India). Div. of Radiation Protection

1977-02-15

Neutron spectra of /sup 242/Cm-Be(..cap alpha..,n) and /sup 244/Cm-Be(..cap alpha..,n) sources have been calculated including the spontaneous fission contribution which is negligible for /sup 242/Cm and amounts to about 4% for /sup 244/Cm. The agreement of the present work with experimental results is poor.

Neutron emission from deuterium plasma focus

International Nuclear Information System (INIS)

Antanasijevic, R.; Banjanac, R.; Dragic, A.; Djordjevic, D.; Joksimovic, D.; Maric, Z.; Udovicic, V.; Vukovic, J.

1998-01-01

The anisotropy of emitted neutrons is investigated on a small 'Mather-type' plasma focus device (PF). This problem is of importance for determining the nature of the fusion reaction mechanisms. Mica detectors together with thick uranium foils were used for both detection and angular distribution measurements of the neutrons. Previously, the annealing of the detectors was done and measured after the irradiation with neutrons from Am-Be source. Also, annealing ability of H-plasma focus has been tested. Geometry of detectors in both experiments was the same. (authors)

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

Directory of Open Access Journals (Sweden)

SANG IN KIM

2014-04-01

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

Determination of copper in some Myanmar indigenous medicines by neutron activation analysis

International Nuclear Information System (INIS)

Myint-U; Kyi Kyi San

1994-01-01

Copper was determined in two Myanmar indigenous medicines by neutron activation analysis using an Am(Be) radionuclide neutron source. The activity of 511 keV peak of the 64 Cu was measured. (author) 2 refs.; 2 tabs

Determination of copper in some Myanmar indigenous medicines by neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Myint-U,; San, Kyi Kyi [Yangon Univ. (Myanmar). Dept. of Chemistry

1994-09-15

Copper was determined in two Myanmar indigenous medicines by neutron activation analysis using an Am(Be) radionuclide neutron source. The activity of 511 keV peak of the [sup 64]Cu was measured. (author) 2 refs.; 2 tabs.

Thermoresponsive behaviour of AM2O8 materials

International Nuclear Information System (INIS)

Allen, Simon

2003-01-01

This thesis investigates the synthesis and structural characterisation of AM 2 O 8 phases, many of which show negative thermal expansion (NTE); relevant literature is reviewed in Chapter One. Chapter Two describes the synthesis, structure solution, and mechanistic role of a new family of low-temperature (LT) orthorhombic AM 2 O 8 polymorphs (A IV = Zr, Hf; M VI = Mo, W). These materials are key intermediates in the preparation of cubic AM 2 O 8 phases from AM 2 O 7 (OH) 2 (H 2 O) 2 . The structure of LT-AM 2 O 8 has been elucidated by combined laboratory X-ray and neutron powder diffraction. Variable temperature X-ray diffraction (VTXRD) studies have shown LT-AMo 2 O 8 phases exhibit anisotropic NTE. LT-ZrMo 2 O 8 has been shown to undergo spontaneous rehydration, allowing preparation of ZrMo 2 O 7 (OD) 2 (D 2 O) 2 and assignment of D 2 O/OD positions within the structure by neutron diffraction. Using this result, a reversible topotactic dehydration pathway from AM 2 O 7 (OH) 2 (H 2 O) 2 to LT-AM 2 O 8 is proposed. Chapter Three investigates the order-disorder phase transition with concurrent oxygen mobility in cubic AM 2 O 8 materials; studies include comprehensive VT neutron diffraction of cubic ZrMo 2 O 8 to reveal a static to dynamic transition at 215 K, and novel quench-anneal/quench-warm variable temperature/time diffraction experiments on ZrWMoO 8 which lead to an activation energy of 40 kJmol -1 for oxygen migration. In Chapter Four 17 O-labelled cubic ZrW 2 O 8 has been prepared to understand the oxygen migration process by VT MAS NMR. In situ hydrothermal studies of cubic ZrMo 2 O 8 using synchrotron radiation have shown direct hydration to ZrMo 2 O 7 (OH) 2 (H 2 O) 2 . In Chapter Five VTXRD of trigonal α-AMo 2 O 8 phases reveals a previously unknown second-order phase transition at 487 K (A = Zr) or 463 K (A = Hf) from P3-bar 1c to P3-bar m1. Rigid-body Rietveld refinements have shown this is due to alignment of apical Mo-O groups with the c axis in the

Measurement of the 241Am and the 243Am Neutron Capture Cross Sections at the n_TOF Facility at CERN

CERN Document Server

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

2014-01-01

The capture cross sections of Am-241 and Am-243 were measured at the n\\_TOF facility at CERN in the epithermal energy range with a BaF2 Total Absorption Calorimeter. A preliminary analysis of the Am-241 and a complete analysis of the Am-243 measurement, including the data reduction and the resonance analysis, have been performed.

Bulk moisture determination in building materials by fast neutron/gamma technique

International Nuclear Information System (INIS)

Padron Diaz, I.; Felipe Desdin, L.; Martin Hernandez, G.; Shtejer, K.; Perez Tamayo, N.; Ceballos, C.; Lemus, O.

1998-01-01

Fast Neutron/Gamma Transmission technique has been improved to allow to measure moisture content in building materials. In order to improve fast neutron/gamma discrimination in the transmission system employing the NE-213 scintillation detector a pulse shape discrimination system was constructed at the CEADEN. A separate neutron/gamma detection approach was used with neutron transmission measurement using an Am-Be neutron source and a BF 3 detector and gamma transmission measurement using a collimated 137 Cs source and a NaI scintillator

Behavior of 241Am in fast reactor systems - a safeguards perspective

International Nuclear Information System (INIS)

Beddingfield, David H.; Lafleur, Adrienne M.

2009-01-01

Advanced fuel-cycle developments around the world currently under development are exploring the possibility of disposing of 241 Am from spent fuel recycle processes by burning this material in fast reactors. For safeguards practitioners, this approach could potentially complicate both fresh- and spent-fuel safeguards measurements. The increased (α,n) production in oxide fuels from the 241 Am increases the uncertainty in coincidence assay of Pu in MOX assemblies and will require additional information to make use of totals-based neutron assay of these assemblies. We have studied the behavior of 241 Am-bearing MOX fuel in the fast reactor system and the effect on neutron and gamma-ray source-terms for safeguards measurements. In this paper, we will present the results of simulations of the behavior of 241 Am in a fast breeder reactor system. Because of the increased use of MOX fuel in thermal reactors and advances in fuel-cycle designs aimed at americium disposal in fast reactors, we have undertaken a brief study of the behavior of americium in these systems to better understand the safeguards impacts of these new approaches. In this paper we will examine the behavior of 241 Am in a variety of nuclear systems to provide insight into the safeguards implications of proposed Am disposition schemes.

Potential therapeutic gain from using p(66)/Be neutrons

International Nuclear Information System (INIS)

Slabbert, J.P.; Jones, D.T.L.; Theron, C.; Serafin, A.; Bohm, L.; Schmitt, G.

1997-01-01

Neutron therapy will be beneficial to patients with tumor types which are resistant to photons but relatively sensitive to high-LET radiation. In this work 15 different cell types, mostly of human tumor decent, were exposed in vitro to 60 Co γ-rays and p(66)/Be neutrons. Micronuclei frequencies in bi-nucleated cells and surviving fractions were determined for each cell type. Following exposure to either 1 or 1.5 Gy neutrons, micronuclei frequencies were significantly correlated with that observed from 2 Gy photons. A strong correlation between mean inactivation doses determined for these radiation modalities from survival curve inactivation parameters, was also noted. In spite of this a significant correlation between the variation in neutron RBE values and photon resistance was established. It is concluded that although neutron and photo sensitivities are related in the group of cell types studies, the use of this high energy neutron source may constitute a potential therapeutic gain for some tumor types. (authors)

Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

Science.gov (United States)

Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

2018-03-01

In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

Neutron sources and its dosimetric characteristics; Fuentes de neutrones y sus caracteristicas dosimetricas

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A. [Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico); Gallego D, E.; Lorente F, A. [Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)

2005-07-01

By means of Monte Carlo methods the spectra of the produced neutrons {sup 252} Cf, {sup 252} Cf/D{sub 2}O, {sup 241} Am Be, {sup 239} Pu Be, {sup 140} La Be, {sup 239} Pu{sup 18}O{sub 2} and {sup 226} Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H{sup *}(10), H{sub p,sIab} (10, 0{sup 0}), E{sub AP} and E{sub ISO}. During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of {sup 239} Pu Be and {sup 241} Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)

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

Neutron fluence spectrometry using disk activation

International Nuclear Information System (INIS)

Loevestam, Goeran; Hult, Mikael; Fessler, Andreas; Gasparro, Joel; Kockerols, Pierre; Okkinga, Klaas; Tagziria, Hamid; Vanhavere, Filip; Wieslander, J.S. Elisabeth

2009-01-01

A simple and robust detector for spectrometry of environmental neutrons has been developed. The technique is based on neutron activation of a series of different metal disks followed by low-level gamma-ray spectrometry of the activated disks and subsequent neutron spectrum unfolding. The technique is similar to foil activation but here the applied neutron fluence rates are much lower than usually in the case of foil activation. The detector has been tested in quasi mono-energetic neutron fields with fluence rates in the order of 1000-10000 cm -2 s -1 , where the obtained spectra showed good agreement with spectra measured using a Bonner sphere spectrometer. The detector has also been tested using an AmBe source and at a neutron fluence rate of about 40 cm -2 s -1 , again, a good agreement with the assumed spectrum was achieved

Neutron spectrometry with the NE-213 organic scintillator detector

International Nuclear Information System (INIS)

Silva, A.A. da.

1980-12-01

A neutron spectrometer with the NE-213 organic scintillator detector (5,08cm x 5,08cm) was mounted, tested, and calibrated at the Argonaut Reactor Laboratory of the Instituto de Engenharia Nuclear, to measure and study spectra of available fast neutron sources. The time zero-crossover technique was employed to discriminate the pulse of neutrons and gammas. The neutron spectrum from a 241 Am-Be source was determined experimentally in the range 1,0 MeV to 12,0 MeV and good agreement with other researchers was obtained. (Author) [pt

Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.

1981-08-01

An active neutron interrogation method has been developed for the measurement of 235 U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the 235 U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described

Study of calculated and measured time dependent delayed neutron yields

International Nuclear Information System (INIS)

Waldo, R.W.

1980-05-01

Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of 232 U, 237 Np, 238 Pu, 241 Am, /sup 242m/Am, 245 Cm, and 249 Cf were studied for the first time. The delayed neutron emission from 232 Th, 233 U, 235 U, 238 U, 239 Pu, 241 Pu, and 242 Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from 232 Th to 252 Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables

Review of Livermore-Led Neutron Capture Studies Using DANCE

International Nuclear Information System (INIS)

Parker, W; Sheets, S; Agvaanluvsan, U; Becker, J; Becvar, F; Bredeweg, T; Clement, R; Couture, A; Esch, E; Haight, R; Jandel, M; Krticka, M; Mitchell, G; Macri, R; O'Donnell, J; Reifarth, R; Rundberg, R; Schwantes, J; Ullmann, J; Vieira, D; Wouters, J; Wilk, P

2007-01-01

We have made neutron capture cross-section measurements using the white neutron source at the Los Alamos Science Center, the DANCE detector array (Detector for Advanced Neutron Capture Experiments) and targets important for basic science and stockpile stewardship. In this paper, we review results from (n,γ) reactions on 94,95 Mo, 152,154,157,160,nat Gd, 151,153 Eu and 242m Am for neutron energies from 94,95 Mo, we focused on the spin and parity assignments of the resonances and the determination of the photon strength functions for the compound nuclei 95,96 Mo. Future plans include measurements on actinide targets; our immediate interest is in 242m Am

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...

Determination of manganese in some pyrolusite ores of Myanmar by neutron activation analysis

International Nuclear Information System (INIS)

Myint, U.; Swe, M.T.

1994-01-01

Manganese in pyrolusite ores from various regions of Myanmar was determined by thermal neutron activation analysis using an Am(Be) neutron source. The induced activities of 56 Mn were monitored by a γ-counting technique. (author) 2 refs.; 1 tab

Determination of manganese in some pyrolusite ores of Myanmar by neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Myint, U; Swe, M T [Yangon Univ. (Myanmar). Dept. of Chemistry

1994-05-17

Manganese in pyrolusite ores from various regions of Myanmar was determined by thermal neutron activation analysis using an Am(Be) neutron source. The induced activities of [sup 56]Mn were monitored by a [gamma]-counting technique. (author) 2 refs.; 1 tab.

Determination of U-235 quantity in fresh fuel elements by neutron coincidence collar technique

International Nuclear Information System (INIS)

Almeida, M.C.M. de; Almeida, S.G. de; Marzo, M.A.S.; Moita, L.P.M.

1990-01-01

The U-235 quantity per lenght of fresh fuel assemblies of the Angra-I first recharge was determined by Neutron Coincidence Collar technique (N.C.C.). This technique is well-founded in fresh fuel assemblies activation by thermal neutrons from AmLi source to generate U-235 fission neutrons. These neutrons are detected by coincidence method in polyethylene structure where 18 He-3 detectors were placed. The coincidence counting results, in active mode (AmLi), showed 0,7% to standard deviation and equal to 1,49% to mass in 1000s of counting. The accuracies of different calibration methods were evaluated and compared. The results showed that the operator declared values are consistent. This evaluation was part of technical-exchange program between Safeguards Laboratory from C.N.E.N. and Los Alamos National Lab., United States. (author)

Neutron Detection with Large Plastic Scintillators for RPM Applications

International Nuclear Information System (INIS)

Corre, G.; Boudergui, K.; Sannie, G.; Kondrasovs, V.

2015-01-01

Homeland security requests the use Radiation Portal Monitor (RPM). They must be able to detect and differentiate gamma and neutron radiation. Gamma detection is required for illicit transportation of radioactive matter detection. Neutron detection is important to control nonproliferation of enriched material. Manufacturers worldwide propose sensors based on 3 He which give the actual state of art in term of neutron detection. The imminent shortage of 3 He forces manufacturers to find viable alternative. From 10 years sensors providers have the challenge to replace previous 3 He detectors that are known to be the most commonly deployed neutron sensor. As 3 He detectors can only detect neutron, they must be completed with gamma detector. The proposed approach is based on pulse time correlation between adjacent sensors from signal collected by EJ200 plastic scintillators. Results obtained during FP7 Scintilla project test campaigns show the system relevance for replacement of today's 3 He detectors. (authors)

Preliminary study of possible ORELA replacement options

International Nuclear Information System (INIS)

Olsen, D.K.; Martin, J.A.; Horen, D.J.

1984-06-01

Based on two conceptual design studies performed by the LANL Accelerator Technology Division, the possibilities in terms of accelerator systems for replacing ORELA with a more intense Maxwellian-type continuous-energy neutron source are summarized and discussed. The neutron intensities from ORELA are compared with those from existing or potential accelerator systems used for cross-section and condensed-matter studies. The best replacement options seem to involve a spallation source from 200- to 400-MeV protons on an ORELA-like target. Pulsing and intensity desiderata with such a source are discussed which correspond to a spectrum-averaged 100-fold improved figure of merit over ORELA for TOF measurements with only a tenfold increased source strength. Existing accelerator designs seem to be inadequate for such a source. Consequently, two conceptual designs were developed for this study by the LANL Accelerator Technology Division. The first conceptual design is for a 200-MeV large linac capable of accelerating 1.3 A during a macropulse; this linac standing alone could serve as an ORELA replacement source. The second conceptual design is for a much smaller 250-MeV PIGMI linac with a 28-mA macropulse current which feeds a proton accumulator ring and bunch-compressor transport line. This linac-ring-compressor (LIRIC) option would give a more cost-effective neutron source for cross-section measurements, whereas the large linac, or a modified version of it, would give a much simpler system more suitable for expansion. In particular, both conceptual designs would incorporate the present ORELA building and would provide approximately 100-fold improved neutron sources over ORELA for cross-section measurements. The total estimated cost of the LIRIC system would be $43M (1984), whereas the cost of the large linac would be about a factor of two more. 55 references, 11 figures, 19 tables

Neutron moisture gaging of agricultural soil

International Nuclear Information System (INIS)

Pospisil, S.; Janout, Z.; Kovacik, M.

1987-01-01

The design is described of a neutron moisture gage which consists of a measuring probe, neutron detector, small electronic recording device and a 241 Am-Be radionuclide source. The neutron detector consists of a surface barrier semiconductor silicon detector and a conversion layer of lithium fluoride. The detection of triton which is the reaction product of lithium with neutrons by the silicon detector is manifested as a voltage pulse. The detector has low sensitivity for fast neutrons and for gamma radiation and is suitable for determining moisture values in large volume samples. Verification and calibration measurements were carried out of chernozem, brown soil and podzolic soils in four series. The results are tabulated. Errors of measurement range between 0.8 to 1.0%. The precision of measurement could be improved by the calibration of the device for any type of soil. (E.S.). 4 tabs., 6 refs., 5 figs

Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C

International Nuclear Information System (INIS)

Suman, H.; Kharita, M. H.; Yousef, S.

2008-02-01

In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)

Measurement and analysis of the 241Am neutron capture cross section at the n_TOF facility at CERN

Science.gov (United States)

Mendoza, E.; Cano-Ott, D.; Altstadt, S.; Andriamonje, S.; Andrzejewski, J.; Audouin, L.; Balibrea, J.; Bécares, V.; Barbagallo, M.; Bečvář, F.; Belloni, F.; Berthier, B.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brugger, M.; Calviño, F.; Calviani, M.; Carrapiço, C.; Cerutti, F.; Chiaveri, E.; Chin, M.; Colonna, N.; Cortés, G.; Cortés-Giraldo, M. A.; Diakaki, M.; Dillmann, I.; Domingo-Pardo, C.; Durán, I.; Dzysiuk, N.; Eleftheriadis, C.; Ferrari, A.; Fraval, K.; Furman, V.; Gómez-Hornillos, M. B.; Ganesan, S.; García, A. R.; Giubrone, G.; Gonçalves, I. F.; González, E.; Goverdovski, A.; Gramegna, F.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Gurusamy, P.; Heftrich, T.; Heinitz, S.; Hernández-Prieto, A.; Heyse, J.; Jenkins, D. G.; Jericha, E.; Käppeler, F.; Kadi, Y.; Karadimos, D.; Katabuchi, T.; Ketlerov, V.; Khryachkov, V.; Koehler, P.; Kokkoris, M.; Kroll, J.; Krtička, M.; Lampoudis, C.; Langer, C.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Leong, L. S.; Lerendegui-Marco, J.; Licata, M.; López, D.; Losito, R.; Manousos, A.; Marganiec, J.; Martínez, T.; Massimi, C.; Mastinu, P.; Mastromarco, M.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Mondelaers, W.; Paradela, C.; Pavlik, A.; Perkowski, J.; Plompen, A. J. M.; Praena, J.; Quesada, J. M.; Rauscher, T.; Reifarth, R.; Riego-Perez, A.; Robles, M.; Roman, F.; Rubbia, C.; Ryan, J. A.; Sabaté-Gilarte, M.; Sarmento, R.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Sedyshev, P.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vermeulen, M. J.; Versaci, R.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Ware, T.; Weigand, M.; Weiss, C.; Wright, T.; Žugec, P.; n TOF Collaboration

2018-05-01

The 241Am(n ,γ ) cross section has been measured at the n_TOF facility at CERN with the n_TOF BaF2 Total Absorption Calorimeter in the energy range between 0.2 eV and 10 keV. Our results are analyzed as resolved resonances up to 700 eV, allowing a more detailed description of the cross section than in the current evaluations, which contain resolved resonances only up to 150-160 eV. The cross section in the unresolved resonance region is perfectly consistent with the predictions based on the average resonance parameters deduced from the resolved resonances, thus obtaining a consistent description of the cross section in the full neutron energy range under study. Below 20 eV, our results are in reasonable agreement with JEFF-3.2 as well as with the most recent direct measurements of the resonance integral, and differ up to 20-30% with other experimental data. Between 20 eV and 1 keV, the disagreement with other experimental data and evaluations gradually decreases, in general, with the neutron energy. Above 1 keV, we find compatible results with previously existing values.

Personnel neutron dosimetry using TLD elements at PNC

International Nuclear Information System (INIS)

Ishiguro, Hideharu

1985-01-01

The evaluation method of neutron dose equivalent was studied on the basis of the albedo type neutron dosimetory to design the personnel dosimeter. The dosimeter was composed of three 6 Li 2 10 B 4 O 7 (Cu) TL elements and one 7 Li 2 11 B 4 O 7 (Cu) element. The equations for assessing thermal, epithermal and fast neutron dose equivalents were derived by 252 Cf, 241 Am-Be and PuO 2 neutron sources. The minimum detectable amount of 6 Li 2 10 B 4 O 7 (Cu) element to thermal neutron was 0.02 m rem. The neutron dose equivalent and the gamma one were evaluated separately within about 20 % error in the mixed radiation field. (author)

Neutron fluence spectrometry using disk activation

Energy Technology Data Exchange (ETDEWEB)

Loevestam, Goeran [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium)], E-mail: goeran.loevestam@ec.europa.eu; Hult, Mikael; Fessler, Andreas; Gasparro, Joel; Kockerols, Pierre; Okkinga, Klaas [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Tagziria, Hamid [EC-JRC-Institute for the Protection and the Security of the Citizen (IPSC), Via E. Fermi 1, I-21020 Ispra (Vatican City State, Holy See,) (Italy); Vanhavere, Filip [SCK-CEN, Boeretang, 2400 Mol (Belgium); Wieslander, J.S. Elisabeth [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Department of Physics, P.O. Box 35 (YFL), FIN-40014, University of Jyvaeskylae (Finland)

2009-01-15

A simple and robust detector for spectrometry of environmental neutrons has been developed. The technique is based on neutron activation of a series of different metal disks followed by low-level gamma-ray spectrometry of the activated disks and subsequent neutron spectrum unfolding. The technique is similar to foil activation but here the applied neutron fluence rates are much lower than usually in the case of foil activation. The detector has been tested in quasi mono-energetic neutron fields with fluence rates in the order of 1000-10000 cm{sup -2} s{sup -1}, where the obtained spectra showed good agreement with spectra measured using a Bonner sphere spectrometer. The detector has also been tested using an AmBe source and at a neutron fluence rate of about 40 cm{sup -2} s{sup -1}, again, a good agreement with the assumed spectrum was achieved.

Neutron sources and its dosimetric characteristics

International Nuclear Information System (INIS)

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

2005-01-01

By means of Monte Carlo methods the spectra of the produced neutrons 252 Cf, 252 Cf/D 2 O, 241 Am Be, 239 Pu Be, 140 La Be, 239 Pu 18 O 2 and 226 Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H * (10), H p,sIab (10, 0 0 ), E AP and E ISO . During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of 239 Pu Be and 241 Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)

Determination of gold in some Myanmar indigenous medicines by neutron activation analysis

International Nuclear Information System (INIS)

Myint U.; Sein Sein Yi

1995-01-01

Gold has been determined in two Myanmar indigenous medicines TMF 14 (Devaauthada), TMF 15 (Shwe Thwe Say) by neutron activation analysis using an Am(Be) radionuclide neutron source. The activity of 411 keV of the 198 Au has been measured. (author). 2 refs., 1 fig., 1 tab

Replacement research reactor for Australia

International Nuclear Information System (INIS)

Miller, Ross

1998-01-01

In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

Instrumental neutron activation determination of gold in mineral raw materials using a californium neutron source

International Nuclear Information System (INIS)

Shilo, N.A.; Ippolitov, E.G.; Ivanenko, V.V.; Kustov, B.N.; Zheleznov, V.V.; Aristov, G.N.; Kovalenko, V.V.; Kondrat'ev, N.B.

1983-01-01

A facility using a californium neutron source and a method for the neutron activation analysis of gold were developed. The sensitivity of the determination is 0.1 g/t. The causes of random and systematic errors have been studied. It is concluded that in prospection and evaluation of gold ore deposists, the traditional test tube analysis for gold may be replaced with the developed method. (author)

How to replace a reactor pressure vessel

International Nuclear Information System (INIS)

Huber, R.

1996-01-01

A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

Energy response study of modified CR-39 neutron personnel dosimeter

International Nuclear Information System (INIS)

Sathian, Deepa; Bakshi, A.K.; Datta, D.; Nair, Sreejith S.; Sathian, V.; Mishra, Jitendra; Sen, Meghnath

2018-01-01

Personnel neutron dosimetry is an integral part of radiation protection. No single dosimeter provides the satisfactory energy response, sensitivity, angular dependence characteristics and accuracy necessary to meet the requirement of an ideal personnel neutron dosimeter. The response of a personnel neutron dosimeter is critically dependent upon the energy distribution of the neutron field. CR-39 personnel neutron dosimeters were typically calibrated in the standard neutron field of 252 Cf and 241 Am-Be in our laboratory, although actual neutron fields may vary from the calibration neutron spectrum. Recently the badge cassette of the personnel neutron dosimeter was changed due to frequent damage of the PVC badge used earlier. This paper discusses energy response of CR-39 solid state nuclear track detector loaded in this modified badge cassette as per latest ISO recommendation

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)

The behavior of moisture content in Durian after harvesting by neutron reflection and transmission techniques

International Nuclear Information System (INIS)

Chimoye, T.; Fuangfoong, M.

1998-01-01

The study aimed at development of a neutron reflection and transmission technique to determine moisture content in Durian fruit as a function of time after harvesting. A system of a 3 mCi Am-Be neutron source with a BF 3 detector as a neutron probe was developed. The results obtained were validated using weighting method

Evaluation of neutron shielding properties of lead glass using bubble detector

International Nuclear Information System (INIS)

Viswanathan, S.; Vishwa Prasad, K.; Srinivasan, T.K.; Ponraju, D.

1999-01-01

Neutron shielding properties of lead glass had been studied using a 241 Am-Be neutron source. Indigenously developed bubble detector was used as neutron detector. Attenuation curves were determined experimentally for the lead glass under the conditions of broad beam geometry. Theoretical calculations were made using Monte Carlo code MCNP3. Measurements were made for polyethylene and concrete to serve as reference. The measured and calculated neutron removal cross sections of lead glass, polyethylene and concrete are reported in this paper. Good agreement is observed between the experimental results and theoretical calculations. (author)

High-energy two-neutron removal from Be{sup 10}

Energy Technology Data Exchange (ETDEWEB)

Ashwood, N.I.; Freer, M.; Ahmed, S.; Clarke, N.M.; Curtis, N.; Soic, N.; Ziman, V.A. [Birmingham Univ., School of Physics and Astronomy, (United Kingdom); Millener, D.J. [Brookhaven National Lab., Upton, NY (United States); Orr, N.A.; Carstoiu, F.; Angelique, J.C.; Catford, W.N.; Lecouey, J.L.; Marques, F.M.; Normand, G.; Timis, C. [Caen Univ., Lab. de Physique Corpusculaire, ISMRA, IN2P3-CNRS, 14 (France); Carsoiu, F. [Horia Hulubei National institute of Physics and Nuclear Engineering (IFIN-HH), Bucharest-Magurele (Romania); Bouchat, V.; Hanappe, F.; Kerckx, Y.; Materna, T. [Universite Libre de Bruxelles (Belgium); Catford, W.N.; Pain, S.; Timis, C. [Surrey Univ., School of Electronics and Physical Sciences, Guildford (United Kingdom); Horoi, M. [Central Michigan Univ., Physics Dept., Mount Pleasant, MI (United States); Unshakova, A. [Joint Institute for Nuclear Research Dubna (Russian Federation)

2005-09-15

A kinetically complete measurement of the {sup 12}C({sup 10}Be, {alpha}+{alpha}+n) and ({sup 10}Be, {alpha}+{alpha}) reactions has been performed at a beam energy of 30 MeV/nucleon. The charged beam velocity particles were detected in an array of Si-CsI detectors placed at zero degrees, and the neutrons in an 81-element neutron array. The coincident detection of the final-state particles, produced in the breakup of {sup 10}Be, allowed the reconstruction of the excitation energy in the {sup 8}Be and {sup 9}Be systems. States in {sup 8}Be were identified, in particular the ground and first-excited states; and in {sup 9}Be, states at 1.68, 2.43, and (2.78, 3.05) MeV were observed. The population of these levels, in particular the 2.43 MeV 5/2- level, suggests that collective excitations play an important role in the neutron removal process. Distorted wave Born approximation and Glauber-type calculations have been used to model the direct neutron removal from the {sup 10}Be ground state and the two-step removal via inelastic excitations of the {sup 10}Be(2{sup +}) and {sup 9}Be(5/2{sup -}) excited states. (authors)

How should the JAERI neutron source be designed?

International Nuclear Information System (INIS)

Watanabe, Noboru

1996-01-01

The importance of a next-generation neutron source in JAERI is discussed. The feasibility and the performances of three types of neutron sources, namely continuous wave spallation source (CWSS), long-pulse spallation source (LPSS) and short-pulse spallation source (SPSS), are compared based on a proposed JAERI accelerator, a superconducting (SC) proton linac (1-1.5 GeV, 25-16 mA in peak current, finally CW). How to realize one of the world's best neutron source using such a linac with a modest beam-current and what type of neutron source is the best for such a linac are the most important current problems. Since the accelerator is not favorable for LPSS due to a lower peak current and there exist serious technical problems for a CWSS target, a short-pulse spallation source would be the best candidate to realize a 5 MW-class SPSS like ESS, provided that the H - -injection to a compressor ring over a long pulse duration (>2 ms) is feasible. (author)

Neutron spectrometry with organic scintillation detector

International Nuclear Information System (INIS)

Butragueno Casado, J. L.

1972-01-01

This work describes a fast neutron spectrometer using a stilbene crystal as head detector with pulse shape discrimination (P.S.D.) to reject gamma background. Tre experimental procedure involves the P.S.D., the measurements to calibrate the spectrometer and the corrections for several factors, mainly the non-linear response of the stilbene. Results of the measurements with the reaction D 2 (d,n)He 3 , and with an Am-Be neutron source are presented. It is also presented the measurement of the spectrum of the fast reactor CCRAl-1. (Author) 17 refs

Magnetic collapse of a neutron gas: Can magnetars indeed be formed?

International Nuclear Information System (INIS)

Martinez, A. Perez; Rojas, H. Perez; Cuesta, H.J.M.

2003-01-01

A relativistic degenerate neutron gas in equilibrium with a background of electrons and protons in a magnetic field exerts its pressure anisotropically, having a smaller value perpendicular to than along the magnetic field. For critical fields the magnetic pressure may produce the vanishing of the equatorial pressure of the neutron gas. Taking this as a model for neutron stars, the outcome could be a transverse collapse of the star. This fixes a limit to the fields to be observable in stable neutron star pulsars as a function of their density. The final structure left over after the implosion might be a mixed phase of nucleons and a meson condensate, a strange star, or a highly distorted black hole or black ''cigar'', but not a magnetar, if viewed as a superstrongly magnetized neutron star. However, we do not exclude the possibility of superstrong magnetic fields arising in supernova explosions which lead directly to strange stars. In other words, if any magnetars exist, they cannot be neutron stars. (orig.)

Radioactive source recovery program responses to neutron source emergencies

International Nuclear Information System (INIS)

Dinehart, S.M.; Hatler, V.A.; Gray, D.W.; Guillen, A.D.

1997-01-01

Recovery of neutron sources containing Pu 239 and Be is currently taking place at Los Alamos National Laboratory. The program was initiated in 1979 by the Department of Energy (DOE) to dismantle and recover sources owned primarily by universities and the Department of Defense. Since the inception of this program, Los Alamos has dismantled and recovered more than 1000 sources. The dismantlement and recovery process involves the removal of source cladding and the chemical separation of the source materials to eliminate neutron emissions. While this program continues for the disposal of 239 Pu/Be sources, there is currently no avenue for the disposition of any sources other than those containing Pu 239 . Increasingly, there have been demands from agencies both inside and outside the Federal Government and from the public to dispose of unwanted sources containing 238 Pu/Be and 241 Am/Be. DOE is attempting to establish a formal program to recover these sources and is working closely with the Nuclear Regulatory Commission (NRC) on a proposed Memorandum of Understanding to formalize an Acceptance Program. In the absence of a formal program to handle 238 Pu/Be and 241 Am/Be neutron sources, Los Alamos has responded to several emergency requests to receive and recover sources that have been determined to be a threat to public health and safety. This presentation will: (1) review the established 239 Pu neutron source recovery program at Los Alamos, (2) detail plans for a more extensive neutron source disposal program, and (3) focus on recent emergency responses

Response functions for computing absorbed dose to skeletal tissues from neutron irradiation

Science.gov (United States)

Bahadori, Amir A.; Johnson, Perry; Jokisch, Derek W.; Eckerman, Keith F.; Bolch, Wesley E.

2011-11-01

Spongiosa in the adult human skeleton consists of three tissues—active marrow (AM), inactive marrow (IM) and trabecularized mineral bone (TB). AM is considered to be the target tissue for assessment of both long-term leukemia risk and acute marrow toxicity following radiation exposure. The total shallow marrow (TM50), defined as all tissues lying within the first 50 µm of the bone surfaces, is considered to be the radiation target tissue of relevance for radiogenic bone cancer induction. For irradiation by sources external to the body, kerma to homogeneous spongiosa has been used as a surrogate for absorbed dose to both of these tissues, as direct dose calculations are not possible using computational phantoms with homogenized spongiosa. Recent micro-CT imaging of a 40 year old male cadaver has allowed for the accurate modeling of the fine microscopic structure of spongiosa in many regions of the adult skeleton (Hough et al 2011 Phys. Med. Biol. 56 2309-46). This microstructure, along with associated masses and tissue compositions, was used to compute specific absorbed fraction (SAF) values for protons originating in axial and appendicular bone sites (Jokisch et al 2011 Phys. Med. Biol. 56 6857-72). These proton SAFs, bone masses, tissue compositions and proton production cross sections, were subsequently used to construct neutron dose-response functions (DRFs) for both AM and TM50 targets in each bone of the reference adult male. Kerma conditions were assumed for other resultant charged particles. For comparison, AM, TM50 and spongiosa kerma coefficients were also calculated. At low incident neutron energies, AM kerma coefficients for neutrons correlate well with values of the AM DRF, while total marrow (TM) kerma coefficients correlate well with values of the TM50 DRF. At high incident neutron energies, all kerma coefficients and DRFs tend to converge as charged-particle equilibrium is established across the bone site. In the range of 10 eV to 100 Me

Neutron collar calibration for assay of LWR [light-water reactor] fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Pieper, J.E.

1987-03-01

The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the 235 U content, and the 238 U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities

Cerebral blood flow and metabolism during controlled hypotension with sodium-nitroprusside and general anaesthesia for total hip replacement a.m. Charnley

International Nuclear Information System (INIS)

Buenemann, L.; Jensen, K.; Thomsen, L.; Riisager, S.

1987-01-01

Cerebral blood flow (CBF) and cerebral metabolic rate of oxygen (CMRo 2 ) were studied during hypotension induced with sodium nitroprusside (SNP) in 10 patients undergoing total hip replacement a.m. Charnley. Cerebral blood flow was measured using an injection of xenon-133 into an arm vein. The decay curve was detected by five scintillation counters placed over each hemisphere and analysed with the Novo 10a cerebrograph. Blood samples were drawn from the radial artery and the jugular venous bulb to calculate the CMRo 2 . In the gropu as a whole, there were significant decreases in mean arterial pressure and in cerebrovascular resistance. There were no significant changes, in either CBF or CMRo 2 in the gropu as a whole, but there were substantial individual differences. In conclusion, the use of SNP-induced hypotension for extracranial surgery should be used only in patients monitored closely. (author)

Fluorescent converter and neutron absorber being made of boron nitride

International Nuclear Information System (INIS)

Matsumoto, G.; Teramura, M.; Sato, J.; Maeda, M.

1983-01-01

To improve the sensitivity of fluorescent converter is essential to the neutron radiography (NRG) which utilizes portable, not so strong, neutron sources. The fluorescent converter made of boron nitride (BN) is fabricated and tested. The sensitivity is about 1/20 of the NE426, but the homogeneity may be better. If 10 BN is utilized, the sensitivity will be five times as much as that of natural BN. Using the neutron beam of the Kyoto University Research Reactor, the flux of which is about 10 6 n/cm 2 sec, a good neutron television image was gained by X-ray television camera. As a bi-product of this converter, a flexible absorber was fabricated. (Auth.)

To the use of bubble detectors in personal neutron dosimetry

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Oxygen Chemical Diffusion Coefficients of (Pu,Am)O2 Fuels

International Nuclear Information System (INIS)

Watanabe, M.; Kato, M.; Matsumoto, T.

2015-01-01

Minor actinide (MA)-bearing MOX fuels have been developed as candidate fuels which are used in fast neutron spectrum cores such as sodium-cooled fast reactor (SFR) cores and experimental accelerator driven system (ADS) cores. Americium (Am) which is one of the MA elements significantly affects basic properties. It is known that Am content causes oxygen potential to increase and that influences irradiation behaviour such as fuel-cladding chemical interaction (FCCI) and chemical state of fission products. However, the effects of Am content on changes of basic properties are not clear. In this work, the oxygen chemical diffusion coefficients were calculated from measured data and the relationship between oxygen diffusion and oxygen potential of (Pu,Am)O 2-x was discussed. (authors)

Development of a D-D Neutron Generator

International Nuclear Information System (INIS)

Kim, In Jung; Jung, Hwa Dong; Park, Chang Su; Jung, Nam Suk; Jung, Soon Wook; Hwang, Y. S.; Choi, H. D.

2007-01-01

To enhance neutron yield, the ion source of the D-D neutron generator is replaced by a large current helicon plasma ion source. Current and energy of deuteron beam are increased, and hence neutron yield is enhanced. The maximum neutron yield is 2x10 8 n/s

Search for beta-delayed protons from {sup 11}Be

Energy Technology Data Exchange (ETDEWEB)

Forstner, Oliver [VERA Laboratory, University of Vienna (Austria); Stefan-Meyer-Institut, Austrian Academy of Sciences, Vienna (Austria); CERN, Geneva (Switzerland); Collaboration: IS541-Collaboration

2014-07-01

The one-neutron halo nucleus {sup 11}Be can emit a proton in a beta decay of the halo neutron. However, due to the Q-value of this decay channel (280.7±0.3 keV) the expected branching ratio will be very low - most estimates are a few times 10{sup -8} - and the detection of the outgoing proton with a kinetic energy of a few hundred keV is challenging. Therefore our attempt was to detect the remaining nucleus {sup 10}Be with the help of accelerator mass spectrometry (AMS). AMS is a highly sensitive tool to detect radioisotopes at the ultra-trace level. A beam of {sup 11}Be ions was produced at the ISOLDE facility at CERN and implanted in a collection sample. The sample was transferred to the VERA AMS facility at the University of Vienna where the {sup 10}Be content was determined. In my talk I present details of the experiment and results of the successful detection of this rare decay channel.

Preliminary results on bubble detector as personal neutron dosemeter

International Nuclear Information System (INIS)

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

2011-01-01

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

The neutron discovery

International Nuclear Information System (INIS)

Six, J.

1987-01-01

The neutron: who had first the idea, who discovered it, who established its main properties. To these apparently simple questions, multiple answers exist. The progressive discovery of the neutron is a marvellous illustration of some characteristics of the scientific research, where the unforeseen may be combined with the expected. This discovery is replaced in the context of the 1930's scientific effervescence that succeeded the revolutionary introduction of quantum mechanics. This book describes the works of Bothe, the Joliot-Curie and Chadwick which led to the neutron in an unexpected way. A historical analysis allows to give a new interpretation on the hypothesis suggested by the Joliot-Curie. Some texts of these days will help the reader to revive this fascinating story [fr

The determination of neutron energy spectra of radioisotope sources

International Nuclear Information System (INIS)

Lutkin, J.E.

1975-08-01

The neutron energy spectrum of a 241 Am-Be radioisotope neutron source has been determined by use of a time of flight neutron spectrometer; this spectrometer not being subject to the same uncertainties as a scintillation spectrometer. Neutron spectra have been determined using a scintillation spectrometer with which the effects of instrumental uncertainties, particularly the pulse shape discrimination have been assessed. In the course of the development of the time flight spectrometer a zero crossover pulse shape discrimination system was developed in order to reduce the unwanted background. Using this system a quantitative survey of pulse shape discrimination with experimental and commercial liquid and plastic organic scintillators were carried out. In addition the pulse shape discrimination properties of inorganic scintillators were also examined. (author)

Attenuation of fast neutron in concretes for biological shielding

International Nuclear Information System (INIS)

Labrada, A.; Chavez, A.; Gonzalez Mateu, D.; Desdin, F.; Tenjeiro, J.I.; Tellez, E.

1993-01-01

The attenuation of neutrons emitted by an 10 6 n/s. Am-Be source, in concretes elaborated with different aggregates is discussed in this paper. Two measurement methods were used an dosimetric system with Bonner spheres and 6 LiI(Eu) detector, and LAVSAN dielectric nuclear track detectors - with 238 U converts. The concretes elaborated with magnetite is reported as the best for neutron shielding while the Bauxite is not advisable for this purpose

The scientific and technical requirements for biology at Australia's Replacement Research Reactor

International Nuclear Information System (INIS)

2001-01-01

A Symposium and Workshop on Neutrons for Biology was held in the School of Biochemistry and Molecular Biology at the University of Melbourne, under the auspices of AINSE, Univ of Melbourne and ANSTO. Invited talks were given on the subjects of Genome, small-angle neutron scattering (SANS) as a critical framework for understanding bio-molecular, neutron diffraction at high and low resolution, and the investigation of viruses and large-scale biological structures using neutrons. There were also talks from prominent NMR practitioners and X-ray protein crystallographers, with substantial discussion about how the various methods might fit together in the future. Significant progress was made on defining Australia's needs, which include a strong push to use SANS and reflectometry for the study of macromolecular complexes and model membranes, and a modest network of supporting infrastructure in Brisbane, Melbourne and the Sydney Basin. Specific recommendations were that the small-angle neutron scattering and reflectometry instruments in the Replacement Research Reactor (RRR) be pursued with high priority, that there be no specific effort to provide high-resolution protein-crystallography facilities at the RRR, but that a watching brief be kept on instrumentation and sample-preparation technologies elsewhere. A watch be kept on inelastic and quasielastic neutron scattering capabilities elsewhere, although these methods will not initially be pursued at the RRR and that should be input from this community into the design of the biochemistry/chemistry laboratories at the Replacement Research Reactor. It was also recommended that a small number of regional facilities be established (or enhanced) to allow users to perform deuteration of biomolecules. These facilities would be of significant value to the NMR and neutron scattering communities

Current situation for exoelectron dosimeters of BeO ceramic in neutron dosimetry

International Nuclear Information System (INIS)

Gammage, R.B.

1977-01-01

Much of the early enthusiasm for using exoelectron dosimeters (ceramic BeO Thermalox 995) in neutron dosimetry was predicted on the belief that the response to fast neutrons, relative to gamma rays, was 0.18 to 0.28 on a R/sub γ/ equiv/tissue rad n/sub f/ basis for neutron energies between 0.1 and 16 MeV. Pairs of BeO disks had to be used, one covered with a polyethylene radiator for producing recoil protons, and the other covered with Teflon. More recent studies indicated a considerably lower ratio of 0.11 for Health Physics Reactor Research fission neutrons. In the earlier work the BeO was coated with gold to enhance the surface conductivity during reading of the thermally stimulated exoelectron emission (TSEE). No metallic coating is now deemed to be necessary. Perhaps thermal neutron contamination of the fast neutron beams due to some thermalization within the hydrogenous radiator was sufficient to cause the high apparent fast neutron sensitivity via n, γ reactions. Whatever the cause, however, the lower value of 0.11 has caused a marked subsidence of enthusiasm in this technique of fast neutron monitoring

A high output, large acceptance injector for the NOSAMS Tandetron AMS system

Energy Technology Data Exchange (ETDEWEB)

Longworth, Brett E., E-mail: blongworth@whoi.edu; Reden, Karl F. von; Long, Pat; Roberts, Mark L.

2015-10-15

We have completed a major upgrade of the National Ocean Sciences AMS Facility (NOSAMS) Tandetron AMS system in two stages. First, the simultaneous (recombinator) injector was replaced with a fast-cycling sequential injector and changes to the low-energy acceleration section. Data after the injector commissioning show an improvement in background, with mean machine background (commercial graphite) of Fm 0.0004 (62 ka). Second, we replaced the original ion source with a high-output 40 sample MCSNICS source. This improved beam currents and raw ratio fractionation, and increased sample to detection efficiency fivefold.

Calibration and evaluation of neutron survey meters used at linac facility

Energy Technology Data Exchange (ETDEWEB)

Salgado, A.P. [Instituto de Radioprotecao e Dosimetria - IRD, Av. Salvador Allende s/n, Recreio dos Bandeirantes, CEP 22780-160 Rio de Janeiro (Brazil); Pereira, W.W., E-mail: walsan@ird.gov.b [Instituto de Radioprotecao e Dosimetria - IRD, Av. Salvador Allende s/n, Recreio dos Bandeirantes, CEP 22780-160 Rio de Janeiro (Brazil); Fonseca, E.S. da; Patrao, K.C.S. [Instituto de Radioprotecao e Dosimetria - IRD, Av. Salvador Allende s/n, Recreio dos Bandeirantes, CEP 22780-160 Rio de Janeiro (Brazil); Batista, D.V.S. [Instituto Nacional do Cancer - INCa, Praca Cruz Vermelha, 23 - centro, CEP 20230-130 Rio de Janeiro (Brazil)

2010-12-15

Calibrated survey meters from the Neutron Laboratory of the Instituto de Radioprotecao e Dosimetria (IRD) were used to determine the ambient dose-equivalent rate in a 15 MV linear accelerator treatment room at the Instituto Nacional do Cancer (INCa). Three different models of neutron survey meters were calibrated using four neutron radionuclide neutron sources: {sup 241}AmBe({alpha},n), {sup 252}Cf(f,n), heavy-water moderated {sup 252}Cf(f,n), and {sup 238}PuBe({alpha},n). All neutron sources were standardized in a Manganese Sulphate Bath (MSB) absolute primary system. The response of each of these instruments was compared with reference values of ambient dose-equivalent rate. The results demonstrate the complexity of making measurements in the mixed neutron/photon field produced in electron linear accelerator radiotherapy treatment rooms.

Effect of double false pulses in calibrated neutron coincidence collar during measuring time-correlated neutrons from PuBe neutron sources

Energy Technology Data Exchange (ETDEWEB)

Nguyen, Tam Cong, E-mail: tam.nguyen.cong@energia.mta.hu; Huszti, Jozsef; Nguyen, Quan Van

2015-09-01

Effect of double false pulses of preamplifiers in neutron coincidence collar was investigated to explain non-parallel shape of calibrated D/S–M{sub Pu} curves of two commercial neutron coincidence collars, JCC-31 and JCC-13. Two curves, which were constructed from D/S ratio (doubles and singles count rate), and Pu content M{sub Pu}, of the same set of secondary standard PuBe neutron sources, should be parallel. Non-parallelism rises doubt about usability of the method based on this curve for determination of Pu content in PuBe neutron sources. We have shown in three steps that the problem originates from double false pulses of preamplifiers in JCC-13. First we used a pulse train diagram for analyzing the non-parallel shape, second we used Rossi-Alpha distribution measured by pulse train recorder developed in our institute and finally, we investigated the effect of inserted noise pulses. This implies a new type of QA test option in traditional multiplicity shift registers for excluding presence of double false pulses.

Use of cellulose nitrate plastic track detectors in neutron personnel monitoring

International Nuclear Information System (INIS)

Venkataraman, G.; Marathe, P.K.; Joshi, R.V.

1975-01-01

Cellulose nitrate, which is a sensitive plastic material wherein even proton tracks could be recorded, was studied with a view to using it for personnel neutron monitoring work. It was found that among the commercially available plastics, the colourless transparent Daicel 6000 variety having a thickness of 0.6mm is satisfactory form the point of view of track recognition. The material was exposed to thermal neutrons, fission neutrons, neutrons from an Am-Be source and to 14 MeV neutrons. As is to be expected, there is no thermal neutron sensitivity. The sensitivity to fast neutrons is less by a factor of two as compared to that of the NTA nuclear track emulsion. It was observed that the background tracks found in the plastic correspond to nearly 100 mrem of fast neutron dose equivalent. It is felt that at present cellulose nitrate is useful for monitoring personnel involved in nuclear criticality accidents. (author)

Neutron scattering science in Australia

International Nuclear Information System (INIS)

Knott, Robert

1999-01-01

Neutron scattering science in Australia is making an impact on a number of fields in the scientific and industrial research communities. The unique properties of the neutron are being used to investigate problems in chemistry, materials science, physics, engineering and biology. The reactor HIFAR at the Australian Nuclear Science and Technology Organisation research laboratories is the only neutron source in Australia suitable for neutron scattering science. A suite of instruments provides a wide range of opportunities for the neutron scattering community that extends throughout universities, government and industrial research laboratories. Plans are in progress to replace the present research reactor with a modern multi-purpose research reactor to offer the most advanced neutron scattering facilities. The experimental and analysis equipment associated with a modern research reactor will permit the establishment of a national centre for world class neutron science research focussed on the structure and functioning of materials, industrial irradiations and analyses in support of Australian manufacturing, minerals, petrochemical, pharmaceuticals and information science industries. (author)

Neutron scattering science in Australia

Energy Technology Data Exchange (ETDEWEB)

Knott, Robert [Australian Nuclear Science and Technology Organisation, Menai, NSW (Australia)

1999-10-01

Neutron scattering science in Australia is making an impact on a number of fields in the scientific and industrial research communities. The unique properties of the neutron are being used to investigate problems in chemistry, materials science, physics, engineering and biology. The reactor HIFAR at the Australian Nuclear Science and Technology Organisation research laboratories is the only neutron source in Australia suitable for neutron scattering science. A suite of instruments provides a wide range of opportunities for the neutron scattering community that extends throughout universities, government and industrial research laboratories. Plans are in progress to replace the present research reactor with a modern multi-purpose research reactor to offer the most advanced neutron scattering facilities. The experimental and analysis equipment associated with a modern research reactor will permit the establishment of a national centre for world class neutron science research focussed on the structure and functioning of materials, industrial irradiations and analyses in support of Australian manufacturing, minerals, petrochemical, pharmaceuticals and information science industries. (author)

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)

A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E. [National Research Nuclear University MEPhI (Russian Federation)

2016-12-15

An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

Estimation of low-level neutron dose-equivalent rate by using extrapolation method for a curie level Am–Be neutron source

International Nuclear Information System (INIS)

Li, Gang; Xu, Jiayun; Zhang, Jie

2015-01-01

Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am–Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am–Be neutron source was investigated by measuring the count rates obtained through a 3 He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, 3 He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter. - Highlights: • The scope of the affected area for

Decay counting in the age of AMS

International Nuclear Information System (INIS)

Arnold, J.R.

1987-01-01

Up to the advent of AMS, all the accomplishments of 14 C dating, and the studies of 10 Be and other long-lived radionuclides, were made by low-level decay counting, the technique pioneered by W.F. Libby. It will hardly be news to people at this conference that, while much was accomplished in the three decades when counting prevailed, the world has now changed decisively. I will try to give an account of where low-level counting was 'before the revolution', and of what its usefulness is today. There are still some remarkable examples of its application, the best being the neutrino experiment of Raymond Davis, and its potential successors. Some cosmogenic nuclides, whose half-lives are less than 10 3 yr, are still best measured by decay; this will continue unless the overall ion yield of AMS systems rises markedly from present levels. One long-lived nuclide, 53 Mn, is still best measured by neutron activation as 312-day 54 Mn, but this may not continue. (orig.)

Decay counting in the age of AMS

International Nuclear Information System (INIS)

Arnold, J.R.

1987-01-01

Up to the advent of AMS, all the accomplishments of 14 C dating, and the studies of 10 Be and other long-lived radionuclides, were made by low-level decay counting, the technique pioneered by W.F. Libby. It will hardly be news to people at this conference that, while much was accomplished in the three decades when counting prevailed, the world has now changed decisively. He will try to give an account of where low-level counting was before the revolution, and of what its usefulness is today. There are still some remarkable examples of its application, the best being the neutrino experiment of Raymond Davis, and its potential successors. Some cosmogenic nuclides, whose half-lives are less than 10 3 yr, are still best measured by decay; this will continue unless the overall ion yield of AMS systems rises markedly from present levels. One long-lived nuclide, 53 Mn, is still best measured by neutron activation as 312-day 54 Mn, but this may not continue

When thin is sexy - neutron reflectometry instrumentation at the Australian replacement research reactor

International Nuclear Information System (INIS)

James, M.

2003-01-01

Full text: Neutron and X-ray reflectometry are techniques used to probe the structure of surfaces, thin-films or buried interfaces as well as processes occurring at surfaces and interfaces such as adsorption, adhesion and inter-diffusion. Applications cover adsorbed surfactant layers, self-assembled monolayers, biological membranes, electrochemical and catalytic interfaces, polymer coatings and photosensitive films. Contrast variation and selective deuteration of hydrogenous materials are important aspects of the neutron-based technique. Neutron reflectometry probes the structure of materials normal to the surface at depths of up to several thousand Angstroms, with an effective depth resolution of a few Angstroms. Neutron reflectometry experiments have been performed by a number of Australian researchers at overseas facilities for more than a decade, however this capability has previously been absent in this country. A neutron reflectometer has been recognised as one of the highest priority instruments to be constructed at the new 20MW research reactor facility at Lucas Heights (due for completion in 2006). In this presentation we report the design of the time-of-flight reflectometer to be constructed at the new research facility. The reflectometer will operate with a vertical scattering plane and thus be able to measure specular reflectometry from both solid and liquid samples. A series of disc choppers will enable the instrument resolution (ΔQ/Q) to be varied from 2% to > 15%. The detection system will consist of a 2-dimenional position sensitive detector that will also allow the measurement of off-specular reflectivity

MCNP SIMULATION OF THE HP(10) ENERGY RESPONSE OF A BRAZILIAN TLD ALBEDO NEUTRON INDIVIDUAL DOSEMETER, FROM THERMAL TO 20 MeV.

Science.gov (United States)

Freitas, B M; Martins, M M; Pereira, W W; da Silva, A X; Mauricio, C L P

2016-09-01

The Brazilian Instituto de Radioproteção e Dosimetria (IRD) runs a neutron individual monitoring system with a home-made TLD albedo dosemeter. It has already been characterised and calibrated in some reference fields. However, the complete energy response of this dosemeter is not known, and the calibration factors for all monitored workplace neutron fields are difficult to be obtained experimentally. Therefore, to overcome such difficulties, Monte Carlo simulations have been used. This paper describes the simulation of the HP(10) neutron response of the IRD TLD albedo dosemeter using the MCNPX transport code, for energies from thermal to 20 MeV. The validation of the MCNPX modelling is done comparing the simulated results with the experimental measurements for ISO standard neutron fields of (241)Am-Be, (252)Cf, (241)Am-B and (252)Cf(D2O) and also for (241)Am-Be source moderated with paraffin and silicone. Bare (252)Cf are used for normalisation. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Portable gamma and thermal neutron probe using a 6LiI(Eu) crystal

International Nuclear Information System (INIS)

Carneiro, Clemente J.G.; Araujo, Geraldo P.; Milian, Felix M.; Barbosa, Jurandir C.; Garcia, Fermin; Oliveira, Arno H.; Silva, Mario R.S.; Penna, Rodrigo

2011-01-01

Europium-activated lithium-6 iodide is a scintillator used for gamma and neutron counting. A portable detection system was built based on this scintillator. This system has three modules: the scintillator, a 10 m liquid light guide, and a Hamamatsu photon counting head H9319 used as a light pulse digitizer. Data transfer, measurement time and other necessary adjustment can be controlled by software from the PC through the RS-232C interface. The scintillator, a crystal of 6 LiI(Eu), is a small cylinder with 3 mm diameter and 40 mm length completely sealed in an aluminum tube coupled to the light guide. The small size of the scintillator increases the neutron/gamma count ratio, since 2 to 3 mm of thickness of this crystal absorbs all thermal neutrons. Intensities of X and gamma rays, and thermal neutrons can be recorded for time intervals of 10 ms to 1 s storing up to 10000 countings. The system was calibrated for measuring radiation doses for validating numerical models in dosimetry. Two characteristic reinforce this application, measurements can be done at several meters away from the radiation source and also inside of water. In addition, it was used to build nuclear probes based on Compton scattering or neutron moderation in porous media by attaching an AmBe source to the top of the aluminum tube. Tests were done to determine the reproducibility of counting rates. Background counting was measured at several temperatures to verify the influence of dark current of PMT. Sealed AmBe, low activity Am, and X rays sources were used for studies of radiation counting statistics. X rays apparatus was used to correlate counting rates measured with the 6 LiI(Eu) detection system and doses measured with an ionization chamber at several distances from the X ray source. (author)

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

Fast neutron and gamma-ray transmission technique in mixed samples. MCNP calculations

International Nuclear Information System (INIS)

Perez, N.; Padron, I.

2001-01-01

In this paper the moisture in sand and also the sulfur content in toluene have been described by using the simultaneous fast neutron/gamma transmission technique (FNGT). Monte Carlo calculations show that it is possible to apply this technique with accelerator-based and isotopic neutron sources in the on-line analysis to perform the product quality control, specifically in the building materials industry and the petroleum one. It has been used particles from a 14MeV neutron generator and also from an Am-Be neutron source. The estimation of optimal system parameters like the efficiency, detection time, hazards and costs were performed in order to compare both neutron sources

Determination of V, W and Mn in high-speed steel by neutron activation source of 241Am/Be

International Nuclear Information System (INIS)

Villar, H.P.; Galdino, S.M.L.; Godoy, M.O.; Dantas, C.C.

1982-01-01

Alloying elements are responsible for certain characteristics of the steels which enable their utilization for specific purposes. The concentrations of these elements must comply with strict standards, and the determination of these concentrations involve chemical analyses which are as a rule tedious and expensive. It is proposed here a fast and precise analytical process based on the neutron activation analysis. A significant correlation (r = 0.998) between manganese concentration and mean specific count rate of 56 Mn was obtained for activated tool steel samples. Later on, bases for tungsten vanadium determinations were set. (Author) [pt

Determination of V, W and Mn in fast steels by neutron activation analysis of 241Am/Be source

International Nuclear Information System (INIS)

Villar, H.P.; Galdino, S.M.L.; Godoy, M.O.; Dantas, C.C.

1982-01-01

Alloying elements are responsible for certain characteristics of the steels which enable their utilization for specific purposes. The concentrations of these elements must comply with strict standards, and the determination of these concentrations involve chemical analyses which are as a rule tedious and expensive. It is proposed here a fast and precise analytical process based on the neutron activation analysis. A significant correlation (r = 0.998) between manganese concentration and mean specific count rate of 56 Mn was obtained for activated tool steel sampes. Later on, bases for tungsten vanadium determinations were set. (Author) [pt

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)

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.

Slow neutron mapping technique for level interface measurement

Science.gov (United States)

Zain, R. M.; Ithnin, H.; Razali, A. M.; Yusof, N. H. M.; Mustapha, I.; Yahya, R.; Othman, N.; Rahman, M. F. A.

2017-01-01

Modern industrial plant operations often require accurate level measurement of process liquids in production and storage vessels. A variety of advanced level indicators are commercially available to meet the demand, but these may not suit specific need of situations. The neutron backscatter technique is exceptionally useful for occasional and routine determination, particularly in situations such as pressure vessel with wall thickness up to 10 cm, toxic and corrosive chemical in sealed containers, liquid petroleum gas storage vessels. In level measurement, high energy neutrons from 241Am-Be radioactive source are beamed onto a vessel. Fast neutrons are slowed down mostly by collision with hydrogen atoms of material inside the vessel. Parts of thermal neutron are bounced back towards the source. By placing a thermal detector next to the source, these backscatter neutrons can be measured. The number of backscattered neutrons is directly proportional to the concentration of the hydrogen atoms in front of the neutron detector. As the source and detector moved by the matrix around the side of the vessel, interfaces can be determined as long as it involves a change in hydrogen atom concentration. This paper presents the slow neutron mapping technique to indicate level interface of a test vessel.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Measurement of epithermal neutrons by a coherent demodulation technique

CERN Document Server

Horiuchi, N; Takahashi, H; Kobayashi, H; Harasawa, S

2000-01-01

Epithermal neutrons have been measured using a neutron dosimeter via a coherent demodulation technique. This dosimeter consists of CsI(Tl)-photodiode scintillation detectors, four of which are coupled to neutron-gamma converting foils of various sizes. Neutron-gamma converting foils of In, Au and Co materials were used, each of which has a large capture cross section which peaks in the epithermal neutron energy region. The type of foil was selected according to the material properties that best correspond to the energy of the epithermal neutrons to be measured. In addition, the proposed technique was applied using Au-foils in order to measure the Cd ratio. The validity of the proposed technique was examined using an sup 2 sup 4 sup 1 Am-Be source placed in a testing stack of polyethylene blocks, and the results were compared with the theoretical values calculated by the Monte Carlo calculation. Finally, the dosimeter was applied for measuring epithermal neutrons and the Cd ratio in an experimental beam-tube o...

Identification of new neutron-rich actinide isotopes

Energy Technology Data Exchange (ETDEWEB)

Oura, Yasuji; Sakama, Minoru; Ohyama, T. [Tokyo Metropolitan Univ. (Japan)] [and others

1999-10-01

To advance research on new neutron-deficient actinide isotopes using an on-line isotope separator combined with a gas-jet injector installed in the JAERI Tandem accelerator, Tokai, performance test of the equipment was carried out. Efficiency of the product isotopes being transported from the target chamber to the measuring system was greatly improved by employing lead iodides (PbI{sub 2}) as the aerosol carrier. With the help of this technique, the authors succeeded in synthesizing and identifying actinide isotopes, {sup 235}Am and {sup 236}Am, and measured their alpha-decay half-life. (S. Ohno)

AMS prepares for long stay in space

CERN Multimedia

CERN Bulletin

2010-01-01

Following the successful space qualification tests at the ESA Technology Centre (ESTEC) in Noordwijk in the Netherlands, AMS is now back in the integration hall at CERN Prévessin. The collaboration is replacing the superconducting magnet with a permanent (non-superconducting) one, which will ensure reliable operation of the experiment for the recently planned longer run on board the International Space Station (ISS).   Work is under way at the AMS integration hall at CERN Prévessin. Following a trip to ESTEC in Noordwijk in the Netherlands, where tests confirmed its fitness for launch into space on board the International Space Station (ISS), the AMS experiment is now back at CERN for final modifications. “The collaboration agreed to adopt a modified configuration that, among other things, re-uses the permanent magnet of the AMS-01 prototype that was flown into space in 1998”, says Samuel Ting, Spokesperson of the AMS experiment. Althoug...

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)

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

International Nuclear Information System (INIS)

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

The development of an automatic scanning method for CR-39 neutron dosimeter

International Nuclear Information System (INIS)

Tawara, Hiroko; Miyajima, Mitsuhiro; Sasaki, Shin-ichi; Hozumi, Ken-ichi

1989-01-01

A method of measuring low level neutron dose has been developed with CR-39 track detectors using an automatic scanning system. It is composed of the optical microscope with a video camera, an image processor and a personal computer. The focus point of the microscope and the X-Y stage are controlled from the computer. The minimum detectable neutron dose is estimated at 4.6 mrem in the uniform field of neutron with equivalent energy spectrum to Am-Be source from the results of automatic measurements. (author)

Neutron shielding properties of a borated high-density glass

Directory of Open Access Journals (Sweden)

Saeed Aly Abdallah

2017-01-01

Full Text Available The neutron shielding properties of a borated high density glass system was characterized experimentally. The total removal macroscopic cross-section of fast neutrons, slow neutrons as well as the linear attenuation coefficient of total gamma rays, primary in addition to secondary, were measured experimentally under good geometric condition to characterize the attenuation properties of (75-x B2O3-1Li2O-5MgO-5ZnO-14Na2O-xBaO glassy system. Slabs of different thicknesses from the investigated glass system were exposed to a collimated beam of neutrons emitted from 252Cf and 241Am-Be neutron sources in order to measure the attenuation properties of fast and slow neutrons as well as total gamma rays. Results confirmed that barium borate glass was suitable for practical use in the field of radiation shielding.

Measurement of the 243Am capture cross section at the n{sub T}OF facility; Medida de la sección eficaz de captura del 243Am en la instalación n{sub T}OF

Energy Technology Data Exchange (ETDEWEB)

Mendoza Cembranos, E.

2014-07-01

Nuclear data for minor actinides are necessary for improving the design and performance of advanced reactors and transmutation devices for the incineration of radioactive nuclear waste [Sal08, Gon09, Ali04, Ali06]. In particular, the 243Am isotope is relevant since it is the minor actinide which contributes more to the radiotoxicity of the nuclear waste between s3 03 and s3 04 years. In addition, the neutron capture in 243Am is the main gate to the creation of 244Cm and higher mass isotopes. The purpose of the this work is to provide experimental data on the 243Am(n, ) for improving the current evaluations. At present, there is no published neutron capture measurement of 243Am below 250 eV, and all the existing evaluations of the elastic and capture cross sections are based essentially on a single transmission measurement [Sim74]. Above 250 eV there are only a few capture measurements available [Wes85, Wis83], which show discrepancies that make them incompatible. Due to the lack of experimental data on 243Am the standard ENDF-6 format libraries present sizeable di rences between each other...(Author)

Portable gamma and thermal neutron probe using a {sup 6}LiI(Eu) crystal

Energy Technology Data Exchange (ETDEWEB)

Carneiro, Clemente J.G.; Araujo, Geraldo P.; Milian, Felix M.; Barbosa, Jurandir C.; Garcia, Fermin [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Centro de Pesquisas em Ciencias e Tecnologias das Radiacoes (CPqCTR); Oliveira, Arno H.; Silva, Mario R.S.; Penna, Rodrigo [Universidade Federal de Minas Gerais (DEN-UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2011-07-01

Europium-activated lithium-6 iodide is a scintillator used for gamma and neutron counting. A portable detection system was built based on this scintillator. This system has three modules: the scintillator, a 10 m liquid light guide, and a Hamamatsu photon counting head H9319 used as a light pulse digitizer. Data transfer, measurement time and other necessary adjustment can be controlled by software from the PC through the RS-232C interface. The scintillator, a crystal of {sup 6}LiI(Eu), is a small cylinder with 3 mm diameter and 40 mm length completely sealed in an aluminum tube coupled to the light guide. The small size of the scintillator increases the neutron/gamma count ratio, since 2 to 3 mm of thickness of this crystal absorbs all thermal neutrons. Intensities of X and gamma rays, and thermal neutrons can be recorded for time intervals of 10 ms to 1 s storing up to 10000 countings. The system was calibrated for measuring radiation doses for validating numerical models in dosimetry. Two characteristic reinforce this application, measurements can be done at several meters away from the radiation source and also inside of water. In addition, it was used to build nuclear probes based on Compton scattering or neutron moderation in porous media by attaching an AmBe source to the top of the aluminum tube. Tests were done to determine the reproducibility of counting rates. Background counting was measured at several temperatures to verify the influence of dark current of PMT. Sealed AmBe, low activity Am, and X rays sources were used for studies of radiation counting statistics. X rays apparatus was used to correlate counting rates measured with the {sup 6}LiI(Eu) detection system and doses measured with an ionization chamber at several distances from the X ray source. (author)

Thermoresponsive behaviour of AM{sub 2}O{sub 8} materials

Energy Technology Data Exchange (ETDEWEB)

Allen, Simon

2003-07-01

This thesis investigates the synthesis and structural characterisation of AM{sub 2}O{sub 8} phases, many of which show negative thermal expansion (NTE); relevant literature is reviewed in Chapter One. Chapter Two describes the synthesis, structure solution, and mechanistic role of a new family of low-temperature (LT) orthorhombic AM{sub 2}O{sub 8} polymorphs (A{sup IV} = Zr, Hf; M{sup VI} = Mo, W). These materials are key intermediates in the preparation of cubic AM{sub 2}O{sub 8} phases from AM{sub 2}O{sub 7}(OH){sub 2}(H{sub 2}O){sub 2}. The structure of LT-AM{sub 2}O{sub 8} has been elucidated by combined laboratory X-ray and neutron powder diffraction. Variable temperature X-ray diffraction (VTXRD) studies have shown LT-AMo{sub 2}O{sub 8} phases exhibit anisotropic NTE. LT-ZrMo{sub 2}O{sub 8} has been shown to undergo spontaneous rehydration, allowing preparation of ZrMo{sub 2}O{sub 7}(OD){sub 2}(D{sub 2}O){sub 2} and assignment of D{sub 2}O/OD positions within the structure by neutron diffraction. Using this result, a reversible topotactic dehydration pathway from AM{sub 2}O{sub 7}(OH){sub 2}(H{sub 2}O){sub 2} to LT-AM{sub 2}O{sub 8} is proposed. Chapter Three investigates the order-disorder phase transition with concurrent oxygen mobility in cubic AM{sub 2}O{sub 8} materials; studies include comprehensive VT neutron diffraction of cubic ZrMo{sub 2}O{sub 8} to reveal a static to dynamic transition at 215 K, and novel quench-anneal/quench-warm variable temperature/time diffraction experiments on ZrWMoO{sub 8} which lead to an activation energy of 40 kJmol{sup -1} for oxygen migration. In Chapter Four {sup 17}O-labelled cubic ZrW{sub 2}O{sub 8} has been prepared to understand the oxygen migration process by VT MAS NMR. In situ hydrothermal studies of cubic ZrMo{sub 2}O{sub 8} using synchrotron radiation have shown direct hydration to ZrMo{sub 2}O{sub 7}(OH){sub 2}(H{sub 2}O){sub 2}. In Chapter Five VTXRD of trigonal {alpha}-AMo{sub 2}O{sub 8} phases reveals a

Stable isotope applications of AMS in geology

International Nuclear Information System (INIS)

Rucklidge, J.C.

1981-01-01

The subject of geochemistry has become increasingly concerned with the distribution of trace elements in and between mineral phases. Part per million detection is routine, but part per billion measurements are, for certain elements, beyond the range of such sensitive analytical methods as neutron activation analysis (NAA). Tandem AMS has the ability to extend this limit several orders of magnitude for those elements which readily form negative ions. There is no doubt that such information can be most valuable for elements which are partitioned strongly between different mineral phases. While bulk analyses may indicate trace levels of certain elements to be present in a rock, it has always been difficult to state with certainty whether the trace element occurs at a uniformly low level throughout the various phases, or whether it is concentrated at a high level in small grains of an extremely rare phase scattered through the rock. The milli- or micro-probe analytical capability, which can be part of AMS, enables such problems concerning ultra-low level element concentrations to be tackled. With the same approach isotopic ratios of both major and minor elements in microgram amounts of material may be undertaken

Neutron multiplier alternative for fusion reactor blankets

International Nuclear Information System (INIS)

Taczanowski, S.

1980-01-01

A proposal is given to replace neutron multiplier needed to enable low lithium and tritium inventories simultaneously assuring sufficient production of tritium, by an efficient moderator ( 7 LiH or 7 LiD). The advantageous effect of the intensified neutron energy degradation is due to the 1/v character of the main tritium producing reaction. The slowing-down medium is designed to be the source of moderated neutrons for the surrounding Li ( 6 Li enriched) region where the most of tritium is to be produced. The surplus tritium production remains stored in the moderator zone. Some preliminary calculations illustrating the above concept were carried out and the neutron flux and tritium production distributions are presented. The indications regarding further studies are also suggested. (author)

Neutron spectrometry and dosimetry with ANNs

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Gallego, E.; Lorente, A.

2009-10-01

Artificial neural networks technology has been applied to unfold the neutron spectra and to calculate the effective dose, the ambient equivalent dose, and the personal dose equivalent for 252 Cf and 241 AmBe neutron sources. A Bonner sphere spectrometry with a 6 LiI(Eu) scintillator was utilized to measure the count rates of the spheres that were utilized as input in two artificial neural networks, one for spectrometry and another for dosimetry. Spectra and the ambient dose equivalent were also obtained with BUNKIUT code and the UTA4 response matrix. With both procedures spectra and ambient dose equivalent agrees in less than 10%. (author)

Neutron scattering science at the Australian Nuclear Science and Technology Organisation (ANSTO)

International Nuclear Information System (INIS)

Knott, Robert

2000-01-01

Neutron scattering science at ANSTO is integrated into a number of fields in the Australian scientific and industrial research communities. The unique properties of the neutron are being used to investigate problems in chemistry, materials science, physics, engineering and biology. The reactor HIFAR at the Australian Nuclear Science and Technology Organisation research laboratories is the only neutron source in Australia suitable for neutron scattering science. A suite of instruments provides a range of opportunities for the neutron scattering community that extends throughout universities, government and industrial research laboratories. Plans to replace the present research reactor with a modern multi-purpose research reactor are well advanced. The experimental and analysis equipment associated with a modern research reactor will permit the establishment of a national centre for world class neutron science research focussed on the structure and functioning of materials, industrial irradiations and analyses in support of Australian manufacturing, minerals, petrochemical, pharmaceuticals and information science industries. A brief overview will be presented of all the instruments presently available at ANSTO with emphasis on the SANS instrument. This will be followed by a description of the replacement research reactor and its instruments. (author)

Neutron scattering science at the Australian Nuclear Science and Technology Organisation (ANSTO)

Energy Technology Data Exchange (ETDEWEB)

Knott, Robert [Australian Nuclear Science and Technology Organisation (Australia)

2000-10-01

Neutron scattering science at ANSTO is integrated into a number of fields in the Australian scientific and industrial research communities. The unique properties of the neutron are being used to investigate problems in chemistry, materials science, physics, engineering and biology. The reactor HIFAR at the Australian Nuclear Science and Technology Organisation research laboratories is the only neutron source in Australia suitable for neutron scattering science. A suite of instruments provides a range of opportunities for the neutron scattering community that extends throughout universities, government and industrial research laboratories. Plans to replace the present research reactor with a modern multi-purpose research reactor are well advanced. The experimental and analysis equipment associated with a modern research reactor will permit the establishment of a national centre for world class neutron science research focussed on the structure and functioning of materials, industrial irradiations and analyses in support of Australian manufacturing, minerals, petrochemical, pharmaceuticals and information science industries. A brief overview will be presented of all the instruments presently available at ANSTO with emphasis on the SANS instrument. This will be followed by a description of the replacement research reactor and its instruments. (author)

Preliminary characterization of the passive neutron dose equivalent monitor with TLDs

Energy Technology Data Exchange (ETDEWEB)

Tsujimura, Norio; Kanai, Katsuta; Momose, Takumaro; Hayashi, Naomi [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan); Chen Erhu [Beijing Institute of Nuclear Engineering, Beijing (China)

2001-02-01

The passive neutron dose equivalent monitor with TLDs is composed of a cubic polyethylene moderator and TLDs at the center of moderator. This monitor was originally designed for measurements of neutron doses over long-term period of time around the nuclear facilities. In this study, the energy response of this monitor was calculated by Monte Carlo methods and experimentally obtained under {sup 241}Am-Be, {sup 252}Cf and moderated {sup 252}Cf neutron irradiation. Additionally, the responses of two types of conventional neutron dose equivalent meters (rem counters) were also investigated as comparison. The authors concluded that this passive neutron monitor with TLDs had a good energy response similar to conventional rem counters and could evaluate neutron doses within 10% of accuracy to the moderated fission spectra. (author)

Low energy neutrons from a sup 2 sup 3 sup 9 PuBe isotopic neutron source inserting in moderating media

CERN Document Server

Vega, H R

2002-01-01

Several neutron applications share a common problem: the neutron source design. In this work MCNP computer code has been used to design a moderated sup 2 sup 3 sup 9 PuBe neutron source to produce low energy neutrons. The design involves the source located at the center of a spherical moderator. Moderator media studied were light water, heavy water and a heterogeneous combination of light water and heavy water. Similar moderating features were found between the 24.5 cm-radius container filled with heavy water (23.0-cm-thick) and that made with light water (3.5-cm-thick) plus heavy water (19.5-cm-thick). A sup 2 sup 3 sup 9 PuBe neutron source inserted in this moderator produces, at 27 cm, a neutron fluence of 1.8 x 10 sup - sup 4 n-cm sup - sup 2 per source neutron, with an average neutron energy of 0.34 MeV, where 47.8 % have an energy <= 0.4 eV. A further study of this moderator was carried out using a reflector medium made of graphite. Thus, 15-cm-thickness reflector improves the neutron field producing...

SUSANS With Polarized Neutrons.

Science.gov (United States)

Wagh, Apoorva G; Rakhecha, Veer Chand; Strobl, Makus; Treimer, Wolfgang

2005-01-01

Super Ultra-Small Angle Neutron Scattering (SUSANS) studies over wave vector transfers of 10(-4) nm(-1) to 10(-3) nm(-1) afford information on micrometer-size agglomerates in samples. Using a right-angled magnetic air prism, we have achieved a separation of ≈10 arcsec between ≈2 arcsec wide up- and down-spin peaks of 0.54 nm neutrons. The SUSANS instrument has thus been equipped with the polarized neutron option. The samples are placed in a uniform vertical field of 8.8 × 10(4) A/m (1.1 kOe). Several magnetic alloy ribbon samples broaden the up-spin neutron peak significantly over the ±1.3 × 10(-3) nm(-1) range, while leaving the down-spin peak essentially unaltered. Fourier transforms of these SUSANS spectra corrected for the instrument resolution, yield micrometer-range pair distribution functions for up- and down-spin neutrons as well as the nuclear and magnetic scattering length density distributions in the samples.

Effect of Gamma Rays on Fast Neutron Registration in CR-39

CERN Document Server

Kobzev, A P; El-Halem, A A; Abdul-Ghaphar, U S; Salama, T A

2002-01-01

A set of CR-39 plastic detectors with front PE radiator was exposed to Am-Be neutron source, which has an emission rate of 0.86\\cdot 10^{7} sec^{-1}, and the neutron dose equivalent rate 1 m apart from the source is equal to 11 mrem/hr. Another set of samples was irradiated by a neutron dose of 4 rem, then exposed to different gamma-ray doses using ^{60}Co source. It was found that the track density grows with the increase of neutron dose and etching time. It was also found that the bulk etching rate V_{B}, the track diameter and the sensitivity of the CR-39 plastic detector with respect to the neutron irradiation increased with increasing gamma-ray dose in the range 1?10 Mrad. These results show that CR-39 can be considered as a promising fast neutron dosimeter and gamma-ray dosimeter.

Neutron scattering in soft matter physics and chemistry

International Nuclear Information System (INIS)

White, J.W.

1999-01-01

Recent experiments area of soft matter science show that self assembly on the micron scale as well as the nanometer scale can be directed chemically. This lecture illustrates how such processes can be studied using the contrast variation available in neutron scattering through isotopic replacement and the techniques of neutron small angle scattering and neutron reflectivity. Related dynamical information at nanometer resolution and on time scales between a nanosecond and a few tenths of a picosecond will become accessible with brighter neutron sources. The examples presented concern the template induced crystallisation of zeolites, the liquid crystal template induced synthesis of mesoporous materials and the structure of thin films at the air water interface. (J.P.N.)

Comparison between PGAA and ID-AMS analysis for determining chlorine content in whole rock basalt

Science.gov (United States)

di Nicola, L.; Schnabel, C.; Wilcken, K. M.; Gméling, K.

2009-04-01

Accurate determination of chlorine concentrations in terrestrial rocks is of importance for the interpretation of terrestrial in-situ cosmogenic 36Cl. Neutron capture by 35Cl, together with production from Ca and K, is one of the three major production pathways of 36Cl in rocks. Here, we present an inter-comparison of chlorine determinations by two procedures. The first approach is an independent Cl determination by prompt gamma (neutron) activation analysis (PGAA). The second method is isotope dilution based on isotopically-enriched stable chlorine carrier added during chemical sample preparation for accelerator mass spectrometry (ID-AMS). Twenty six (26) whole rock samples have been processed for PGAA and ID-AMS analyses. Elemental analysis by PGAA provides concentrations of major, minor and trace elements including the target elements for 36Cl production (K, Ca, Ti, and Fe), as well as of neutron absorbers and neutron moderators (H, B, Cl, Sm and Gd). The Cl concentrations determined during this study constitute the first inter-comparison for concentrations below 100 μCl/g. Our results show no significant difference in Cl concentrations between methods, and comparable uncertainties. This agreement guarantees that during the procedure we employ for whole rock sample no significant loss of stable chlorine from either the spike or the sample occurs before isotopic equilibration, prior to AgCl precipitation. Furthermore, we show that the elemental analysis by PGAA offers anadvance for the interpretation of 36Cl measurements. It allows simultaneous measurement of major and most trace element concentrations with a precision necessary for calculating the relative contributions to 36Cl production rates of the different mechanisms. Finally, the Cl concentration can be used to optimize the amount of isotopically-enriched spike for AMS-ID sample preparation for 36Cl.

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

CERN Document Server

Csikai, J; Sanami, T; Michikawa, T

2002-01-01

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

Activation measurements for thermal neutrons. Part F. 36Cl measurements in Japan

International Nuclear Information System (INIS)

Nagashima, Yasuo; Seki, Riki; Matsuhiro, Takeshi; Takahashi, Tsutomu; Sasa, Kimikazu; Usui, Toshihide; Sueki, Keisuke

2005-01-01

The development of the accelerator mass spectrometry (AMS) system at the Tandem Accelerator Center of the University of Tsukuba was started in 1995, using the university's own molecular pilot beam technique. Presently, it is the only facility in Japan used to measure 36 Cl (Nagashima et al. 2000). The sensitivity of the 36 Cl AMS system is around 10 -14 36 Cl/Cl atom ratio, which is enough to measure the natural level of the 36 Cl/Cl ratio. The system is characterized by long-term stability, enabling high-quality, continuous measurements over many hours. Our AMS system was used to measure 36 Cl produced in soil by neutrons released into the environment at the time of the JCO criticality accident in Tokai-mura in 1999 (Seki et al. 2003). At the beginning of 2001, our group joined the collaborative efforts to investigate and clarify the discrepancy observed between measurements and calculations of neutron activities induced by the atomic bombings in Hiroshima and Nagasaki. Using our AMS system, 36 Cl was measured in granite samples from Hiroshima exposed to atomic-bomb neutrons and in distant, unexposed samples. (author)

Neutron shielding point kernel integral calculation code for personal computer: PKN-pc

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Sakamoto, Yukio; Nakane, Yoshihiro; Tomita, Ken-ichi; Kurosawa, Naohiro.

1994-07-01

A personal computer version of PKN code, PKN-pc, has been developed to calculate neutron and secondary gamma-ray 1cm depth dose equivalents in water, ordinary concrete and iron for neutron source. Characteristics of PKN code are, to able to calculate dose equivalents in multi-layer three-dimensional system, which are described with two-dimensional surface, for monoenergetic neutron source from 0.01 to 14.9 MeV, 252 Cf fission and 241 Am-Be neutron source quick and easily. In addition to these features, the PKN-pc is possible to process interactive input and to get graphical system configuration and graphical results easily. (author)

Single-crystal neutron diffraction at the Australian Replacement Research Reactor

International Nuclear Information System (INIS)

Klooster, W.T.

2001-01-01

The purpose of the workshop was to: identify the future needs and opportunities for single-crystal neutron diffraction, and specify instrument requirements. important number of experiments. The conclusion of the workshop deliberation was that Australia has a diverse community of users of single-crystal neutron diffraction. A (quasi)-Laue image-plate diffractometer allows the fastest throughput by far, but would exclude an important number of experiments. Most of these could be covered by the additional possibility to locate the image-plate detector on a monochromatic beam. Therefore it was recommend both a white thermal beam and a monochromatic beam (λ= 1 to 2.4 Angstroms) for an image-plate detector. At little additional cost the existing 2TanA instrument could be located semi-permanently on the same monochromatic beam, thus offering three quite different types of single-crystal instruments. Small improvements could be made to the 2TanA instrument to cater for the remaining experiments not suited to an image-plate diffractometer: exchange of the Eulerian cradle for an automated tilt goniometer for extremely bulky sample environment (cryomagnets, large pressure cells), optional larger area detector, analyser crystal. It was recommended that an Instrument Advisory Team will be assembled, and will help in specifying, designing and commissioning the instrument

Thermal neutron calibration channel at LNMRI/IRD

International Nuclear Information System (INIS)

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

2014-01-01

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

The new Munich neutron source

International Nuclear Information System (INIS)

Herrmann, W.A.

1998-01-01

The Munich FRM II neutron source currently under construction is to replace the FRM I research reactor in Munich, also known as 'atomic egg'. The project is executed by the Free State of Bavaria as a construction project of the Munich Technical University and managed by the University. As main contractor for the construction project, Siemens AG is also co-applicant in the licensing procedure under the Atomic Energy Act for the construction phase. The project is carried out to build a modern high flux neutron source required for a broad range of applications in research and technology mainly with thermal and cold neutrons. The 'neutron gap' existing in Germany is to be closed with the FRM II. As a national research installation, the FRM II is available to all interested scientists from a variety of disciplines. (orig.) [de

Calibration of a TLD system to estimate personal doses in fields of gamma-neutrons radiation

International Nuclear Information System (INIS)

Villegas, E.N.; Somarriba, S.I.

2016-01-01

Currently Nicaragua has no personal neutron dosimetry system. The calibration of a batch of albedo neutron dosimeters consisting of two pairs of "6LiF and "7LiF (Mg, Ti) detectors was done. The dosimeter and reader sensitivities were obtained using a "1"3"7Cs source, and a neutron calibration factor was found with a "2"4"1AmBe source. Reproducibility and homogeneity tests were performed, and the detection limit of the system was determined. This calibration will allow the beginning of neutron personal monitoring in the country. (author)

Delayed neutron spectra and their uncertainties in fission product summation calculations

Energy Technology Data Exchange (ETDEWEB)

Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

1997-03-01

Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

Influenza B virus M2 protein can functionally replace its influenza A virus counterpart in promoting virus replication

International Nuclear Information System (INIS)

Wanitchang, Asawin; Wongthida, Phonphimon; Jongkaewwattana, Anan

2016-01-01

The M2 protein (AM2 and BM2) of influenza A and B viruses function as a proton channel essential for viral replication. They also carry a cytoplasmic tail whose functions are not fully delineated. It is currently unknown whether these proteins could be replaced functionally in a viral context. Here, we generated single-cycle influenza A viruses (scIAV-ΔHA) carrying various M2-2A-mCherry constructs in the segment 4 (HA) and evaluated their growth in complementing cells. Intriguingly, the scIAV-ΔHA carrying AM2 and that bearing BM2 grew comparably well in MDCK-HA cells. Furthermore, while the virus carrying chimeric B-AM2 in which the BM2 transmembrane fused with the AM2 cytoplasmic tail produced robust infection, the one bearing the AM2 transmembrane fused with the BM2 cytoplasmic tail (A-BM2) exhibited severely impaired growth. Altogether, we demonstrate that AM2 and BM2 are functionally interchangeable and underscore the role of compatibility between transmembrane and cytoplasmic tail of the M2 protein. -- Highlights: •Flu A M2 protein (AM2) can be functionally replaced by that of Flu B (BM2). •Both AM2 and BM2 with extended cytoplasmic tail are functional. •Compatibility between the ion channel and the cytoplasmic tail is critical for M2 function. •M2 with higher ion channel activity may augment influenza virus replication.

Influenza B virus M2 protein can functionally replace its influenza A virus counterpart in promoting virus replication

Energy Technology Data Exchange (ETDEWEB)

Wanitchang, Asawin; Wongthida, Phonphimon; Jongkaewwattana, Anan, E-mail: anan.jon@biotec.or.th

2016-11-15

The M2 protein (AM2 and BM2) of influenza A and B viruses function as a proton channel essential for viral replication. They also carry a cytoplasmic tail whose functions are not fully delineated. It is currently unknown whether these proteins could be replaced functionally in a viral context. Here, we generated single-cycle influenza A viruses (scIAV-ΔHA) carrying various M2-2A-mCherry constructs in the segment 4 (HA) and evaluated their growth in complementing cells. Intriguingly, the scIAV-ΔHA carrying AM2 and that bearing BM2 grew comparably well in MDCK-HA cells. Furthermore, while the virus carrying chimeric B-AM2 in which the BM2 transmembrane fused with the AM2 cytoplasmic tail produced robust infection, the one bearing the AM2 transmembrane fused with the BM2 cytoplasmic tail (A-BM2) exhibited severely impaired growth. Altogether, we demonstrate that AM2 and BM2 are functionally interchangeable and underscore the role of compatibility between transmembrane and cytoplasmic tail of the M2 protein. -- Highlights: •Flu A M2 protein (AM2) can be functionally replaced by that of Flu B (BM2). •Both AM2 and BM2 with extended cytoplasmic tail are functional. •Compatibility between the ion channel and the cytoplasmic tail is critical for M2 function. •M2 with higher ion channel activity may augment influenza virus replication.

On Jewish Being: Notes on Jean Améry

Directory of Open Access Journals (Sweden)

Andrew Benjamin

2017-02-01

Full Text Available That the question of identity takes on a sense of urgency, one with its own possibilities and impossibilities, the moment that identity is bound up with death, is hardy surprising. What follows are a series of reflections on the question of identity, Jewish identity, raised by Jean Améry’s remarkable text On the Necessity and Impossibility of Being a Jew (Über Zwang und Unmöglichkeit, Jude zu sein. Améry’s text was of course published in the wake of his own experiences as an active member of the resistance, as having been imprisoned in Auschwitz and as the victim of torture. Philosophically, rather than biographically, if there were a point of comparison, then it is to Levinas’s 1947 text Etre juif. Both pose the problem of how the question of Jewish identity, Jewish being, is to be understood in the wake of the Shoah. The meaning of the formulations - Jude zu sein, Jude sein, Etre juif, Jewish being – delimits the question to be addressed. This will be the case even if its point of address, namely what the question stages, is itself far from straightforward. Moreover, while what is demanded within that question is itself philosophically important, it is equally the case that the question of Jewish being is at work within both communities and synagogues across the Jewish world. As a consequence it is as much a philosophical question as it is one that has a structuring effect on how Jewish survival is conceived (and thus equally on what that survival is taken to be. How survival is understood is an issue that continues to exert its force. Who is the subject of survival? What is the subject of survival? Who or what has been subjected to the issue of survival? Survival is both more nuanced and complex than the brute fact of an afterlife. Jewish being as a present question – a question of the present - continues therefore.

Neutron Scattering at HIFAR—Glimpses of the Past

Directory of Open Access Journals (Sweden)

Margaret Elcombe

2017-04-01

Full Text Available This article attempts to give a description of neutron scattering down under for close on forty-six years. The early years describe the fledgling group buying parts and cobbling instruments together to its emergence as a viable neutron scattering group with up to ten working instruments. The second section covers the consolidation of this group, despite tough higher level management. The Australian Science and Technology Council (ASTEC enquiry in 1985 and the Government decision not to replace the HIgh Flux Australian Reactor (HIFAR, led to major expansion and upgrading of the existing neutron beam facilities during the 1990s. Finally, there were some smooth years of operation while other staff were preparing for the replacement reactor. It has concentrated on the instruments as they were built, modified, replaced with new ones, and upgraded at different times.

Development and characterization of two-component albedo based neutron individual monitoring system using thermoluminescent detectors; Desenvolvimento e caracterizacao de um sistema de monitoracao individual de neutrons tipo albedo de duas componentes usando detectores termoluminescentes

Energy Technology Data Exchange (ETDEWEB)

Martins, Marcelo Marques

2008-07-01

A TLD-albedo based two-component neutron individual monitoring system was developed and characterized in this work. The monitor consists of a black plastic holder, an incident neutron boron loaded shield, a moderator polyethylene body (to increase its response), two pairs of TLD-600 and TLD-700 (one pair to each component) and an adjustable belt. This monitoring system was calibrated in thermal neutron fields and in 70 keV, 144 keV, 565 keV, 1.2 MeV and 5 MeV monoenergetic neutron fields. In addition, it was calibrated in {sup 252C}f(D{sub 2}O), {sup 252}Cf, {sup 241}Am-B, {sup 241}Am-Be and {sup 238}Pu-Be source fields. For the latter, the lower detection levels are, respectively, 0.009 mSv, 0.06 mSv, 0.12 mSv, 0.09 mSv and 0.08 mSv. The participation in an international intercomparison sponsored by IAEA with simulated workplace fields validated the system. The monitoring system was successfully characterized in the ISO 21909 standard and in an IRD - the Brazilian Institute for Radioprotection and Dosimetry - technical regulation draft. Nowadays, the neutron individual system is in use by IRD for whole body individual monitoring of five institutions, which comprehend several activities. (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.

Design characteristics of a three-component AEOI Neutriran Albedo Neutron Personnel Dosimeter

International Nuclear Information System (INIS)

Sohrabi, M.; Katouzi, M.

1991-01-01

In establishing a national personnel neutron dosimetry service in Iran, different parameters of the AEOI Neutriran Albedo Neutron Personnel Dosimeter (NANPD) have been optimized. A NANPD was designed with three dosimetry components to measure (a) direct thermal neutrons, (b) direct fast neutrons and (C) direct neutrons by the detection of the albedo neutrons reflected from the body. The dosimeter consists of one or more Lexan polycarbonate and/or CR-39 foils and two 10 B (n,α) 7 Li converters in a cadmium cover so arranged as to efficiently measure the three neutron dose components separately. The boron converter thickness, its position relative to the beam direction and its distance from the PC foil were studied and the results were incorporated into the design. The dose response of the dosimeter, its lower detection limit as well as the correction factors related to the field neutrons and albedo neutrons were also determined for a 238 Pu-Be, an 241 Am-Be and a 252 Cf sources. In this paper, the dosimeter design and its dosimetric characteristics are presented and discussed. (author)

Polarized neutrons for Australian scientific research

International Nuclear Information System (INIS)

Kennedy, Shane J.

2005-01-01

Polarized neutron scattering has been a feature at ANSTO's HIFAR research reactor since the first polarization analysis (PA) spectrometer Longpol began operation over 30 years ago. Since that time, we have improved performance of Longpol and added new capabilities in several reincarnations of the instrument. Most of the polarized neutron experiments have been in the fields of magnetism and superconductivity, and most of that research has involved PA. Now as we plan our next generation neutron beam facility, at the Replacement Research Reactor (RRR), we intend to continue the tradition of PA but with a far broader scope in mind. Our new capabilities will combine PA and energy analysis with both cold and thermal neutron source spectra. We will also provide capabilities for research with polarized neutrons in small-angle neutron scattering and in neutron reflectometry. The discussion includes a brief historical account of the technical developments with a summary of past and present applications of polarized neutrons at HIFAR, and an outline of the polarized neutron capabilities that will be included in the first suite of instruments, which will begin operation at the new reactor in 2006

Preliminary research on measuring grease in petroleum pipeline using fast neutron transmission method

International Nuclear Information System (INIS)

Liu Qingwei; Liu Shengkang; Zhang Zhiping; Ding Xiaoping

2006-01-01

The principle, experiment and conclusion on the grease stain measurement using fast neutron are reported. The experiment equipment consist of 241 Am-Be fast neutron source, ZnS detector and BH1224 multichannel spectrometer. Paraffin is used instead of real grease stain. Steel plates are used instead of pipeline. The results of the experiment indicate that there is a good linearship between the logarithm of the reciprocal of the neutron transmissivity and the paraffin thickness. The measuring accuracy of the paraffin thickness is 0.6 mm in this experiment. (authors)

New thermal neutron calibration channel at LNMRI/IRD

International Nuclear Information System (INIS)

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

2015-01-01

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

Neutron spectrum determination of d(20)+Be source reaction by the dosimetry foils method

Science.gov (United States)

Stefanik, Milan; Bem, Pavel; Majerle, Mitja; Novak, Jan; Simeckova, Eva

2017-11-01

The cyclotron-based fast neutron generator with the thick beryllium target operated at the NPI Rez Fast Neutron Facility is primarily designed for the fast neutron production in the p+Be source reaction at 35 MeV. Besides the proton beam, the isochronous cyclotron U-120M at the NPI provides the deuterons in the energy range of 10-20 MeV. The experiments for neutron field investigation from the deuteron bombardment of thick beryllium target at 20 MeV were performed just recently. For the neutron spectrum measurement of the d(20)+Be source reaction, the dosimetry foils activation method was utilized. Neutron spectrum reconstruction from resulting reaction rates was performed using the SAND-II unfolding code and neutron cross-sections from the EAF-2010 nuclear data library. Obtained high-flux white neutron field from the d(20)+Be source is useful for the intensive irradiation experiments and cross-section data validation.

Development and characterization of two-component albedo based neutron individual monitoring system using thermoluminescent detectors

International Nuclear Information System (INIS)

Martins, Marcelo Marques

2008-01-01

A TLD-albedo based two-component neutron individual monitoring system was developed and characterized in this work. The monitor consists of a black plastic holder, an incident neutron boron loaded shield, a moderator polyethylene body (to increase its response), two pairs of TLD-600 and TLD-700 (one pair to each component) and an adjustable belt. This monitoring system was calibrated in thermal neutron fields and in 70 keV, 144 keV, 565 keV, 1.2 MeV and 5 MeV monoenergetic neutron fields. In addition, it was calibrated in 252C f(D 2 O), 252 Cf, 241 Am-B, 241 Am-Be and 238 Pu-Be source fields. For the latter, the lower detection levels are, respectively, 0.009 mSv, 0.06 mSv, 0.12 mSv, 0.09 mSv and 0.08 mSv. The participation in an international intercomparison sponsored by IAEA with simulated workplace fields validated the system. The monitoring system was successfully characterized in the ISO 21909 standard and in an IRD - the Brazilian Institute for Radioprotection and Dosimetry - technical regulation draft. Nowadays, the neutron individual system is in use by IRD for whole body individual monitoring of five institutions, which comprehend several activities. (author)

The measurement of thermal neutron constants of the soil; application to the calibration of neutron moisture gauges and to the pedological study of soil

International Nuclear Information System (INIS)

Couchat, P.; Marcesse, J.; Carre, C.; Le Ho, J.

1975-01-01

The neutronic method for measuring the water content of soils is more and more used by agronomists, hydrogeologists and pedologists. On the other hand the studies on the phenomena of slowing down and diffusion process have shown a narrow relation between the thermal absorption (Σ(a)) and diffusion (Σ(d)) constants and the thermal flux developed in the soil around a fast neutron source like Am-Be. Two original applications of the direct measurement of Σ(a) and Σ(d) are then presented. The method described consists in the measurement, in a cube of graphite with Am-Be source in the middle, on one side of the perturbation of the thermal flux, obtained by the introduction of 300g of soil, and on the other side of the transmitted thermal flux measured through the same sample of soil, on a side of the cube. After calibrating the device, these two parameters give Σ(a) and Σ(d) which are easily introduced in the calibration equation of neutron moisture gauge. Also these two values are useful for the pedologists because Σ(d) is connected to clay content in the soil and Σ(a) is connected to the type of clay by the way of rare earth contents [fr

Simulation of a room for neutron instrument calibration at LCR/UERJ

International Nuclear Information System (INIS)

Medeiros, M.P.C.; Estrada, J.J.S.; Gomes, R.G.; Santos, R.F.G.; Leite, S.P.; Alves, C.F.E.; Rebello, W.F.; Almeida, C.E. de

2013-01-01

In this work the MCNPX code was used to design a calibrating room for neutron detectors to be implemented in the Laboratorio de Ciencias Radiologicas of UERJ. The calibration room containing a neutron irradiator with a 241 Am-Be source, a linear positioning system, radiation detectors and a shadow cone was modeled. The ambient dose equivalent rate, ııı ∗ ı10ı, in adjacent to the calibration room areas, as well as neutron scattering caused by the room itself were calculated. Using an occupancy factor of 1/16 for all adjacent areas, 3.8 cm of 5% borated polyethylene or 5.5 cm of concrete for shielding is enough to satisfy radiation safety requirements. (author)

Characterization of Crystallographic Structures Using Bragg-Edge Neutron Imaging at the Spallation Neutron Source

Directory of Open Access Journals (Sweden)

Gian Song

2017-12-01

Full Text Available Over the past decade, wavelength-dependent neutron radiography, also known as Bragg-edge imaging, has been employed as a non-destructive bulk characterization method due to its sensitivity to coherent elastic neutron scattering that is associated with crystalline structures. Several analysis approaches have been developed to quantitatively determine crystalline orientation, lattice strain, and phase distribution. In this study, we report a systematic investigation of the crystal structures of metallic materials (such as selected textureless powder samples and additively manufactured (AM Inconel 718 samples, using Bragg-edge imaging at the Oak Ridge National Laboratory (ORNL Spallation Neutron Source (SNS. Firstly, we have implemented a phenomenological Gaussian-based fitting in a Python-based computer called iBeatles. Secondly, we have developed a model-based approach to analyze Bragg-edge transmission spectra, which allows quantitative determination of the crystallographic attributes. Moreover, neutron diffraction measurements were carried out to validate the Bragg-edge analytical methods. These results demonstrate that the microstructural complexity (in this case, texture plays a key role in determining the crystallographic parameters (lattice constant or interplanar spacing, which implies that the Bragg-edge image analysis methods must be carefully selected based on the material structures.

Further improvement for {sup 10}Be measurement on an upgraded compact AMS radiocarbon facility

Energy Technology Data Exchange (ETDEWEB)

Fu, Dongpo; Ding, Xingfang [State Key Laboratory of Nuclear Physics and Technology and Institute of Heavy Ion Physics, Peking University, Beijing 100871,China (China); Liu, Kexin, E-mail: kxliu@pku.edu.cn [State Key Laboratory of Nuclear Physics and Technology and Institute of Heavy Ion Physics, Peking University, Beijing 100871,China (China); Müller, Arnold Milenko; Suter, Martin; Christl, Marcus [Laboratory of Ion Beam Physics, ETH Zürich, 8093 Zürich (Switzerland); Zhou, Liping [Department of Geography, Peking University, Beijing 100871 (China); Synal, Hans-Arno [Laboratory of Ion Beam Physics, ETH Zürich, 8093 Zürich (Switzerland)

2015-10-15

The Peking University 500 kV NEC compact AMS radiocarbon facility (PKU-CAMS) has been modified in order to have additionally the possibility to measure {sup 10}Be. In the preliminary experiment a silicon nitride foil was mounted in front of the electrostatic deflector as passive boron degrader, and the original Si detector for radiocarbon detection was replaced by an ETHZ-designed high-resolution ΔE − E{sub res} gas ionization chamber (GIC) for {sup 10}Be identification. This simple arrangement has yielded an overall {sup 10}Be transmission of 2.2% and a {sup 10}Be/{sup 9}Be background level of 3.5 × 10{sup −14}. To further reduce the background and increase the transmission by re-focusing the {sup 10}Be ions, an additional 90° bending magnet with 350 mm radius was installed after the electrostatic deflector. The silicon detector was shifted slightly relative to its position of original NEC system setup in opposite direction of beam and can be lifted up manually without breaking vacuum when {sup 10}Be measurements are carried out. In this way the system can be easily and fast set up for {sup 10}Be without affecting any parameters for radiocarbon measurement. The gas detector for {sup 10}Be was mounted at the end of the beam line after the additional magnet. The lay-out of the upgraded spectrometer is very compact and does not require more space than the original instrument. Using this compact setup, the overall transmission for {sup 10}Be was doubled to 5–6% and the {sup 10}Be/{sup 9}Be background level was reduced to radios as low as 2.4 × 10{sup −15}.

The new 6 MV multi-nuclide AMS facility at the University of Tsukuba

Science.gov (United States)

Sasa, Kimikazu; Takahashi, Tsutomu; Matsumura, Masumi; Matsunaka, Tetsuya; Satou, Yukihiko; Izumi, Daiki; Sueki, Keisuke

2015-10-01

The former accelerator mass spectrometry (AMS) system installed on the 12UD Pelletron tandem accelerator at the University of Tsukuba was completely destroyed by the Great East Japan Earthquake on 11 March 2011. A replacement has been designed and constructed at the university as part of the post-quake reconstruction project. It consists of a 6 MV Pelletron tandem accelerator, two multiple cathode AMS ion sources (MC-SNICSs), and a rare-particle detection system. The 6 MV Pelletron tandem accelerator will be applied not only to AMS, but also to areas such as nanotechnology, ion beam analysis, heavy ion irradiation, and nuclear physics. The rare-particle detection system will be capable of measuring environmental levels of long-lived radioisotopes of 10Be, 14C, 26Al, 36Cl, 41Ca, and 129I. It is also expected to measure other radioisotopes such as 32Si and 90Sr. The 6 MV Pelletron tandem accelerator was installed in the spring of 2014 at the University of Tsukuba. Routine beam delivery and AMS experiments will start in 2015.

The neutron field perturbation effect in the Dalat Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The perturbation effect of the thermal neutron field of the Dalat reactor is investigated when a fuel element is replaced by a water column or a plexiglass rod. In consequence, it is possible to replace the measurement of the relative distribution of the thermal neutron field on the surface of fuel element by that in the water column or in the plexiglass rod. (author). 5 refs. 4 figs. 4 tabs.

Insight into the Am-O Phase Equilibria: A Thermodynamic Study Coupling High-Temperature XRD and CALPHAD Modeling.

Science.gov (United States)

Epifano, Enrica; Guéneau, Christine; Belin, Renaud C; Vauchy, Romain; Lebreton, Florent; Richaud, Jean-Christophe; Joly, Alexis; Valot, Christophe; Martin, Philippe M

2017-07-03

In the frame of minor actinide transmutation, americium can be diluted in UO 2 and (U, Pu)O 2 fuels burned in fast neutron reactors. The first mandatory step to foresee the influence of Am on the in-reactor behavior of transmutation targets or fuel is to have fundamental knowledge of the Am-O binary system and, in particular, of the AmO 2-x phase. In this study, we coupled HT-XRD (high-temperature X-ray diffraction) experiments with CALPHAD thermodynamic modeling to provide new insights into the structural properties and phase equilibria in the AmO 2-x -AmO 1.61+x -Am 2 O 3 domain. Because of this approach, we were able for the first time to assess the relationships between temperature, lattice parameter, and hypostoichiometry for fcc AmO 2-x . We showed the presence of a hyperstoichiometric existence domain for the bcc AmO 1.61+x phase and the absence of a miscibility gap in the fcc AmO 2-x phase, contrary to previous representations of the phase diagram. Finally, with the new experimental data, a new CALPHAD thermodynamic model of the Am-O system was developed, and an improved version of the phase diagram is presented.

Characteristics of neutron-irradiated CR-39 foils treated by sequential chemical and electrochemical etching

International Nuclear Information System (INIS)

Somogyi, G.; Dajko, G.; Turek, K.; Spurny, F.

1982-01-01

The density of background spots revealed by chemical (CE) and electrochemical (ECE) etching and by their sequential application (CE + ECE) has been measured in several sorts of CR-39 material. The trends in the variation of sensitivity to Am-Be neutrons have been determined in CR-39 sheets covered by thick proton-radiator, when changing the field strength, frequency, etchant concentration, pre-etch duration and the fluence of neutrons. The results are analyzed in order to find out an optimum set of experimental parameters which may be proposed to attain high registration sensitivity to neutrons. (author)

Experimental evaluation of neutron dose in radiotherapy patients: Which dose?

Energy Technology Data Exchange (ETDEWEB)

Romero-Expósito, M., E-mail: mariateresa.romero@uab.cat; Domingo, C.; Ortega-Gelabert, O.; Gallego, S. [Grup de Recerca en Radiacions Ionizants (GRRI), Departament de Física, Universitat Autònoma de Barcelona, Bellaterra 08193 (Spain); Sánchez-Doblado, F. [Departamento de Fisiología Médica y Biofísica, Universidad de Sevilla, Sevilla 41009 (Spain); Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41009 (Spain)

2016-01-15

Purpose: The evaluation of peripheral dose has become a relevant issue recently, in particular, the contribution of secondary neutrons. However, after the revision of the Recommendations of the International Commission on Radiological Protection, there has been a lack of experimental procedure for its evaluation. Specifically, the problem comes from the replacement of organ dose equivalent by the organ-equivalent dose, being the latter “immeasurable” by definition. Therefore, dose equivalent has to be still used although it needs the calculation of the radiation quality factor Q, which depends on the unrestricted linear energy transfer, for the specific neutron irradiation conditions. On the other hand, equivalent dose is computed through the radiation weighting factor w{sub R}, which can be easily calculated using the continuous function provided by the recommendations. The aim of the paper is to compare the dose equivalent evaluated following the definition, that is, using Q, with the values obtained by replacing the quality factor with w{sub R}. Methods: Dose equivalents were estimated in selected points inside a phantom. Two types of medical environments were chosen for the irradiations: a photon- and a proton-therapy facility. For the estimation of dose equivalent, a poly-allyl-diglicol-carbonate-based neutron dosimeter was used for neutron fluence measurements and, additionally, Monte Carlo simulations were performed to obtain the energy spectrum of the fluence in each point. Results: The main contribution to dose equivalent comes from neutrons with energy higher than 0.1 MeV, even when they represent the smallest contribution in fluence. For this range of energy, the radiation quality factor and the radiation weighting factor are approximately equal. Then, dose equivalents evaluated using both factors are compatible, with differences below 12%. Conclusions: Quality factor can be replaced by the radiation weighting factor in the evaluation of dose

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Use of neutron activation and X-ray fluorescence with radioactive sources (Cf-252 and Am-241) for the instrumental qualiquantitative simultaneous analysis of some elements in samples of mineral supplement for animals

International Nuclear Information System (INIS)

Simabuco, S.M.

1984-01-01

To study the possibility of using two non-destructive (neutron activation and X-ray fluorescence) analyses in simultaneous quali-quantitative evaluations of some elements in mineral supplement for animals, a Cf-252 neutron source (11.3 mCi; 21.1 μgrams) and a Am-241 low energy gamma-ray emitter source (59.5 KeV; 100 mCi) were employed. For these sources, shieldings and sample irradiation systems were built. For the neutron activation analysis a reservoir of 72 cm height and 43 cm diameter was filled with paraffine, and the samples and neutron sources were put inside this reservoir using polypropilene and nylon tubes. To detect the gamma-rays emitted by the radioisotopes a well-type solid NaI(Tl) crystal scintillator (3x3') was used, coupled to a multi-channel analyser. For the X-ray fluorescence analysis a lead cylinder of 9.75 cm height and 5.6 cm diameter (with 0.7 cm thickness) was made and internally lined with a 0.36 mm copper and 0.1 mm aluminium foil. (Author) [pt

. Estimating soil contamination from oil spill using neutron backscattering technique

International Nuclear Information System (INIS)

Okunade, I.O.; Jonah, S.A.; Abdulsalam, M.O.

2009-01-01

An analytical facility which is based on neutron backscattering technique has been adapted for monitoring oil spill. The facility which consists of 1 Ci Am-Be isotopic source and 3 He neutron detector is based on the principle of slowing down of neutrons in a given medium which is dominated by the elastic process with the hydrogen nucleus. Based on this principle, the neutron reflection parameter in the presence of hydrogenous materials such as coal, crude oil and other hydrocarbon materials depends strongly on the number of hydrogen nuclei present. Consequently, the facility has been adapted for quantification of crude oil in soil contaminated in this work. The description of the facility and analytical procedures for quantification of oil spill in soil contaminated with different amount of crude oil are provided

Comparative experimental and theoretical investigations of the DM neutron moisture probe

DEFF Research Database (Denmark)

Ølgaard, Povl Lebeck; Haahr, Vagner

1967-01-01

Theoretical and experimental investigations of the Danish produced DM subsurface moisture probe have been carried out at the Research Establishment Risö, and the results obtained are presented in this paper. The DM probe contains an Am-Be fast neutron source and has a glass scintillator containing...

BONDI-97 A novel neutron energy spectrum unfolding tool using a genetic algorithm

CERN Document Server

Mukherjee, B

1999-01-01

The neutron spectrum unfolding procedure using the count rate data obtained from a set of Bonner sphere neutron detectors requires the solution of the Fredholm integral equation of the first kind by using complex mathematical methods. This paper reports a new approach for the unfolding of neutron spectra using the Genetic Algorithm tool BONDI-97 (BOnner sphere Neutron DIfferentiation). The BONDI-97 was used as the input for Genetic Algorithm engine EVOLVER to search for a globally optimised solution vector from a population of randomly generated solutions. This solution vector corresponds to the unfolded neutron energy spectrum. The Genetic Algorithm engine emulates the Darwinian 'Survival of the Fittest' strategy, the key ingredient of the 'Theory of Evolution'. The spectra of sup 2 sup 4 sup 1 Am/Be (alpha,n) and sup 2 sup 3 sup 9 Pu/Be (alpha,n) neutron sources were unfolded using the BONDI-97 tool. (author)

Response of a neutron monitor area with TLDs pairs

Energy Technology Data Exchange (ETDEWEB)

Guzman G, K. A.; Borja H, C. G.; Valero L, C.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Gallego, E.; Lorente, A., E-mail: ing_karen_guzman@yahoo.com.mx [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)

2011-10-15

The response of a passive neutron monitor area has been calculated using the Monte Carlo code MCNP5. The response was the amount of n({sup 6}Li, T){alpha} reactions occurring in a TLD-600 located at the center of a cylindrical polyethylene moderator. Fluence, (n, a) and H*(10) responses were calculated for 47 monoenergetic neutron sources. The H*(10) relative response was compared with responses of commercially available neutron monitors being alike. Due to {sup 6}Li cross section (n, {alpha}) reactions are mainly produced by thermal neutrons, however TLD-600 is sensitive to gamma-rays; to eliminate the signal due to photons monitor area was built to hold 2 pairs of TLD-600 and 2 pairs of TLD-700, thus from the difference between TLD-600 and TLD-700 readouts the net signal due to neutrons is obtained. The monitor area was calibrated at the Universidad Politecnica de Madrid using a {sup 241}AmBe neutron source; net TLD readout was compared with the H*(10) measured with a Bert hold Lb-6411. Performance of the neutron monitor area was determined through two independent experiments, in both cases the H*(10) was statistically equal to H*(10) measured with a Bert hold Lb-6411. Neutron monitor area with TLDs pairs can be used in working areas with intense, mixed and pulsed radiation fields. (Author)

Response of CMS avalanche photo-diodes to low energy neutrons

Science.gov (United States)

Brown, R. M.; Deiters, K.; Ingram, Q.; Renker, D.

2012-12-01

The response of the Avalanche Photo-diodes (APDs) installed in the CMS detector at the LHC to neutrons from 241AmBe and 252Cf sources is reported. Signals in size equivalent to those of up to 106 photo-electrons with the nominal APD gain are observed. Measurements with an APD with the protective epoxy coating removed and with the source placed behind the APD show that there is an important response due to recoil protons from neutron interactions with the hydrogen in the epoxy, in addition to signals from neutron interactions with the silicon of the diode. The effective gain of these signals is much smaller than the diode's nominal gain.

Determination of the scattered radiation at the Neutron Calibration Laboratory of IPEN, SP, Brazil

International Nuclear Information System (INIS)

Alvarenga, Tallyson; Valeriano, Caio C.S.; Caldas, Linda V.E.; Federico, Claudio A.

2016-01-01

With the increased use of techniques using neutron radiation, there has been a considerable growth in the number of detectors for this kind of radiation. A neutron calibration laboratory with neutron radiation ("2"4"1AmBe) was designed. In practical situations of this type of laboratory, one of the main problems is related to the knowledge of scattered radiation. In order to evaluate this scattered radiation, simulations were carried out without the presence of structural elements and with the complete room. Fourteen measuring points were evaluated in different directions at various distances. (author)

Bubble detector's evaluation for neutron field measurement in a very known source

Energy Technology Data Exchange (ETDEWEB)

Ramalho, Eduardo; Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.b, E-mail: jdantas@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Reina, Luiz, E-mail: reina@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Facure, Alessandro, E-mail: facure@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2011-07-01

Measurements on neutron fields, mainly for dosimetric purposes, have been a major concern for ionizing radiation workers, because of the radiation protection issues. The present work aims to study the using of bubble detectors in neutron dosimetry and the Bubble Detector Spectrometer (BDS) was chosen for this task. Several experiments were performed in order to obtain spectra from such devices and their respective analysis and then they were compared to those which were obtained by other ways. An Am-Be calibration neutron source from Instituto de Radioprotecao e Dosimetria/Comissao Nacional de Energia Nuclear (IRD/CNEN) was used and its spectrum was compared to the one obtained by BDS. The possibility of the use of such devices as ambient dosimeters was also evaluated. Despite the uncertainties, especially in the lowest energy thresholds, the spectrum from BDS is in good agreement with the known ones and the use of BDS as a dosimeter demands a more detailed study due to some characteristics of the Am-Be source that produce high uncertainties in low energy thresholds. (author)

Fission Cross-section Measurements of (233)U, (245)Cm and (241,243)Am at CERN n_TOF Facility

CERN Document Server

Calviani, M; Andriamonje, S; Chiaveri, E; Vlachoudis, V; Colonna, N; Meaze, M H; Marrone, S; Tagliente, G; Terlizzi, R; Belloni, F; Abbondanno, U; Fujii, K; Milazzo, P M; Moreau, C; Aerts, G; Berthoumieux, E; Dridi, W; Gunsing, F; Pancin, J; Perrot, L; Plukis, A; Alvarez, H; Duran, I; Paradela, C; Alvarez-Velarde, F; Cano-Ott, D; Gonzalez-Romero, E; Guerrero, C; Martinez, T; Villamarin, D; Vicente, M C; Andrzejewski, J; Marganiec, J; Assimakopoulos, P; Karadimos, D; Karamanis, D; Papachristodoulou, C; Patronis, N; Audouin, L; David, S; Ferrant, L; Isaev, S; Stephan, C; Tassan-Got, L; Badurek, G; Jericha, E; Leeb, H; Oberhummer, H; Pigni, M T; Baumann, P; Kerveno, M; Lukic, S; Rudolf, G; Becvar, F; Krticka, M; Calvino, F; Capote, R; Carrillo De Albornoz, A; Marques, L; Salgado, J; Tavora, L; Vaz, P; Cennini, P; Dahlfors, M; Ferrari, A; Gramegna, F; Herrera-Martinez, A; Kadi, Y; Mastinu, P; Praena, J; Sarchiapone, L; Wendler, H; Chepel, V; Ferreira-Marques, R; Goncalves, I; Lindote, A; Lopes, I; Neves, F; Cortes, G; Poch, A; Pretel, C; Couture, A; Cox, J; O'brien, S; Wiescher, M; Dillman, I; Heil, M; Kappeler, F; Mosconi, M; Plag, R; Voss, F; Walter, S; Wisshak, K; Dolfini, R; Rubbia, C; Domingo-Pardo, C; Tain, J L; Eleftheriadis, C; Savvidis, I; Frais-Koelbl, H; Griesmayer, E; Furman, W; Konovalov, V; Goverdovski, A; Ketlerov, V; Haas, B; Haight, R; Reifarth, R; Igashira, M; Koehler, P; Kossionides, E; Lampoudis, C; Lozano, M; Quesada, J; Massimi, C; Vannini, G; Mengoni, A; Oshima, M; Papadopoulos, C; Vlastou, R; Pavlik, A; Pavlopoulos, P; Plompen, A; Rullhusen, P; Rauscher, T; Rosetti, M; Ventura, A

2011-01-01

Neutron-induced fission cross-sections of minor actinides have been measured using the n_TOF white neutron source at CERN, Geneva, as part of a large experimental program aiming at collecting new data relevant for nuclear astrophysics and for the design of advanced reactor systems. The measurements at n_TOF take advantage of the innovative features of the n_TOF facility, namely the wide energy range, high instantaneous neutron flux and good energy resolution. Final results on the fission cross-section of 233U, 245Cm and 243Am from thermal to 20 MeV are here reported, together with preliminary results for 241Am. The measurement have been performed with a dedicated Fast Ionization Chamber (FIC), a fission fragment detector with a very high efficiency, relative to the very well known cross-section of 235U, measured simultaneously with the same detector.

{sup 124}Sb–Be photo-neutron source for BNCT: Is it possible?

Energy Technology Data Exchange (ETDEWEB)

Golshanian, Mohadeseh [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Department of Physics, Shahrood University, Shahrood (Iran, Islamic Republic of); Rajabi, Ali Akbar [Department of Physics, Shahrood University, Shahrood (Iran, Islamic Republic of); Kasesaz, Yaser, E-mail: ykasesaz@aeoi.org.ir [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

2016-11-01

In this research a computational feasibility study has been done on the use of {sup 124}SbBe photo-neutron source for Boron Neutron Capture Therapy (BNCT) using MCNPX Monte Carlo code. For this purpose, a special beam shaping assembly has been designed to provide an appropriate epithermal neutron beam suitable for BNCT. The final result shows that using 150 kCi of {sup 124}Sb, the epithermal neutron flux at the designed beam exit is 0.23×10{sup 9} (n/cm{sup 2} s). In-phantom dose analysis indicates that treatment time for a brain tumor is about 40 min which is a reasonable time. This high activity {sup 124}Sb could be achieved using three 50 kCi rods of {sup 124}Sb which can be produced in a research reactor. It is clear, that as this activity is several hundred times the activity of a typical cobalt radiotherapy source, issues related to handling, safety and security must be addressed.

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

International Nuclear Information System (INIS)

Cardenas, Jose Patricio Nahuel

2010-01-01

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

Neutron activation: an invaluable technique for teaching applied radiation

International Nuclear Information System (INIS)

Trainer, Matthew

2002-01-01

This experiment introduces students to the important method of neutron activation. A sample of aluminium was irradiated with neutrons from an isotropic 241 Am-Be source. Using γ-ray spectroscopy, two radionuclide products were identified as 27 Mg and 28 Al. Applying a cadmium cut-off filter and an optimum irradiation time of 45 min, the half-life of 27 Mg was determined as 9.46±0.50 min. The half-life of the 28 Al radionuclide was determined as 2.28±0.10 min using a polythene moderator and an optimum irradiation time of 10 min. (author)

Optimization of CR-39 for fast neutron dosimetry applications

International Nuclear Information System (INIS)

Vilela, E.; Fantuzzi, E.; Giacomelli, G.; Giorgini, M.; Morelli, B.; Patrizii, L.; Serra, P.; Togo, V.

1999-01-01

We present the results of an experimental work aimed at improving the performances of the CR-39[reg] (Registered Trademark of PPG Industries Inc.) nuclear track detector for neutron dosimetry applications. The work was done in collaboration with the Intercast Europe S.p.A., producer of CR-39 for commercial and scientific applications. We compare the CR-39 made with different additives concentrations and different polymerization processes. We evaluate the response of the CR-39 to fast neutrons from three sources: 241 Am-Be, 252 Cf and 238 Pu-Li. Particular attention was paid to background fluctuations that limit the lower detectable dose

Measurement of low neutron-fluences using electrochemically etched PC and PET track detectors

International Nuclear Information System (INIS)

Somogyi, G.; Dajko, G.; Turek, K.; Spurny, F.

1979-01-01

Systematic investigations have been carried out to study different properties of electrochemically etched (ECE) polycarbonate (PC) and polyethylene-terephthalate (PET) foils. The dependence of the density of background discharge spots on surface-thickness removal, electrical field strength and frequency of voltage is given. The effect of these parameters on the neutron sensitivity of polycarbonate and polyethylene-terephthalate foils irradiated at right angles to 14.7 MeV, 241 Am-Be and 252 Cf neutrons is also studied. With knowledge of the background and sensitivity data, the etching and electrical parameters are optimized for low neutron-fluence measurements. (author)

New thermal neutron calibration channel at LNMRI/IRD

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Measurement of highly enriched uranium metal buttons with the high-level neutron coincidence counter operating in the active mode

International Nuclear Information System (INIS)

Foley, J.E.

1980-10-01

The portable High-Level Neutron Coincidence Counter is used in the active mode with the addition of AmLi neutron sources to assay the 235 U content of highly enriched metal pieces or buttons. It is concluded that the portable instrument is a practical instrument for assaying uranium metal buttons with masses in the range 1.5 to 4 kg

Neutron reference spectra measurements with the Bonner multi-spheres spectrometer; Medidas de espectros de referencia de neutrons com o espectrometro de multiesferas de Bonner

Energy Technology Data Exchange (ETDEWEB)

Lemos Junior, Roberto Mendonca de

2004-07-01

This paper aims to define a procedure to use the Bonner Multisphere Spectrometer with a {sup 6}LiI(Eu) detector in order to determine of neutron spectra. It was measured {sup 238}PuBe spectra and same of reference ({sup 241}AmBe, {sup 252}Cf e {sup 252}Cf+D{sub 2}O) published in ISO 8529-1 (2001) Norm. The data were processed by a computer program (BUNKI), which presents the results in neutrons energy fluency. Each input parameter of the program was studied in order to establish their influence in the adjustment result. The environment dose equivalent rate obtained placing the detector 1 m from the {sup 241}AmBe source was 122 {+-} 4 {mu}Sv/h with 7% of uncertainty and 95% of confidence level. The procedure established in this work was tested with the {sup 238}PuBe spectrum, obtaining an environment dose equivalent rate of 286 {+-} 9 {mu}Sv/h, 8% lower than the value measured experimentally used as reference. Through this procedure will be possible to measure neutron spectra in different work places where neutrons sources are used. Knowing these spectra, it will be possible to evaluate which area monitors, are more suitable, as well as, to study better the response of individual neutron monitors, as for instance, to obtain a conversion coefficient more appropriate to the albedo dosimeter used in different work places. As the measurements need a long time to be accomplished, the work optimization is fundamental to reduce the exposing time of the Bonner spectrometer operator. For this reason, an important parameter examined in this paper was the possibility of reducing the number of spheres used during the measurement without changing the final result. Considering the radiation protection standards, this parameter has a huge importance when the measurements are performed in work places where the neutron fluency and gamma rate offer risks to the operator's health, as for instance, in nuclear centrals. Studying this parameter, it was possible to conclude that

Neutron Standards Laboratory of the CIEMAT

International Nuclear Information System (INIS)

Guzman G, K. A.; Mendez V, R.; Vega C, H. R.

2014-08-01

By means of a calculation series with Monte Carlo methods and the code MCNPX was characterized the neutrons field produced by the existent calibration sources in the Neutron Standards Laboratory of the Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT). The laboratory has two neutron calibration sources one of 241 AmBe and other 252 Cf that are stored in a water pool. A detailed three-dimensional model of the room was built with the base of stainless steel remarking in the selector to the sources that situates them to 4 m of the floor to be irradiated on the irradiation table and the storage pool. Each one of the sources was defined on the model in its double steel encapsulated. The spectra were calculated with different cases with the purpose of to calculate the contribution of each element that impacts to the neutrons transport. The spectra of the calibration sources were calculated to different distances regarding the source from 0, 15, 35, 50 to 300 cm on the base and in a same way the values of the ambient dose equivalent using the approaches of the ICRP-74. The results show clearly that the great contribution in the modification of the spectrum is attributed to the walls, and floor of the Neutron Standards Laboratory installations. (Author)

Soil moisture determination with Tesla NZK 203 neutron gage

International Nuclear Information System (INIS)

Hally, J.

1977-01-01

Soil moisture was measured using the NZK 203 neutron probe manufactured by Tesla Premysleni. The individual measuring sites were spaced at a distance of 100 m. The NZK 203 set consists of a NPK 202 moisture gage and a NSK 301 scintillation detector and features the following specifications: moisture density measuring range 20 to 500 kg/m 3 , 241 Am-Be fast neutron source having a neutron flux of 7.5x10 4 n.sec -1 +-10%, operating temperature -10 to +45 degC. The measured counting rate was primarily affected by the statistical fluctuation of ionizing radiation and by instrument instability. In order that these effects should be limited each measurement was repeated 10 times with the optimum measurement time at an interval of 20 to 100 sec. The NZK 203 Tesla set was proven to be suitable for rapid and reproducible determination of moisture profiles. (J.P.)

Progress on multi-nuclide AMS of JAEA-AMS-TONO

Science.gov (United States)

Saito-Kokubu, Yoko; Matsubara, Akihiro; Miyake, Masayasu; Nishizawa, Akimitsu; Ohwaki, Yoshio; Nishio, Tomohiro; Sanada, Katsuki; Hanaki, Tatsumi

2015-10-01

The JAEA-AMS-TONO (Japan Atomic Energy Agency's Accelerator Mass Spectrometer established at the Tono Geoscience Center) facility has been used for the dating of geological samples. The AMS system is versatile, based on a 5 MV tandem Pelletron-type accelerator. Since its establishment in 1997, the AMS system has been used for measurement of carbon-14 (14C) mainly for 14C dating studies in neotectonics and hydrogeology, in support of JAEA's research on geosphere stability applicable to the long-term isolation of high-level radioactive waste. Results of the measurement of 14C in soils and plants has been applied to the dating of fault activity and volcanism. Development of beryllium-10 (10Be) and aluminum-26 (26Al) AMS systems are now underway to enhance the capability of the multi-nuclide AMS in studies of dating by cosmogenic nuclides. The 10Be-AMS system has already been used for routine measurements in applied studies and improvements of the measurement technique have been made. Now we plan to fine tune the system and perform test measurements to develop the 26Al-AMS system.

Progress on multi-nuclide AMS of JAEA-AMS-TONO

Energy Technology Data Exchange (ETDEWEB)

Saito-Kokubu, Yoko, E-mail: kokubu.yoko@jaea.go.jp [Japan Atomic Energy Agency, Toki, Gifu 509-5102 (Japan); Matsubara, Akihiro [Japan Atomic Energy Agency, Toki, Gifu 509-5102 (Japan); Miyake, Masayasu; Nishizawa, Akimitsu; Ohwaki, Yoshio; Nishio, Tomohiro; Sanada, Katsuki [Pesco Corp., Ltd., Toki, Gifu 509-5123 (Japan); Hanaki, Tatsumi [Japan Atomic Energy Agency, Toki, Gifu 509-5102 (Japan)

2015-10-15

The JAEA-AMS-TONO (Japan Atomic Energy Agency’s Accelerator Mass Spectrometer established at the Tono Geoscience Center) facility has been used for the dating of geological samples. The AMS system is versatile, based on a 5 MV tandem Pelletron-type accelerator. Since its establishment in 1997, the AMS system has been used for measurement of carbon-14 ({sup 14}C) mainly for {sup 14}C dating studies in neotectonics and hydrogeology, in support of JAEA’s research on geosphere stability applicable to the long-term isolation of high-level radioactive waste. Results of the measurement of {sup 14}C in soils and plants has been applied to the dating of fault activity and volcanism. Development of beryllium-10 ({sup 10}Be) and aluminum-26 ({sup 26}Al) AMS systems are now underway to enhance the capability of the multi-nuclide AMS in studies of dating by cosmogenic nuclides. The {sup 10}Be-AMS system has already been used for routine measurements in applied studies and improvements of the measurement technique have been made. Now we plan to fine tune the system and perform test measurements to develop the {sup 26}Al-AMS system.

A high-resolution neutron spectra unfolding method using the Genetic Algorithm technique

CERN Document Server

Mukherjee, B

2002-01-01

The Bonner sphere spectrometers (BSS) are commonly used to determine the neutron spectra within various nuclear facilities. Sophisticated mathematical tools are used to unfold the neutron energy distribution from the output data of the BSS. This paper highlights a novel high-resolution neutron spectra-unfolding method using the Genetic Algorithm (GA) technique. The GA imitates the biological evolution process prevailing in the nature to solve complex optimisation problems. The GA method was utilised to evaluate the neutron energy distribution, average energy, fluence and equivalent dose rates at important work places of a DIDO class research reactor and a high-energy superconducting heavy ion cyclotron. The spectrometer was calibrated with a sup 2 sup 4 sup 1 Am/Be (alpha,n) neutron standard source. The results of the GA method agreed satisfactorily with the results obtained by using the well-known BUNKI neutron spectra unfolding code.

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)

Fast-neutron and gamma-ray imaging with a capillary liquid xenon converter coupled to a gaseous photomultiplier

Science.gov (United States)

Israelashvili, I.; Coimbra, A. E. C.; Vartsky, D.; Arazi, L.; Shchemelinin, S.; Caspi, E. N.; Breskin, A.

2017-09-01

Gamma-ray and fast-neutron imaging was performed with a novel liquid xenon (LXe) scintillation detector read out by a Gaseous Photomultiplier (GPM). The 100 mm diameter detector prototype comprised a capillary-filled LXe converter/scintillator, coupled to a triple-THGEM imaging-GPM, with its first electrode coated by a CsI UV-photocathode, operated in Ne/5%CH4 at cryogenic temperatures. Radiation localization in 2D was derived from scintillation-induced photoelectron avalanches, measured on the GPM's segmented anode. The localization properties of 60Co gamma-rays and a mixed fast-neutron/gamma-ray field from an AmBe neutron source were derived from irradiation of a Pb edge absorber. Spatial resolutions of 12± 2 mm and 10± 2 mm (FWHM) were reached with 60Co and AmBe sources, respectively. The experimental results are in good agreement with GEANT4 simulations. The calculated ultimate expected resolutions for our application-relevant 4.4 and 15.1 MeV gamma-rays and 1-15 MeV neutrons are 2-4 mm and ~ 2 mm (FWHM), respectively. These results indicate the potential applicability of the new detector concept to Fast-Neutron Resonance Radiography (FNRR) and Dual-Discrete-Energy Gamma Radiography (DDEGR) of large objects.

Calibration of neutrons monitors with moderators and application in the calibration factors of albedo dosemeters

International Nuclear Information System (INIS)

Schuch, L.A.

1978-11-01

The calibration factors and the reproducibility of an Albedo Dosimeter designed for personal neutron monitoring were determined. These factor were obtained simulating the dosimeter reading and the equivalent dose in the locality by a convenient combination of responses of the Bonner Sphere Spectrometer. The results obtained in the simulation were verified experimentally for different spectra employing the Am-Be, bare 252 Cf source and 253 Cf source with graphite sields of varying thickness. Different standards were used in the procedures necessary for the determination of the calibration factors. An Am-Be neutron source, standardized by the activation of a manganese sulphate bath was used as a primary standard. As a secondary standard, for the measurement of the neutron fluence, a De Pangher Long Counter was used and the scattering effects were determined using the shadow cone method. The other monitors such as the Rem-Counter and the Bonner Sphere Spectrometer were also calibrated with reference to the secondary standard with a view to comparing the results obtained with those furnished by the Albedo Dosimeter. (Author) [pt

Study of a transportable neutron radiography system

International Nuclear Information System (INIS)

Souza, S.N.A. de.

1991-05-01

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

Anisotropy of neutrons sources of the Neutron Metrology Laboratory; Anisotropia de fontes de nêutrons do Laboratório de Metrologia de Nêutrons

Energy Technology Data Exchange (ETDEWEB)

Silva, A.C.F., E-mail: alexander.camargo@oi.com.br [Fundação Técnico Educacional Souza Marques, Rio de Janeiro, RJ (Brazil); Silva, F.S.; Creazolla, P.G.; Patrão, K.C.S.; Fonseca, E.S. da; Pereira, W.W. [Instituto de Radioproteção e Dosimetria (LNMRI/IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Laboratório Nacional de Metrologia das Radiações Ionizantes

2017-07-01

The anisotropy measurements have as main objective to define the emission of the radiation by different angles of an encapsulated neutron source. Measurements were performed using a Precision Long Counter (PLC) detector in the Laboratório de Baixo Espalhamento of the LNMRI / IRD. In this study were used an {sup 241}AmBe (α,n) 5.92 GBq and a {sup 238}PuBe (α,n) 1.85 TBq. The anisotropy factor was 8.65% to {sup 241}AmBe and 4.36% to {sup 238}PuBe, due to variations in the source encapsulation. The results in this work will focus mainly on the area of radiation protection and studies that will improve the process of routine measurements in laboratories and instrument calibrations. (author)

Neutron field features in a calibration hall

International Nuclear Information System (INIS)

Vega C, H.R.; Gallego, E.; Lorente, A.

2004-01-01

A new source facility ( 241 Am-Be) has been installed in a large size bunker-type room. To characterize the neutron fields in the facility, detailed calculations have been made with MCNP-4C, showing the different components of the neutron radiation reaching the reference points (direct, in scattered, backscattered). The contribution from neutrons scattered in the walls to the total ambient dose equivalent remains reasonably low ( 6 LiI(Eu) scintillator (0.4 cm 0 x 0.4 cm), UTA4 response matrix and BUNKIUT unfolding code. The calculated and experimentally obtained spectra are compared, with small differences found in the epithermal and thermal region, attributable to the concrete composition used in the calculations. The H*(10) rate has been determined from the spectra, and then compared to the reading of an active dosemeter (LB 6411), with differences found lower than 8%. (Author)

Intermediate-energy neutron beam for NCT at MURR

International Nuclear Information System (INIS)

Brugger, R.M.; Less, T.J.; Passmore, G.G.

1986-01-01

The University of Missouri Research Reactor (MURR) is one of the high-flux reactors in the USA and it can be used to produce an intense beam of intermediate-energy neutrons for neutron capture therapy. Two methods are being evaluated at MURR to produce such a beam. The first uses a moderator of Al 2 O 3 replacing part of the graphite and water on one side of the core of the reactor to produce a source of predominantly intermediate-energy neutrons. The second method is a filter of 238 U between the core and the patient position to pass only intermediate-energy neutrons. The results of these evaluations are presented in this paper along with an outline of the other resources at the University of Missouri-Columbia that are available to support an NCT program. 4 references, 7 figures, 1 table

RackSaver neutron absorbing device development and testing

International Nuclear Information System (INIS)

Lambert, R.; O'Leary, P.; Roberts, P.

1996-01-01

Siemens Power Corporation (SPC), in cooperation with the Electric Power Research Institute (EPRI), has developed the RackSaver neutron absorbing insert. The RackSaver insert can be installed onto spent nuclear fuel assemblies to replace deteriorating Boraflex neutron absorbing material installed in some spent-fuel storage racks. This paper describes results of a development and in-pool demonstration program performed to support potential utilization of the RackSaver neutron absorbing insert by affected utilities. The program objective was to advance the RackSaver concept into a field-demonstrated product. This objective was accomplished through three phases: design, licensing and criticality evaluations, and demonstration testing

Neutron transport from targets to moderators

International Nuclear Information System (INIS)

Taylor, A.D.

1980-01-01

The title of this meeting is 'Targets for Neutron Beam Spallation Sources', but so far all the emphasis in the talks has been on how to produce the fast neutron flux. I would like to stress that that is just the beginning of the story. What we are required to produce are beams of thermal and epithermal neutrons with time and spectral characteristics tailored to the instrumental requirements. The real source of our neutrons is not uranium arrays or thorium cylinders but a small volume of hydrogenous material, some 10 x 10 x 5 cm 3 . This is really what the whole thing is about - the target produces a copious field of fast neutrons, but if we fail to moderate them with the right energy and time characteristics, we will not match to what is happening downstream. In this talk, I am going to deal specifically with what we have done for SNS to optimise the target-moderator-reflector and decoupler system in this respect. (orig.)

The new 6 MV multi-nuclide AMS facility at the University of Tsukuba

Energy Technology Data Exchange (ETDEWEB)

Sasa, Kimikazu, E-mail: ksasa@tac.tsukuba.ac.jp; Takahashi, Tsutomu; Matsumura, Masumi; Matsunaka, Tetsuya; Satou, Yukihiko; Izumi, Daiki; Sueki, Keisuke

2015-10-15

The former accelerator mass spectrometry (AMS) system installed on the 12UD Pelletron tandem accelerator at the University of Tsukuba was completely destroyed by the Great East Japan Earthquake on 11 March 2011. A replacement has been designed and constructed at the university as part of the post-quake reconstruction project. It consists of a 6 MV Pelletron tandem accelerator, two multiple cathode AMS ion sources (MC-SNICSs), and a rare-particle detection system. The 6 MV Pelletron tandem accelerator will be applied not only to AMS, but also to areas such as nanotechnology, ion beam analysis, heavy ion irradiation, and nuclear physics. The rare-particle detection system will be capable of measuring environmental levels of long-lived radioisotopes of {sup 10}Be, {sup 14}C, {sup 26}Al, {sup 36}Cl, {sup 41}Ca, and {sup 129}I. It is also expected to measure other radioisotopes such as {sup 32}Si and {sup 90}Sr. The 6 MV Pelletron tandem accelerator was installed in the spring of 2014 at the University of Tsukuba. Routine beam delivery and AMS experiments will start in 2015.

The new 6 MV multi-nuclide AMS facility at the University of Tsukuba

International Nuclear Information System (INIS)

Sasa, Kimikazu; Takahashi, Tsutomu; Matsumura, Masumi; Matsunaka, Tetsuya; Satou, Yukihiko; Izumi, Daiki; Sueki, Keisuke

2015-01-01

The former accelerator mass spectrometry (AMS) system installed on the 12UD Pelletron tandem accelerator at the University of Tsukuba was completely destroyed by the Great East Japan Earthquake on 11 March 2011. A replacement has been designed and constructed at the university as part of the post-quake reconstruction project. It consists of a 6 MV Pelletron tandem accelerator, two multiple cathode AMS ion sources (MC-SNICSs), and a rare-particle detection system. The 6 MV Pelletron tandem accelerator will be applied not only to AMS, but also to areas such as nanotechnology, ion beam analysis, heavy ion irradiation, and nuclear physics. The rare-particle detection system will be capable of measuring environmental levels of long-lived radioisotopes of "1"0Be, "1"4C, "2"6Al, "3"6Cl, "4"1Ca, and "1"2"9I. It is also expected to measure other radioisotopes such as "3"2Si and "9"0Sr. The 6 MV Pelletron tandem accelerator was installed in the spring of 2014 at the University of Tsukuba. Routine beam delivery and AMS experiments will start in 2015.

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)

Development of a portable system to test area monitors for neutrons

International Nuclear Information System (INIS)

Souza, Luciane de Rezende

2011-02-01

The objective is to develop a portable system to test the reliability in terms of calibration of area monitors for neutrons. For the production of this system, thickness and location of the source within the system were simulated using the code of radiation transport MCNP5. The thicknesses were set for a 241 Am-Be source with an activity of 395 mCi, which will be in a polyethylene cylinder which will provide a ambient dose equivalent rate chosen through the points of calibration settings' used by the Laboratory of Neutrons (IRD / CNEN). The results obtained in this study show the feasibility of mounting the portable system as a tool to test the area monitors for neutrons, which will provide the user of neutron area monitors to check the instrument's response in the same field of operation, thus avoiding the use of an inadequate equipment. (author)

Evaluation and calculation of neutron transactinide cross-sections

International Nuclear Information System (INIS)

Konshin, V.A.

1980-01-01

This paper reviews the state of the art of nuclear theory and its application to the evaluation and calculation of neutron reaction cross sections of transactinium isotopes. In particular, the paper describes the current evaluation of the total files of neutron reaction data for 240 Pu and 241 Pu in the energy range between 10 -5 eV and 15 MeV based on a thorough analysis of available experimental data and on the use of modern theoretical concepts, and the work in progress on the evaluation of the total neutron reaction data file for 242 Pu and 241 Am. (author)

Neutron reference spectra measurements with the Bonner multi-spheres spectrometer

International Nuclear Information System (INIS)

Lemos Junior, Roberto Mendonca de

2004-01-01

This paper aims to define a procedure to use the Bonner Multisphere Spectrometer with a 6 LiI(Eu) detector in order to determine of neutron spectra. It was measured 238 PuBe spectra and same of reference ( 241 AmBe, 252 Cf e 252 Cf+D 2 O) published in ISO 8529-1 (2001) Norm. The data were processed by a computer program (BUNKI), which presents the results in neutrons energy fluency. Each input parameter of the program was studied in order to establish their influence in the adjustment result. The environment dose equivalent rate obtained placing the detector 1 m from the 241 AmBe source was 122 ± 4 μSv/h with 7% of uncertainty and 95% of confidence level. The procedure established in this work was tested with the 238 PuBe spectrum, obtaining an environment dose equivalent rate of 286 ± 9 μSv/h, 8% lower than the value measured experimentally used as reference. Through this procedure will be possible to measure neutron spectra in different work places where neutrons sources are used. Knowing these spectra, it will be possible to evaluate which area monitors, are more suitable, as well as, to study better the response of individual neutron monitors, as for instance, to obtain a conversion coefficient more appropriate to the albedo dosimeter used in different work places. As the measurements need a long time to be accomplished, the work optimization is fundamental to reduce the exposing time of the Bonner spectrometer operator. For this reason, an important parameter examined in this paper was the possibility of reducing the number of spheres used during the measurement without changing the final result. Considering the radiation protection standards, this parameter has a huge importance when the measurements are performed in work places where the neutron fluency and gamma rate offer risks to the operator's health, as for instance, in nuclear centrals. Studying this parameter, it was possible to conclude that removing the 20,32 cm diameter sphere it will be

Qualification of the monitor Pug-7N like dosimeter for neutrons; Habilitacion del monitor PUG-7N como dosimetro para neutrones

Energy Technology Data Exchange (ETDEWEB)

Benites R, J. L. [Centro Estatal de Cancerologia de Nayarit, Av. Enfermeria, Fracc. Fray Junipero Serra, 63000 Tepic, Nayarit (Mexico); Vega C, H. R.; Murillo O, R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Velazquez F, J. B., E-mail: jlbenitesr@prodigy.net.mx [Universidad Autonoma de Nayarit, Postgrado CBAP, Carretera Tepic Compostela Km. 9, Xalisco, Nayarit (Mexico)

2011-10-15

By means of an inter-comparison method, the monitor for neutrons Pug-7N was enabled like dosimeter for neutrons of two magnitudes: the environmental equivalent dose, H*(10), and the H equivalent dose. The monitor Pug-7N has a plastic detector of scintillation Pns-20 that can be used inside or outside of its polyethylene cylindrical moderator. This designed to detect the neutrons presence that is shown in ana logical form by means of a fast count. Although the instrument is useful to detect the neutrons presence its design it does not allow to estimate the dose. With the purpose of enabling it as dosimeter for neutrons, their response was compared with the response of the area monitor for neutrons Bert hold Lb 6411 and Eberline NRD model Asp-1. Under the same irradiation conditions the 3 instruments were exposed to a source of {sup 241}AmBe of 3.7E(9) Bq (100 mCi) of activity whose spectrum and dosimetric magnitudes were determined with a spectrometric system of Bonner spheres with scintillator of {sup 6}Lil(Eu) and the NSDUAZ code. Conversion factors of H*(10)/cpm and H/cpm were obtained for the two options of the monitor detector Pug-7N, with this procedure the monitor Pug-7N besides determining the presence of neutrons, it has been enabled for their use as dosimeter for neutrons. (Author)

Dual-sided microstructured semiconductor neutron detectors (DSMSNDs)

International Nuclear Information System (INIS)

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

2015-01-01

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

Neutron detection using CR-39 and Atomic Force Microscopy (AFM)

International Nuclear Information System (INIS)

Vazquez L, C.; Fragoso, R.; Felix, R.; Golzarri, J.I.; Espinosa, G.; Castillo, F.

2007-01-01

AFM has been applied in many CR-39 track formation analyses. In this paper, the use of AFM in the neutron detection and analysis of the track formation is reported. The irradiation was made with an 1.5 GBq (0.5 Ci) 241 Am-Be neutron source, with and without a polyethylene radiator. The surface analysis was made to the CR-39 fresh material without irradiation, after the irradiation, and after a very short etching time. The results show important differences between the irradiation, with and without polyethylene radiator, and the latent tracks of the neutron in the CR-39 polycarbonate. The development of track formation after very short etching time and pits characterization were measured too using the AFM facilities. (Author)

Neutron spectrum measurement in D + Be reaction

CERN Document Server

Abbasi-Davani, F; Aslani, G R; Etaati, G R; Koohi-Fayegh, R

2002-01-01

In this project the neutron spectra from the reaction of deuteron on beryllium nuclei is measured. The energies of deuterons were 7, 10, 13 and 15 MeV, and these measurements are performed at 10,30 and 50 degrees relative to the beam of deuterons. The detector used is 76 by 76 mm right circular cylinder of N E-213 liquid scintillator. The zero crossing technique is used for gamma discrimination. For the elimination of the background radiation, a Polyethylene block, 40 cm in thickness, with inserted cadmium sheets, and a lead block, 5 cm in thickness, were used. In order to obtain the background radiation spectrum, the latter blocks were placed between the target and the detector to eliminate neutron and gamma radiations reaching the detector directly. sup F ORIST sup c ode is used to unfold the neutron spectra from the measured pulse high t spectra and sup O 5S sup a nd sup R ESPMG sup c odes are used to obtain the detector response matrix.

Application of a gamma spectroscopy system to the measurement of neutron cross sections necessary to the development of nuclear energy

International Nuclear Information System (INIS)

Deruelle, O.

2002-09-01

This work concerns the development of nuclear energy and nuclear waste management in particular. Two parts of this study can be distinguished. In the first part (theoretical), a thorium-plutonium fuel based on MOX and dedicated for PWR was investigated in order to transmute plutonium in a potentially low waste fuel cycle. It was shown that this type of fuel is not regenerative but could be used for a transition to the industrial thorium fuel cycle without building new reactors. Thanks to moderated neutron spectra and high loaded actinide mass in the core, U-233 is quickly created (∼300 kg/y) for a loss of about ∼1200 kg of fissile plutonium. In the second part (experimental), we have developed and built a new reaction chamber to measure neutron cross sections of actinides by alpha-gamma spectroscopy. This experimental device (in principle transportable) was commissioned in the high flux reactor of ILL Grenoble. Neutron flux was measured by gamma spectroscopy of irradiated Al and Co samples and was found to be of the order of 6,0. 10 14 n.cm -2 .s -1 (4%). By the irradiation of 11μg of Am-243 and Pu-242, corresponding capture cross sections were measured in the thermal neutron flux at 50 deg C. These are the results: 243 Am(n,γ) 244fond. Am = 4,72±1,42b; 243 Am(n,γ) 244total Am = 74,8±3,25b; 242 Pu (n,γ) 243 Pu = 22,7±1,09b. Uncertainties of the measurements are mostly due to the determination of the neutron flux, efficiency of the electronics and ambiguities related to the definition of the area under α-γ spectra. Although our measured cross sections deviate (by 10-30%) from the corresponding values widely used in evaluated data libraries such as ENDF, JEF and JENDL, in this work we have demonstrated the feasibility and principle of our experimental method. Furthermore, the value for the 243-americium capture cross-section is in very good agreement with the last two measurements done in 1975 and 1997. These facts allowed us to think of new experiments

A technique of measuring neutron spectrum

International Nuclear Information System (INIS)

Sarkar, P.K.; Kirthi, K.N.; Ganguly, A.K.

1975-01-01

Plastic scintillators have been used to measure fast neutron spectrum from various sources. Gamma background discrimination has been done by selecting thin scintillators and thereby achieving near 100% transmission of Compton-edge electrons. The measured distribution has been unfolded by using an iterative least square technique. This gives minimum variance and maximum likelihood estimate with error minimised. Smoothening of the observed distribution has been done by Fourier and time series analyses. The method developed is applicable in principle for the determination of spectra of high energy neutrons ranging from 1 MeV to 70 MeV and beyond. However, practical application of the method is limited by the non-availability of cross-section data for various neutron induced reactions with carbon and hydrogen present in the polymerised polystyrene scintillator. This procedure has been adopted in the present work for spectral determination up to 14 MeV neutrons using the published value of reaction and scattering cross-sections. The spectra of Po-Be, Pu-Be, Am-Be and Ra-Be arrived at agree well with the published spectra obtained by other methods. Spectrum from spontaneous fission of Cf-252 have also been measured and fitted to the expression N(E)=Esup(1/2)exp(-E/T). The fitted parameter T and spectral details agree well with those in published literature

Elemental analysis technique based on detecting gamma-rays from interactions of neutrons with medium

International Nuclear Information System (INIS)

Pospisil, S.; Janout, Z.; Vobecky, M.

1979-01-01

The methods are discussed of carbon content determination in large amounts of material by detecting 4438 keV gamma radiation accompanying inelastic scattering of neutrons from a radionuclide neutron source. Presented are the methodological analysis of the problem, the results of test measurements, and methodological recommendations for the practical application of the method. Test measurements were conducted on fly ash, limestone and brown coal in amounts of approximately 5 kg for each material sample, using an Am-Be neutron source. The determined sensitivity thresholds corresponded to the carbon concentration of 5 to 10% w.w. (S.P.)

Prompt Neutron Decay Constant Determination Of Silicide Transition Core Using Noise Method

International Nuclear Information System (INIS)

Jujuratisbela, Uju; Yulianto, Yusi Eko; Cahyana

2001-01-01

Chairman of BATAN had decided to replace the Oxide fuel element type of RSG-GAS into silicide element type step by step. The replacement will create core transitions. Kinetic characteristic of the transition cores have to be monitored in order to know the deviation of core behavior. For that reason, the kinetic parameters have to be measured. Prompt neutron decay constant (alpha) is one of the kinetic parameters that has to be monitored continuously in the transition cores. In order not to disturb the normal operation of reactor, alpha parameter should be measured by using noise analysis method. The voltage of neutron flux at power of 15 MW is connected to preamplifier and filter then to the Dynamic Signal Analyzer Version-2 and then the auto power spectral density (APSD) was determined by using Fast Fourier transform. From the APSD curve of each channel of JKT03, the cut off frequency of each channel can be determined by using linear regression technique such that the prompt neutron decay constant can be estimated

Am(VI) Extraction Final Report: FY16

Energy Technology Data Exchange (ETDEWEB)

Mincher, Bruce Jay [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grimes, Travis Shane [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tillotson, Richard Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Law, Jack Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

This report summarizes activities related to hexavalent Am extraction for FY16, in completion of FCR&D Milestone M3FT-16IN030103027. Activities concentrated on three areas of research: 1) centrifugal contactor hot testing, 2) Am(VI) stability studies, and 3) alternative oxidant studies. A brief summary of each task follows. Hot Testing: A new engineering-scale oxidation and solvent extraction test bed was built at Idaho National Laboratory to allow for solvent extraction testing of minor actinide separation concepts. The test bed consists of an oxidation vessel, filtration apparatus, four, 3D printed, 2-cm diameter centrifugal contactors, feed/product vessels, and sample ports. This system replaced the previous 3 stage, 5-cm contactor test bed that was used for the initial testing in FY14. In the FY16 hot test, a feed simulant was spiked with 243Am and 139Ce and treated with 60 g/L sodium bismuthate for two hours to oxidize the Am(III) to Am(VI). This solution was then pumped through a filter and into the four-stage centrifugal contactor setup. The organic phase solvent formulation was 1 M diethylhexylbutyramide (DEHBA)/dodecane. The test showed that Am(VI) was produced by bismuthate oxidation and the residual oxidant was successfully filtered without back pressure buildup. Sixty-four percent of Am was extracted in the contactors using DEHBA. Both Am and Ce were quantitatively stripped by 0.1 M H2O2. Successful demonstration of the utility of small, printable contactors suggests that hot testing of separations concepts can now be conducted more often, since it is cheaper, generates less waste, and entails much less radcon risk than previous testing. Am(VI) stability: A rigorous examination of reagents was conducted to determine if contaminants could interfere with Am oxidation and extraction. An series of DAm measurements showed that bismuthate particle size, water source, acid quality, and DAAP batch or pre-treatment had little effect on extraction efficiency

Measurement of fast neutron induced fission cross section of minor-actinide

International Nuclear Information System (INIS)

Hirakawa, Naohiro

1997-03-01

In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am, Cm). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA is measured using Dynamitron Accelerator in Tohoku University. The experimental method and the samples, which were developed or introduced during the last year, were improved in this fiscal year: (1) Development of a sealed fission chamber, (2) Intensification of Li neutron target, (3) Improvement of time-resolution of Time-of-Flight (TOF) electronic circuit, (4) Introduction of Np237 samples with large sample mass and (5) Introduction of a U235 sample with high purity. Using these improved tools and samples, the fission cross section ratio of Np237 relative to U235 was measured between 5 to 100 keV, and the fission cross section of Np237 was deduced. On the other hand, samples of Am241 and Am243 were obtained from Japan Atomic Energy Research Institute (JAERI) after investigating fission cross section of two americium isotopes (Am241 and Am 243) which are important for core physics calculation of fast reactors. (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)

The calibration method for personal dosimetry system in photon and neutron radiation fields

Energy Technology Data Exchange (ETDEWEB)

Trousil, J; Plichta, J [CSOD, Prague (Czech Republic); Nikodemova, D [SOD, Bratislava (Slovakia)

1996-12-31

The type testing of dosimetry system was performed with standard photon radiation fields within the energy range 15 keV to 1.25 MeV and electron radiation fields within the range 0.2 MeV to 3 MeV. For type testing of neutron dosimeters {sup 252}Cf and {sup 241}Am-Be radionuclide neutron sources was used, as well as a 14 MeV neutron generator. The neutron sources moderated by various moderating and absorbing materials was also used. The routine calibration of individual photon dosemeters was carried out using a {sup 137}Cs calibration source in the air kerma quality in the dose range 0.2 mGy to 6 Gy. The type testing of neutron dosemeters was performed in collaboration with Nueherberg laboratory on neutron generator with neutron energies -.57; 1.0;; 5.3 and 15.1 MeV. The fading and angular dependence testing was also included in the tests of both dosemeter systems. (J.K.).

Characterization of the storage pool of the Neutron Standards Laboratory of CIEMAT, using Monte Carlo techniques

Energy Technology Data Exchange (ETDEWEB)

Campo B, X.; Mendez V, R.; Embid S, M. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Av. Complutense 40, 28040 Madrid (Spain); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Sanz G, J., E-mail: xandra.campo@ciemat.es [Universidad Nacional de Educacion a Distancia, Escuela Tecnica Superior de Ingenieros Industriales, C. Juan del Rosal 12, 28040 Madrid (Spain)

2014-08-15

Neutron Standards Laboratory of CIEMAT in Spain is a brand new irradiation facility, with {sup 241}Am-Be (185 GBq) and {sup 252}Cf (5 GBq) calibrated neutron sources which are stored in a water pool with a concrete cover. From this storage place an automated system is able to take the selected source and place it in the irradiation position, 4 m over the ground level and in the geometrical center of the Irradiation Room with 9 m (length) x 7.5 m (width) x 8 m (height). For calibration or irradiation purposes, detectors or materials can be placed on a bench but it is possible to use the pool (1.0 m x 1.5 m and more than 1.0 m depth) for long time irradiations in thermal neutron fields. For this reason it is essential to characterize the pool itself in terms of neutron spectrum. In this document, the main features of this facility are presented and the characterization of the storage pool in terms of neutron fluence rate and neutron spectrum has been carried out using simulations with MCNPX-2.7.e code. The MCNPX-2.7.e model has been validated using experimental measurements outside the pool (Bert hold LB6411). Inside the pool, the fluence rate decreases and the spectra is thermalized with the distance to the {sup 252}Cf source. This source predominates and the effect of the {sup 241}Am-Be source in these magnitudes is not shown until positions closer than 20 cm from it. (author)

Characterization of the storage pool of the Neutron Standards Laboratory of CIEMAT, using Monte Carlo techniques

International Nuclear Information System (INIS)

Campo B, X.; Mendez V, R.; Embid S, M.; Vega C, H. R.; Sanz G, J.

2014-08-01

Neutron Standards Laboratory of CIEMAT in Spain is a brand new irradiation facility, with 241 Am-Be (185 GBq) and 252 Cf (5 GBq) calibrated neutron sources which are stored in a water pool with a concrete cover. From this storage place an automated system is able to take the selected source and place it in the irradiation position, 4 m over the ground level and in the geometrical center of the Irradiation Room with 9 m (length) x 7.5 m (width) x 8 m (height). For calibration or irradiation purposes, detectors or materials can be placed on a bench but it is possible to use the pool (1.0 m x 1.5 m and more than 1.0 m depth) for long time irradiations in thermal neutron fields. For this reason it is essential to characterize the pool itself in terms of neutron spectrum. In this document, the main features of this facility are presented and the characterization of the storage pool in terms of neutron fluence rate and neutron spectrum has been carried out using simulations with MCNPX-2.7.e code. The MCNPX-2.7.e model has been validated using experimental measurements outside the pool (Bert hold LB6411). Inside the pool, the fluence rate decreases and the spectra is thermalized with the distance to the 252 Cf source. This source predominates and the effect of the 241 Am-Be source in these magnitudes is not shown until positions closer than 20 cm from it. (author)

Dose Calibration of the ISS-RAD Fast Neutron Detector

Science.gov (United States)

Zeitlin, C.

2015-01-01

The ISS-RAD instrument has been fabricated by Southwest Research Institute and delivered to NASA for flight to the ISS in late 2015 or early 2016. ISS-RAD is essentially two instruments that share a common interface to ISS. The two instruments are the Charged Particle Detector (CPD), which is very similar to the MSL-RAD detector on Mars, and the Fast Neutron Detector (FND), which is a boron-loaded plastic scintillator with readout optimized for the 0.5 to 10 MeV energy range. As the FND is completely new, it has been necessary to develop methodology to allow it to be used to measure the neutron dose and dose equivalent. This talk will focus on the methods developed and their implementation using calibration data obtained in quasi-monoenergetic (QMN) neutron fields at the PTB facility in Braunschweig, Germany. The QMN data allow us to determine an approximate response function, from which we estimate dose and dose equivalent contributions per detected neutron as a function of the pulse height. We refer to these as the "pSv per count" curves for dose equivalent and the "pGy per count" curves for dose. The FND is required to provide a dose equivalent measurement with an accuracy of ?10% of the known value in a calibrated AmBe field. Four variants of the analysis method were developed, corresponding to two different approximations of the pSv per count curve, and two different implementations, one for real-time analysis onboard ISS and one for ground analysis. We will show that the preferred method, when applied in either real-time or ground analysis, yields good accuracy for the AmBe field. We find that the real-time algorithm is more susceptible to chance-coincidence background than is the algorithm used in ground analysis, so that the best estimates will come from the latter.

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.

{sup 3}He Replacement for Nuclear Safeguards Applications- an integrated test program to compare alternative neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Menlove, H. O.; Henzlova, D.; Evans, L. G.; Swinhoe, M. T.; Marlow, J. B. [Los Alamos National Laboratory, Safeguards Science and Technology Group, Los Alamos, (United States)

2011-12-15

During the past several years, the demand for {sup 3}He gas has far exceeded the gas supply. This shortage of {sup 3}He gas is projected to continue into the foreseeable future. There is a need for alternative neutron detectors that do not require {sup 3}He gas. For more than four decades, neutron detection has played a fundamental role in the safeguarding and control of nuclear materials at production facilities, fabrication plants and storage sites worldwide. Neutron measurements for safeguards applications have requirements that are unique to the quantitative assay of special nuclear materials. These neutron systems measure the neutron multiplicity distributions from each spontaneous fission and/or induced fission event. The neutron time correlation counting requires that two or more neutrons from a single fission event be detected. The doubles and triples neutron counting rate depends on the detector efficiency to the 2nd and 3rd power, respectively, so low efficiency systems will not work for the coincidence measurements, and any detector instabilities are greatly amplified. In the current test program, we will measure the alternative detector properties including efficiency, die-away time, multiplicity precision, gamma sensitivity, dead-time, and we will also consider the detector properties that would allow commercial production to safeguards scale assay systems. This last step needs to be accomplished before the proposed technologies can reduce the demand on {sup 3}He gas in the safeguards world. This paper will present the methodology that includes MCNPX simulations for comparing divergent detector types such as {sup 10}B lined proportional counters with {sup 3}He gas based systems where the performance metrics focus on safeguards applications.

Investigation of neutron resonances of 247Cm in the 0.5-20 eV energy range

International Nuclear Information System (INIS)

Belanova, T.S.; Kolesov, A.G.; Klinov, A.V.; Nikol'skij, S.N.; Poruchikov, V.A.; Nefedov, V.N.; Artamonov, V.S.; Ivanov, R.N.; Kalebin, S.M.

1979-01-01

The neutron resonance parameters of 247 Cm were calculated from the transmission of a curium sample measured by the time-of-flight method. The neutron resonance parameters were calculated by the shape method using the single-level Breit-Wigner formula. Since the neutron resonance parameters of 244 Cm, 245 Cm, 246 Cm, 248 Cm, 243 Am and 240 Pu are well known, it was possible to identify the neutron resonances of 247 Cm from the measured transmission and calculate their parameters. We identified only five neutron resonances of 247 Cm with high values of 2gGAMMAsub(n). This is due to the fact that the 247 Cm content of the sample is low (1.7mg) and the resonances of this isotope are identified against the background of a large number of resonances of 244 Cm, 245 Cm, 246 Cm, 248 Cm, 243 Am and 240 Pu situated in the energy range in question

Measurement of angular distribution of neutrons emitted from plasma focus using NTD

International Nuclear Information System (INIS)

Antanasijevic, R.; Maric, Z.; Banjanac, R.; Dragic, A.; Stanojevic, J.; Dordevic, D.; Joksimovic, D.; Udovicic, V.; Vukovic, J.

1999-01-01

Mica detectors together with thick uranium foils were used for detection and angular distribution measurement of the neutrons emitted from the d-plasma focus (DPF). The distance between detectors and plasma pinch was 4 cm. For reason of detector protection from the thermal shock, they were covered with the mica layer of 1 mm thickness. Annealing of the detectors were measured after the irradiation with neutrons from an Am-Be source and treated with thermal shock H-plasma focus (HPF). Geometry of detectors in both experiments was the same

Measurement of angular distribution of neutrons emitted from plasma focus using NTD

Energy Technology Data Exchange (ETDEWEB)

Antanasijevic, R.; Maric, Z.; Banjanac, R.; Dragic, A.; Stanojevic, J.; Dordevic, D.; Joksimovic, D.; Udovicic, V.; Vukovic, J

1999-06-01

Mica detectors together with thick uranium foils were used for detection and angular distribution measurement of the neutrons emitted from the d-plasma focus (DPF). The distance between detectors and plasma pinch was 4 cm. For reason of detector protection from the thermal shock, they were covered with the mica layer of 1 mm thickness. Annealing of the detectors were measured after the irradiation with neutrons from an Am-Be source and treated with thermal shock H-plasma focus (HPF). Geometry of detectors in both experiments was the same.

Measurement of angular distribution of neutrons emitted from plasma focus using NTD

CERN Document Server

Antanasijevic, R; Banjanac, R; Dragic, A; Stanojevic, J; Dordevic, D; Joksimovic, D; Udovicic, V; Vukovic, J

1999-01-01

Mica detectors together with thick uranium foils were used for detection and angular distribution measurement of the neutrons emitted from the d-plasma focus (DPF). The distance between detectors and plasma pinch was 4 cm. For reason of detector protection from the thermal shock, they were covered with the mica layer of 1 mm thickness. Annealing of the detectors were measured after the irradiation with neutrons from an Am-Be source and treated with thermal shock H-plasma focus (HPF). Geometry of detectors in both experiments was the same.

The fast neutron SEU cross section of a 4 Mb SRAM memory

International Nuclear Information System (INIS)

Pereira Junior, Evaldo C.F.; Goncalez, Odair L.; Cruz, Marco Aurelio da; Prado, Adriane Cristina Mendes; Federico, Claudio Antonio; Gaspar, Felipe de Barros

2013-01-01

The results of a static test of single event upset (SEU) produced by fast neutrons on an ISSI 4Mb SRAM memory are reported in this work. To perform the tests, it was built a platform based on a motherboard which is controlled by microprocessor, whose function is to perform the writing, reading and control of the memories under irradiation. The irradiation was performed with a set of 8 241 Am-Be neutrons source in a quasi-isotropic incidence. The SEU cross was calculated from the accumulated bit flip count. (author)

Embedded data acquisition system for neutron monitors

International Nuclear Information System (INIS)

Población, Ó G; Tejedor, I G; Sánchez, S; Blanco, J J; Gómez-Herrero, R; Medina, J; Steigies, C T

2014-01-01

This article presents the design and implementation of a new data acquisition system to be used as replacement for the old ones that have been in use with neutron monitors for the last decades and, which are eventually becoming obsolete. This new system is also intended to be used in new installations, enabling these scientific instruments to use today's communication networks to send data and receive commands from the operators. This system is currently running in two stations: KIEL2, in the Christian-Albrechts-Universität zu Kiel, Kiel, Germany, and CALMA, in the Castilla-La Mancha Neutron Monitor, Guadalajara, Spain

Measurements of H*(10) in reference neutron fields using Bonner sphere spectrometry and LET spectrometry

CERN Document Server

Golnik, N; Králik, M

2002-01-01

A Bonner sphere spectrometer and the REM-2 recombination chamber were used for inter-comparison measurements of the neutron component of ambient dose equivalent, H sub n *(10) in reference neutron fields. The sup 2 sup 4 sup 1 Am-Be and sup 2 sup 5 sup 2 Cf neutron sources were exposed either free-in-air or placed in iron or paraffin filters. The REM-2 recombination chamber was used as a LET spectrometer. The agreement of H sub n *(10) values measured with both the methods was within experimental uncertainties of few percent. The determined neutron spectra were used for calculations of the REM-2 chamber response to H*(10).

Single-sphere multiple-detector neutron spectrometer. Final report on Phase 1

International Nuclear Information System (INIS)

Sinclair, F.; Stern, I.; Hahn, R.W.; Entine, G.

1987-07-01

To address the problem of accurate, timely estimates of the neutron spectral flux, researchers are developing a monitoring instrument based on a single moderating sphere with a large number of independent sensors. Such a single-sphere spectrometer would allow easy measurement of quality factors. This is made possible by the recent development of a novel digital sensor which detects radiation induced errors in a dynamic random-access memory. During Phase I of the SBIR program, researchers constructed a first prototype of the single-sphere spectrometer, measured its response in a neutron flux from an isotopic Am-Be source in several geometries, and compared these with the results of Monte Carlo simulations of neutron transport. The preliminary results show that the approach is feasible and relatively straightforward

Calibration of a spectrometry multisphere system for neutron fields

International Nuclear Information System (INIS)

Carelli, Jorge L.; Cruzate, Juan A.; Papadopulos, Susana B.; Gregori, Beatriz N.; Ciocci Brazzano, Ligia

2005-01-01

In this work it is presented the calibration of the neutrons spectrometric system of the Nuclear Regulatory Authority (ARN) in the Institut de Protection et Sure te Nucleaires (Ipn), Labourite dadaist et de Recherche s en Dosimetric Extern e, Cadarache, France. The multisphere system is composed of 9 polyethylene spheres of high density, with a gaseous detector of 3 He and associate electronics. The matrix of energy response to the system neutrons was obtained applying the MCNPX code for the range of energies between thermal and 100 MeV with cross sections taken from library ENDF/B-VI. The neutron spectra of the multisphere system were obtained applying the deconvolution code LOUHI82. The relationship between the theoretical responses and the experiences obtained with the AmBe and 252 Cf sources are also presented in this work [es

Neutron beam imaging with GEM detectors

International Nuclear Information System (INIS)

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

2015-01-01

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

Twelve years of follow up of cases with old 241Am internal contamination.

Science.gov (United States)

Malátová, Irena; Vrba, Tomás; Becková, Vera; Pospísilová, Helena

2010-10-01

A group of workers internally contaminated with Am have been followed for about 12 years. The source of contamination was AmO2 powder used for production of AmBe neutron sources and other applications. The production of some radionuclide sources included chemical treatment of the original material, which transformed the americium into the nitrate, but mostly powder metallurgy was used for production of sources for smoke detectors. In vivo measurement of the workers was performed with two LEGe detectors placed near the head of the measured person. Calibration was performed with four different physical skull phantoms of different origin and a voxel phantom with Monte Carlo simulation, which was developed to fit the head sizes of individual persons. Samples of urine and feces were analyzed by means of radiochemical separation followed by alpha-spectrometry. Separation of 241Am from mineralized excreta was performed by combined anion exchange and extraction chromatographic techniques. As a tracer, 243Am was used. When the measured data (83 data on skeletal activity, activity in 389 bioassay samples) were compared with International Commission on Radiological Protection's and Leggett's biokinetic models of americium, it was found that in most cases, after more than 15 y since the intake, the excretion rate was lower (or skeletal activity higher) than predicted. On the other hand, the ratio of excreted activity in urine and feces agrees well with model predictions.

Calculations of radiation damage in target, container and window materials for spallation neutron sources

International Nuclear Information System (INIS)

Wechsler, M.S.; Mansur, L.K.

1996-01-01

Radiation damage in target, container, and window materials for spallation neutron sources is am important factor in the design of target stations for accelerator-driver transmutation technologies. Calculations are described that use the LAHET and SPECTER codes to obtain displacement and helium production rates in tungsten, 316 stainless steel, and Inconel 718, which are major target, container, and window materials, respectively. Results are compared for the three materials, based on neutron spectra for NSNS and ATW spallation neutron sources, where the neutron fluxes are normalized to give the same flux of neutrons of all energies

Fast neutron forward distributions from C, Be and U thick targets bombarded by deuterons

International Nuclear Information System (INIS)

Menard, S.; Clapier, F.; Pauwels, N.; Proust, J.; Donzaud, C.; Guillemaud-Mueller, D.; Lhenry, I.; Mueller, A.C.; Scarpaci, J.A.; Sorlin, O.; Mirea, M.

1999-01-01

In principle, to produce neutron rich radioactive beams with sufficient intensities, a source of isotopes far from the valley of β--stability can be obtained through the fission of 238 U induced by fast neutrons. A very promising way to assess the feasibility of these very intense neutron beams is to break an intense 2 H beam in a dedicated converter. The main objective of SPIRAL and PARRNe R - D projects is the investigation of the optimum parameters for a neutron rich isotope source in accordance with the scheme presented above. In such conditions, the charge particle energy loss can prevent the destruction of the fission target. In the frame of these project, a special attention is dedicated to the energetic and angular distributions of the neutrons emerging from a set of converters at a series of 2 H incident energies. Deuteron beams at energies less than 30 MeV are particularly interesting because it is expected that, after the decay in the 238 U target, the neutron rich radioactive fission products are cold enough, thus avoiding the evaporation of a too large number of neutrons. For such purposes, one needs experimental angular distributions at given energies for different types of converters and to elaborate a theoretical tool in order to estimate accurately the characteristics of the secondary neutron beam. In this paper, the experimental results were obtained with 17, 20 and 28 MeV deuteron energies on Be, C and U converters using the time of flight method. These data are compared to results given by a model valid at higher energy in order to obtain pertinent simulations in a large range of incident energies. Many theoretical tools were developed to characterize the properties of the neutron beams emerging from thick targets. In this contribution the Serber's model, considered with its improvements which account for the Coulomb deflection and the mean straggling of the beam in the material, is compared to experimental data in order to verify the validity

Neutron scattering equipments in JAERI. Current status

International Nuclear Information System (INIS)

Hamaguchi, Yoshikazu; Minakawa, Nobuaki

2003-01-01

24 neutron scattering instruments are installed in the JRR-3M research reactor. Among them JAERI has 12 neutron scattering instruments. Those instruments are HRPD for high-resolution structural analysis, TAS-1 and TAS-2 for elastic and inelastic scattering and for magnetic scattering measurements by the polarized neutron, LTAS for elastic and inelastic scattering measurement at a low energy region, and for neutron device development, PNO for topography and for very small angle scattering measurement in a small Q range, NRG for neutron radiography, RESA for internal strain measurements, SANS for the molecule and semi-macroscopic magnetic structural analysis, BIX-2 and BIX-3 for the biological structural analysis research, and PGA for the research of prompt gamma-ray analysis. The university groups have 12 neutron scattering instruments. Since those instruments were installed at the period when JRR-3M was completed, about 10 years have passed. In order to match the old control systems with the progress of recent computer technologies, and peripheral equipment, numbers of instruments are being renewed. In the neutron guide hall of JRR-3M, the Ni mirror guide tube was replaced by a super mirror guide tube to increase neutron flux. The intensity of 2A flux was increased by a factor of about two. (J.P.N.)

Development of a position-sensitive fission counter and measurement of neutron flux distributions

International Nuclear Information System (INIS)

Yamagishi, Hideshi; Soyama, Kazuhiko; Kakuta, Tsunemi

2001-08-01

A position-sensitive fission counter (PSFC) that operates in high neutron flux and high gamma-ray background such as at the side of a power reactor vessel has been developed. Neutron detection using the PSFC with a solenoid electrode is based on a delay-line method. The PSFC that has the outer diameter of 25 mm and the sensitive length of 1000 mm was manufactured for investigation of the performances. The PSFC provided output current pulses that were sufficiently higher than the alpha noise, though the PSFC has a solenoid electrode and large electrode-capacitance. The S/N ratio of PSFC outputs proved to be higher than that of ordinary fission counters with 200 mm sensitive length. A performance test to measure neutron flux distributions by a neutron measuring system with the PSFC was carried out by the side of a graphite pile, W2.4 x H1.4 x L1.2 m, with neutron sources, Am-Be 370 GBq x 2. It was confirmed that the neutron flux distribution was well measured with the system. (author)

Preliminary study about frequencies of unstable chromosome alterations induced by gamma beam and neutron-gamma mixed field

International Nuclear Information System (INIS)

Mendes, Mariana E.; Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F.; Calixto, Merilane S.; Santos, Neide

2011-01-01

The estimate on approximate dose in exposed individual can be made through conventional cytogenetic analysis of dicentric, this technique has been used to support physical dosimetry. It is important to estimate the absorbed dose in case of accidents with the aim of developing an appropriate treatment and biological dosimetry can be very useful in case where the dosimetry is unavailable. Exposure to gamma and neutron radiation leads to the same biological effects such as chromosomal alterations and cancer. However, neutrons cause more genetic damage, such as mutation or more structural damage, such as chromosome alterations. The aim of research is to compare frequencies of unstable chromosome alterations induced by a gamma beam with those from neutron-gamma mixed field. Two blood samples were obtained from one healthy donor and irradiated at different sources. The first sample was exposed to mixed field neutron-gamma sources 241 AmBe at the Neutron Calibration Laboratory (NCL - CRCN/NE - PE - Brazil) and the second one was exposed to 137 Cs gamma rays at 137 Cs Laboratory (CRCN/NE - PE - Brazil), both exposures resulting in an absorbed dose of 0.66Gy. Mitotic metaphase cells were obtained by lymphocyte culture for chromosomal analysis and slides were stained with Giemsa 5%. These preliminary results showed a similarity in associated dicentrics frequency per cell (0.041 and 0.048) after 137 Cs and 241 AmBe sources irradiations, respectively. However, it was not observed centric rings frequency per cell (0.0 and 0.027). This study will be continue to verify the frequencies of unstable chromosome alterations induced by only gamma beam and neutron-gamma mixed field. (author)

Preliminary study about frequencies of unstable chromosome alterations induced by gamma beam and neutron-gamma mixed field

Energy Technology Data Exchange (ETDEWEB)

Mendes, Mariana E.; Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F. [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Calixto, Merilane S.; Santos, Neide [Universidade Federal de Pernanmbuco (CCB/UFPE), Recife, PE (Brazil). Centro de Ciencias Biologicas. Dept. de Genetica

2011-07-01

The estimate on approximate dose in exposed individual can be made through conventional cytogenetic analysis of dicentric, this technique has been used to support physical dosimetry. It is important to estimate the absorbed dose in case of accidents with the aim of developing an appropriate treatment and biological dosimetry can be very useful in case where the dosimetry is unavailable. Exposure to gamma and neutron radiation leads to the same biological effects such as chromosomal alterations and cancer. However, neutrons cause more genetic damage, such as mutation or more structural damage, such as chromosome alterations. The aim of research is to compare frequencies of unstable chromosome alterations induced by a gamma beam with those from neutron-gamma mixed field. Two blood samples were obtained from one healthy donor and irradiated at different sources. The first sample was exposed to mixed field neutron-gamma sources {sup 241}AmBe at the Neutron Calibration Laboratory (NCL - CRCN/NE - PE - Brazil) and the second one was exposed to {sup 137}Cs gamma rays at {sup 137}Cs Laboratory (CRCN/NE - PE - Brazil), both exposures resulting in an absorbed dose of 0.66Gy. Mitotic metaphase cells were obtained by lymphocyte culture for chromosomal analysis and slides were stained with Giemsa 5%. These preliminary results showed a similarity in associated dicentrics frequency per cell (0.041 and 0.048) after {sup 137}Cs and {sup 241}AmBe sources irradiations, respectively. However, it was not observed centric rings frequency per cell (0.0 and 0.027). This study will be continue to verify the frequencies of unstable chromosome alterations induced by only gamma beam and neutron-gamma mixed field. (author)

An accurate and portable solid state neutron rem meter

Energy Technology Data Exchange (ETDEWEB)

Oakes, T.M. [Nuclear Science and Engineering Institute, University of Missouri, Columbia, MO (United States); Bellinger, S.L. [Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS (United States); Miller, W.H. [Nuclear Science and Engineering Institute, University of Missouri, Columbia, MO (United States); Missouri University Research Reactor, Columbia, MO (United States); Myers, E.R. [Department of Physics, University of Missouri, Kansas City, MO (United States); Fronk, R.G.; Cooper, B.W [Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS (United States); Sobering, T.J. [Electronics Design Laboratory, Kansas State University, KS (United States); Scott, P.R. [Department of Physics, University of Missouri, Kansas City, MO (United States); Ugorowski, P.; McGregor, D.S; Shultis, J.K. [Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS (United States); Caruso, A.N., E-mail: carusoan@umkc.edu [Department of Physics, University of Missouri, Kansas City, MO (United States)

2013-08-11

Accurately resolving the ambient neutron dose equivalent spanning the thermal to 15 MeV energy range with a single configuration and lightweight instrument is desirable. This paper presents the design of a portable, high intrinsic efficiency, and accurate neutron rem meter whose energy-dependent response is electronically adjusted to a chosen neutron dose equivalent standard. The instrument may be classified as a moderating type neutron spectrometer, based on an adaptation to the classical Bonner sphere and position sensitive long counter, which, simultaneously counts thermalized neutrons by high thermal efficiency solid state neutron detectors. The use of multiple detectors and moderator arranged along an axis of symmetry (e.g., long axis of a cylinder) with known neutron-slowing properties allows for the construction of a linear combination of responses that approximate the ambient neutron dose equivalent. Variations on the detector configuration are investigated via Monte Carlo N-Particle simulations to minimize the total instrument mass while maintaining acceptable response accuracy—a dose error less than 15% for bare {sup 252}Cf, bare AmBe, an epi-thermal and mixed monoenergetic sources is found at less than 4.5 kg moderator mass in all studied cases. A comparison of the energy dependent dose equivalent response and resultant energy dependent dose equivalent error of the present dosimeter to commercially-available portable rem meters and the prior art are presented. Finally, the present design is assessed by comparison of the simulated output resulting from applications of several known neutron sources and dose rates.

A transportable system for the determination of phosphorus in sheep bone by in vivo neutron activation analysis

International Nuclear Information System (INIS)

Whineray, S.; Thomas, B.J.; Ternouth, J.H.; Davies, H.M.S.

1980-01-01

An apparatus was constructed which measures the phosphorus in sheep leg bone, non-invasively, by neutron activation analysis. The results obtained show that with two 10 Ci isotopic neutron sources ( 241 Am/Be or 238 Pu/Be) and a single 7.5 x 7.5 cm NaI(Tl) detector, serial changes in leg bone phosphorus may be determined with a precision of 13% in 15 min of experimental time. This precision could be reduced to 5% by incorporating two large detectors into the system. (author)

Determination of Am-241 in lung and bone by gamma spectrometry with semiconductor detectors LEGe

International Nuclear Information System (INIS)

Perez Lopez, B.

2014-01-01

Americium is produced from neutron absorption plutonium atoms within nuclear reactors. The work of dismantling and decontamination of the installations and radioactive waste management makes workers exposed acquire risk of internal exposure and therefore can incorporate Am-241 in his body. (Author)

Critical mass calculations for 241Am, 242mAm and 243Am

International Nuclear Information System (INIS)

Dias, Hemanth; Tancock, Nigel; Clayton, Angela

2003-01-01

Criticality mass calculations are reported for 241 Am, 242m Am and 243 Am using the MONK and MCNP computer codes with the UKNDL, JEF-2.2, ENDF/B-VI and JENDL-3.2 nuclear data libraries. Results are reported for spheres of americium metal and dioxide in bare, water reflected and steel reflected systems. Comparison of results led to the identification of a serious inconsistency in the 241 Am ENDF/B-VI DICE library used by MONK - this demonstrates the importance of using different codes to verify critical mass calculations. The 241 Am critical mass estimates obtained using UKNDL and ENDF/B-VI show good agreement with experimentally inferred data, whilst both JEF-2.2 and JENDL-3.2 produce higher estimates of critical mass. The computed critical mass estimates for 242m Am obtained using ENDF/B-VI are lower than the results produced using the other nuclear data libraries - the ENDF/B-VI fission cross-section for 242m Am is significantly higher than the other evaluations in the fast region and is not supported by recent experimental data. There is wide variation in the computed 243 Am critical mass estimates suggesting that there is still considerable uncertainty in the 243 Am nuclear data. (author)

AMS ready for launch

CERN Multimedia

Katarina Anthony

2011-01-01

On 29 April, the Alpha Magnetic Spectrometer (AMS) will complete its long expedition to the International Space Station on board the space shuttle Endeavour. The Endeavour is set to lift off from NASA’s Kennedy Space Station at 15:47 EST (21:47 CET).   Samuel Ting, principal investigator for the AMS project, and Rolf Heuer, CERN Director-General, visit the Kennedy Space Centre before the AMS launch.  Courtesy of NASA and Kennedy Space Center. AMS is a CERN recognised experiment, created by an internal collaboration of 56 institutes. It will be the first large magnetic spectrometer to be used in space, and has been designed to function as an external module on the ISS. AMS will measure cosmic rays without atmospheric interference, allowing researchers on the ground to continue their search for dark matter and antimatter in the Universe. Data collected by AMS will be analysed in CERN’s new AMS Control Centre in Building 946 (due for completion in June 2011). The End...

The spectrometer PERKEO III and the decay of free neutrons; Das Spektrometer PERKEO III und der Zerfall des freien Neutrons

Energy Technology Data Exchange (ETDEWEB)

Maerkisch, B.M.

2006-10-18

The Standard Model of particle physics describes neutron decay with three parameters. In contrast to that, a multitude of observables are accessible experimentally. With precision measurements sensitive tests of the Standard Model are thus possible. The neutron decay spectrometer PERKEO III was designed and built as part of this work. It will replace its predecessor PERKEO II, which has reached its systematical and statistical limits. With the new instrument measurements with continuous and pulsed neutron beams become feasible. This either provides an increase in statistics of up to two orders of magnitude, or eliminates the two major instrument specific sources of systematical corrections. In our first measurement in winter 2006/2007, the available event rate will be used to determine weak magnetism from the electron asymmetry A. Previously, this value was not statistically accessible in neutron decay. Systematics are analyzed with the help of our measurement with PERKEO II. For this measurement PERKEO III will be installed at the neutron guide H113 at the Institute Laue-Langevin, Grenoble. The neutron beam from this guide is characterized and a model is given, which allows the rapid calculation of beam profiles and absolute event rates from such a beam. In preparation of a future neutron decay instrument the reflective properties of two non-magnetic neutron mirrors were measured. (orig.)

High-resolution neutron diffraction studies of biological and industrial fibres

Energy Technology Data Exchange (ETDEWEB)

Langan, P; Mason, S A [Institut Max von Laue - Paul Langevin (ILL), 38 - Grenoble (France); Fuller, W; Forsyth, V T; Mahendrasingam, A; Shotton, M; Simpson, L [Keele Univ. (United Kingdom); Grimm, H [FZ, Juelich (Germany); Leberman, R [EMBL, (Country Unknown)

1997-04-01

Neutron diffraction is becoming an important tool for studying fibres due to its complementarity to X-ray diffraction. Unlike X-rays, scattering of neutrons by polymer atoms is not a function of their atomic number. In high-resolution studies (1.5-3 A) on D19 deuteration (replacing H by D) is being used to change the relative scattering power of chosen groups making them easier to locate. Recent studies on DNA and cellulose are described. (author). 6 refs.

Modification of Prototype D-D Neutron Generator

International Nuclear Information System (INIS)

Kim, In Jung; Kim, Suk Kwon; Park, Chang Su; Jung, Nam Suk; Jung, Hwa Dong; Park, Ji Young; Hwang, Yong Seok; Choi, H. D.

2005-01-01

The prototype D-D neutron generator was modified in order to enhance the neutron yield. The distance from ion source to target was reduced to increase the ion beam current at target position. Thick Ti target was replaced by thin Ti target which was vacuum-deposited on Cu substrate in order to enhance the target cooling. Performance of the modified device was tested

Evaluation of energy responses for neutron dose-equivalent meters made in Japan

International Nuclear Information System (INIS)

Saegusa, J.; Yoshizawa, M.; Tanimura, Y.; Yoshida, M.; Yamano, T.; Nakaoka, H.

2004-01-01

Energy responses of three types of Japanese neutron dose-equivalent (DE) meters were evaluated by Monte Carlo simulations and measurements. The energy responses were evaluated for thermal neutrons, monoenergetic neutrons with energies up to 15.2 MeV, and also for neutrons from such radionuclide sources as 252 Cf and 241 Am-Be. The calculated results were corroborated with the measured ones. The angular dependence of the response and the DE response were also evaluated. As a result, reliable energy responses were obtained by careful simulations of the proportional counter, moderator and absorber of the DE meters. Furthermore, the relationship between pressure of counting gas and response of the DE meter was discussed. By using the obtained responses, relations between predicted readings of the DE meters and true DE values were studied for various workplace spectra

Diffusion Parameters of BeO by the Pulsed Neutron Method

International Nuclear Information System (INIS)

Joshi, B.V.; Nargundkar, V.R.; Subbarao, K.

1965-01-01

The use of the pulsed neutron method for the precise determination of the diffusion parameters of moderators is described. The diffusion parameters of BeO have been obtained by this method. The neutron bursts were produced from a cascade accelerator by pulsing the ion source and using the Be (d, n) reaction. The detector was an enriched boron trifluoride proportional counter. It is shown that by a proper choice of the counter position arid length, and the source position, most of the space harmonics can be eliminated. Any constant background can be accounted for in the calculation of the decay constant. Very large bucklings were not used to avoid time harmonics. Any remaining harmonic content was rendered ineffective by the use of adequate time delay. The decay constant of the fundamental mode of the thermal neutron population was determined for several bucklings. Conditions to be satisfied for an accurate determination of the diffusion cooling constant C are discussed. The following values are obtained for BeO: λ 0 = absorption constant = 156.02 ± 4.37 s -1 D = diffusion coefficient = (1.3334 ± 0.0128) x 10 5 cm 2 /s C = diffusion cooling constant = (-4.8758 ± 0.5846) x 10 5 cm 4 /s. The effect of neglecting the contribution of the B 6 term on the determination of the diffusion parameters was estimated and is shown to be considerable. The reason for the longstanding discrepancy between the values of C obtained for the same moderator by different workers is attributed to this. (author) [fr

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

Solid thermoluminescent dosemeter of sodium tetraborate and brazilian fluoride sensitive to thermal neutrons

International Nuclear Information System (INIS)

Fratin, L.

1988-01-01

The techniques of compacting sodium tetraborate and natural fluoride mixtures were studied in this work, with the aim of producing a solid dosimeter sensitive to thermal neutrons. The production procedure involves the vitrification of the sodium tetraborate, the grinding, mixture, cold pressing and the sinterization of the pellets. A special arrangement was built for irradiation where paraffin was used as moderator for neutrons from a 241 Am-Be source. Two different mass ratios of sodium tetraborate and flourite showed a linear thermoluminescent response to the neutron fluence in the range of 1.0 to 7.0 x 10 8 n (sub)tcm -2 . Solid dosimeters, manufactured from natural fluorite and sodium chloride, showed a response to gamma radiation similar to the response of the dosimeters sensitive to neutrons. These dosimeters are need to identify the proportion of thermoluminescent response due to gamma radiation present in a neutron field. (author) [pt

Study of the response of a silicon detector irradiated with 1 MeV neutrons; Etude de la reponse d`un detecteur Si irradie par des neutrons de 1 MeV

Energy Technology Data Exchange (ETDEWEB)

Roy, P [Montreal Univ., PQ (Canada). Lab. de Physique Nucleaire

1994-12-31

The author studied the response of an n-type silicon detector irradiated with 1 MeV neutrons at fluences ranging from 0.26x10{sup 13} to 11.19x10{sup 13} neutrons/cm{sup 2}. The response of the irradiated detector to {sup 241}Am alpha particles was measured. 13 refs., 7 figs.

Activation Measurements for Thermal Neutrons, Part D. U.S. Measurements of 36Cl in Mineral Samples from Hiroshima and Nagasaki

International Nuclear Information System (INIS)

Tore Straume; Alfredo A, Marchetti; Stephen D, Egbert; James A, Roberts; Ping Men; Shoichiro Fujita; Kiyoshi Shizuma; Masaharu Hoshi; G, Rugel; W, Ruhm; G, Korschinek; J. E. McAninch; K. L. Carroll; T. Faestermann; K. Knie; R. E. Martinelli; A. Wallner; C. Wallner

2005-01-01

The present paper presents the 36 Cl measurement effort in the US. A large number of 36 Cl measurements have been made in both granite and concrete samples obtained from various locations and distances in Hiroshima and Nagasaki. These measurements employed accelerator mass spectrometry (AMS) to quantify the number of atoms of 36 Cl per atom of total Cl in the sample. Results from these measurements are presented here and discussed in the context of the DS02 dosimetry reevaluation effort for Hiroshima and Nagasaki atomic-bomb survivors. The production of 36 Cl by bomb neutrons in mineral samples from Hiroshima and Nagasaki was primarily via the reaction 35 Cl(n,γ) 36 Cl. This reaction has a substantial thermal neutron cross-section (43.6 b at 0.025 eV) and the product has a long half-life (301,000 y). hence, it is well suited for neutron-activation detection in Hiroshima and Nagasaki using AMS more than 50 years after the bombings. A less important reaction for bomb neutrons, 39 K(n,α) 36 Cl, typically produces less than 10% of the 36 Cl in mineral samples such as granite and concrete, which contain ∼ 2% potassium. In 1988, only a year after the publication of the DS86 final report (Roesch 1987), it was demonstrated experimentally that 36 Cl measured using AMS should be able to detect the thermal neutron fluences at the large distances most relevant to the A-bomb survivor dosimetry. Subsequent measurements in mineral samples from both Hiroshima and Nagasaki validated the experimental findings. The potential utility of 36 Cl as a thermal neutron detector in Hiroshima was first presented by Haberstock et al. who employed the Munich AMS facility to measure 36 Cl/Cl ratios in a gravestone from near the hypocenter. That work subsequently resulted in an expanded 36 Cl effort in Germany that paralleled the US work. More recently, there have also been 36 Cl measurements made by a Japanese group. The impetus for the extensive 36 Cl and other neutron activation

Optimization of CR-39 for fast neutron dosimetry applications

CERN Document Server

Vilela, E; Giacomelli, G; Giorgini, M; Morelli, B; Patrizii, L; Serra, P; Togo, V

1999-01-01

We present the results of an experimental work aimed at improving the performances of the CR-39[reg] (Registered Trademark of PPG Industries Inc.) nuclear track detector for neutron dosimetry applications. The work was done in collaboration with the Intercast Europe S.p.A., producer of CR-39 for commercial and scientific applications. We compare the CR-39 made with different additives concentrations and different polymerization processes. We evaluate the response of the CR-39 to fast neutrons from three sources: sup 2 sup 4 sup 1 Am-Be, sup 2 sup 5 sup 2 Cf and sup 2 sup 3 sup 8 Pu-Li. Particular attention was paid to background fluctuations that limit the lower detectable dose.

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.

Validation of neutron data libraries by backscattered spectra of Pu-Be Neutrons

CERN Document Server

El-Agib, I

1999-01-01

Elastically backscattered spectra of Pu-Be neutrons have been measured for SiO sub 2 , water, graphite, paraffin oil and Al slabs using a proton recoil spectrometer. The results were compared with the calculated spectra obtained by the three-dimensional Monte-Carlo transport code MCNP-4B and point-wise cross sections from the ENDF/B-V, ENDF/B-VI, JENDL-3.1 and BROND-2 data libraries. The good agreement between the measured and calculated results indicates that this procedure can be used for validation of different data libraries. This simple method renders possible the detection of oxygen, carbon and hydrogen in bulk samples. (author)

Search for EC-decayed neutron-deficient actinide isotopes using gas-jet coupled JAERI-ISOL

Energy Technology Data Exchange (ETDEWEB)

Tsukada, Kazuaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-07-01

To study the nuclear properties of unknown neutron deficient actinide isotopes which decay mainly via orbital electron capture (EC), we have developed a composite system consisting of a gas-jet transport apparatus and a thermal ion-source at the JAERI-ISOL. With this system, search for {sup 236}Am produced in the {sup 235}U({sup 6}Li, 5n) reaction has been performed. Pu KX-rays associated with the EC decay of {sup 236}Am are observed at the mass-236 fraction. The half-life of {sup 236}Am is evaluated to be 4.4min. The outline of the gas-jet coupled JAERI-ISOL system and typical performance are given. (author)

AM Envelope. The potential of Additive Manufacturing for facade constructions

Directory of Open Access Journals (Sweden)

Holger Strauss

2017-11-01

and enhanced performance. Advancements can for example be achieved in the semi-finished goods: more effective glueing of window frames can be supported by Snap-On fittings. Solving the most critical part of a free-form structure and allowing for a smart combination with the approved standards has a great potential, as well. Next to those product oriented approaches toward future envelopes, this thesis provides the basic knowledge about AM technologies and AM materials. The basic principle of AM opens a fascinating new world of engineering, no matter what applications can be found: to ‘design for function’ rather to ‘design for production’ turns our way of engineering of the last century upside down. A collection of AM applications therefore offers the outlook to our (built future in combination with the acquired knowledge. AM will never replace established production processes but rather complement them where this seems practical. AM is not the proverbial Swiss-army knife that can resolve all of today’s façade issues! But it is a tool that might be able to close another link in the ‘file-to-factory chain’. AM allows us a better, more precise and safer realization of today’s predominantly free designs that are based on the algorithms of the available software. With such extraordinary building projects, the digital production of neuralgic system components will become reality in the near future – today, an AM Envelope is close at hand. Still, ‘printing’ entire buildings lies in the far future; for a long time human skill and craftsmanship will be needed on the construction site combined with high-tech tools to translate the designers’ visions into reality. AM Envelope is one possible result of this!

Analysis of unstable chromosome alterations frequency induced by neutron-gamma mixed field radiation

Energy Technology Data Exchange (ETDEWEB)

Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Vale, Carlos H.F.P.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F. [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)], e-mail: psouza@cnen.gov.br, e-mail: jodinilson@cnen.gov.br; Calixto, Merilane S.; Santos, Neide [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Genetica

2009-07-01

Nowadays monitoring chromosome alterations in peripheral blood lymphocytes have been used to access the radiation absorbed dose in individuals exposed accidental or occupationally to gamma radiation. However there are not many studies based on the effects of mixed field neutron-gamma. The radiobiology of neutrons has great importance because in nuclear factories worldwide there are several hundred thousand individuals monitored as potentially receiving doses of neutron. In this paper it was observed the frequencies of unstable chromosome alterations induced by a gamma-neutron mixed field. Blood was obtained from one healthy donor and exposed to mixed field neutron-gamma sources {sup 241}AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphases were analyzed for the presence of chromosome alterations by two experienced scorers. The results suggest that there is the possibility of a directly proportional relationship between absorbed dose of neutron-gamma mixed field radiation and the frequency of unstable chromosome alterations analyzed in this paper. (author)

Analysis of unstable chromosome alterations frequency induced by neutron-gamma mixed field radiation

International Nuclear Information System (INIS)

Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Vale, Carlos H.F.P.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F.; Calixto, Merilane S.; Santos, Neide

2009-01-01

Nowadays monitoring chromosome alterations in peripheral blood lymphocytes have been used to access the radiation absorbed dose in individuals exposed accidental or occupationally to gamma radiation. However there are not many studies based on the effects of mixed field neutron-gamma. The radiobiology of neutrons has great importance because in nuclear factories worldwide there are several hundred thousand individuals monitored as potentially receiving doses of neutron. In this paper it was observed the frequencies of unstable chromosome alterations induced by a gamma-neutron mixed field. Blood was obtained from one healthy donor and exposed to mixed field neutron-gamma sources 241 AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphases were analyzed for the presence of chromosome alterations by two experienced scorers. The results suggest that there is the possibility of a directly proportional relationship between absorbed dose of neutron-gamma mixed field radiation and the frequency of unstable chromosome alterations analyzed in this paper. (author)

Determination of arsenic in some Myanmar indigenous medicines by neutron activation analysis

International Nuclear Information System (INIS)

Myint, U.; Than, W.; Htay, H.; Myint, K.O.

1994-01-01

Am(Be) neutron source was used for activation of samples and 76 As radioactivity measured by both β- and γ-counting techniques. The samples analyzed were raw materials traditionally used in formulating Myanmar indigenous medicines. The results were compared with those obtained by volumetric analysis and those reported in the literature. (author) 4 refs.; 5 tabs

Determination of arsenic in some Myanmar indigenous medicines by neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Myint, U; Than, W; Htay, H; Myint, K O [Yangon Univ. (Myanmar). Dept. of Chemistry

1994-08-01

Am(Be) neutron source was used for activation of samples and [sup 76]As radioactivity measured by both [beta]- and [gamma]-counting techniques. The samples analyzed were raw materials traditionally used in formulating Myanmar indigenous medicines. The results were compared with those obtained by volumetric analysis and those reported in the literature. (author) 4 refs.; 5 tabs.

Moderation of neutron energy

International Nuclear Information System (INIS)

Marlatt, G.R.

1986-01-01

This patent describes a nuclear reactor system having a nuclear reactor which has a core including fuel assemblies, means for transmitting through the core a coolant, the coolant having a predetermined neutron-energy moderating property, sealed tubes in the core, each tube containing a material having a different neutron-energy moderating property than the coolant, means, when actuated, to engage at least certain of the tubes, for opening certain of the tubes to permit the coolant to replace the material in the tubes thereby to change the energy spectrum of the neutrons in the reactor, hydraulic means, connected to the opening means, for actuating the opening means to engage certain of the tubes to open the tubes. A device, external to the reactor, connected to the hydraulic means controlls the actuation of the opening means, the opening means being so set with reference to the tubes that only certain of the tubes are opened at any time as the opening means is advanced towards the tubes by the hydraulic means

Actinide neutron-induced fission cross section measurements at LANSCE

Energy Technology Data Exchange (ETDEWEB)

Tovesson, Fredrik K [Los Alamos National Laboratory; Laptev, Alexander B [Los Alamos National Laboratory; Hill, Tony S [INL

2010-01-01

Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications in a wide energy range from sub-thermal energies up to 200 MeV. A parallel-plate ionization chamber are used to measure fission cross sections ratios relative to the {sup 235}U standard while incident neutron energies are determined using the time-of-flight method. Recent measurements include the {sup 233,238}U, {sup 239-242}Pu and {sup 243}Am neutron-induced fission cross sections. Obtained data are presented in comparison with ex isting evaluations and previous data.

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.

Irradiation facilities at the advanced neutron source

International Nuclear Information System (INIS)

West, C.D.

1992-01-01

The Advanced Neutron Source (ANS) is a facility, centered around a new 330MW(f) heavy-water cooled and reflected research reactor, proposed for construction at Oak Ridge. The main scientific justification for the new source is the United States' need for increased capabilities in neutron scattering and other neutron beam research, but the technical objectives of the project also cater for the need to replace the irradiation facilities at the aging High Flux Isotope Reactor and to provide other research capabilities to the scientific community. This document provides a description of the ANS facilities

Reflector modelization for neutronic diffusion and parameters identification

International Nuclear Information System (INIS)

Argaud, J.P.

1993-04-01

Physical parameters of neutronic diffusion equations can be adjusted to decrease calculations-measurements errors. The reflector being always difficult to modelize, we choose to elaborate a new reflector model and to use the parameters of this model as adjustment coefficients in the identification procedure. Using theoretical results, and also the physical behaviour of neutronic flux solutions, the reflector model consists then in its replacement by boundary conditions for the diffusion equations on the core only. This theoretical result of non-local operator relations leads then to some discrete approximations by taking into account the multiscaled behaviour, on the core-reflector interface, of neutronic diffusion solutions. The resulting model of this approach is then compared with previous reflector modelizations, and first results indicate that this new model gives the same representation of reflector for the core than previous. (author). 12 refs

Microscopic integral cross section measurements in the Be(d,n) neutron spectrum for applications in neutron dosimetry, radiation damage and the production of long-lived radionuclides

International Nuclear Information System (INIS)

Smith, D.L.; Meadows, J.W.; Greenwood, L.R.

1990-01-01

Integral neutron-reaction cross sections have been measured, relative to the U-238 neutron fission cross-section standard, for 27 reactions which are of contemporary interest in various nuclear applications (e.g., fast-neutron dosimetry, neutron radiation damage and the production of long-lived activities which affect nuclear waste disposal). The neutron radiation field employed in this study was produced by bombarding a thick Be-metal target with 7-MeV deuterons from an accelerator. The experimental results are reported along with detailed information on the associated measurement uncertainties and their correlations. These data are also compared with corresponding calculated values, based on contemporary knowledge of the differential cross sections and of the Be(d,n) neutron spectrum. Some conclusions are reached on the utility of this procedure for neutron-reaction data testing

Experimental study and Monte Carlo modeling of operational quantities in metrology of ionizing radiation: application to neutrons dosimetry by radio-photoluminescence

International Nuclear Information System (INIS)

Salem, Youbba-Ould

2014-01-01

We characterize a passive dosimeter capable of measuring both fast and thermal neutrons for ambient and personal dosimetry. These neutrons can be detected in a mixed neutron-gamma field with appropriate converters (polyethylene for fast neutrons, cadmium for thermal neutrons). Monte Carlo simulations with MCNPX helped with the geometrical conception of the dosimeter and the choice of materials. The responses of the RPL dosimeter to these neutrons are linear in H * (10) and H p (10) with detection limits of 2 mSv for fast neutrons and 0.19 mSv for thermal neutrons. The angular dependencies are satisfactory according to the ISO 21909 norm. A calibration factor of (9.5 ± 0.5)*10 -2 mSv.cm 2 /RPL signal is obtained to the fast neutrons of the IPHC's 241 Am-Be calibrator. This factor is (9.7 ± 0.3)*10 -3 mSv.cm 2 /RPL signal for the thermalized neutrons. (author)

Investigation of Workplace-like Calibration Fields via a Deuterium-Tritium (D-T) Neutron Generator.

Science.gov (United States)

Mozhayev, Andrey V; Piper, Roman K; Rathbone, Bruce A; McDonald, Joseph C

2017-04-01

Radiation survey meters and personal dosimeters are typically calibrated in reference neutron fields based on conventional radionuclide sources, such as americium-beryllium (Am-Be) or californium-252 (Cf), either unmodified or heavy-water moderated. However, these calibration neutron fields differ significantly from the workplace fields in which most of these survey meters and dosimeters are being used. Although some detectors are designed to yield an approximately dose-equivalent response over a particular neutron energy range, the response of other detectors is highly dependent upon neutron energy. This, in turn, can result in significant over- or underestimation of the intensity of neutron radiation and/or personal dose equivalent determined in the work environment. The use of simulated workplace neutron calibration fields that more closely match those present at the workplace could improve the accuracy of worker, and workplace, neutron dose assessment. This work provides an overview of the neutron fields found around nuclear power reactors and interim spent fuel storage installations based on available data. The feasibility of producing workplace-like calibration fields in an existing calibration facility has been investigated via Monte Carlo simulations. Several moderating assembly configurations, paired with a neutron generator using the deuterium tritium (D-T) fusion reaction, were explored.

Bubble masks for time-encoded imaging of fast neutrons.

Energy Technology Data Exchange (ETDEWEB)

Brubaker, Erik; Brennan, James S.; Marleau, Peter; Nowack, Aaron B.; Steele, John T.; Sweany, Melinda; Throckmorton, Daniel J.

2013-09-01

Time-encoded imaging is an approach to directional radiation detection that is being developed at SNL with a focus on fast neutron directional detection. In this technique, a time modulation of a detected neutron signal is inducedtypically, a moving mask that attenuates neutrons with a time structure that depends on the source position. An important challenge in time-encoded imaging is to develop high-resolution two-dimensional imaging capabilities; building a mechanically moving high-resolution mask presents challenges both theoretical and technical. We have investigated an alternative to mechanical masks that replaces the solid mask with a liquid such as mineral oil. Instead of fixed blocks of solid material that move in pre-defined patterns, the oil is contained in tubing structures, and carefully introduced air gapsbubblespropagate through the tubing, generating moving patterns of oil mask elements and air apertures. Compared to current moving-mask techniques, the bubble mask is simple, since mechanical motion is replaced by gravity-driven bubble propagation; it is flexible, since arbitrary bubble patterns can be generated by a software-controlled valve actuator; and it is potentially high performance, since the tubing and bubble size can be tuned for high-resolution imaging requirements. We have built and tested various single-tube mask elements, and will present results on bubble introduction and propagation as a function of tubing size and cross-sectional shape; real-time bubble position tracking; neutron source imaging tests; and reconstruction techniques demonstrated on simple test data as well as a simulated full detector system.

Spectral fluence of neutrons generated by radiotherapeutic Linacs

International Nuclear Information System (INIS)

Kralik, Miloslav; Solc, Jaroslav; Smoldasova, Jana; Vondracek, Vladimir; Farkasova, Estera; Ticha, Ivana

2015-01-01

Spectral fluences of neutrons generated in the heads of the radiotherapeutic linacs Varian Clinac 2100 C/D and Siemens ARTISTE were measured by means of the Bonner spheres spectrometer whose active detector of thermal neutrons was replaced by an activation detector, i.e. a tablet made of pure manganese. Measurements with different collimator settings reveal an interesting dependence of neutron fluence on the area defined by the collimator jaws. The determined neutron spectral fluences were used to derive ambient dose equivalent rate along the treatment coach. To clarify at which components of the linac neutrons are mainly created, the measurements were complemented with MCNPX calculations based on a realistic model of the Varian Clinac. (authors)

Monte Carlo analysis of the Neutron Standards Laboratory of the CIEMAT

International Nuclear Information System (INIS)

Vega C, H. R.; Mendez V, R.; Guzman G, K. A.

2014-10-01

By means of Monte Carlo methods was characterized the neutrons field produced by calibration sources in the Neutron Standards Laboratory of the Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT). The laboratory has two neutron calibration sources: 241 AmBe and 252 Cf which are stored in a water pool and are placed on the calibration bench using controlled systems at distance. To characterize the neutrons field was built a three-dimensional model of the room where it was included the stainless steel bench, the irradiation table and the storage pool. The sources model included double encapsulated of steel, as cladding. With the purpose of determining the effect that produces the presence of the different components of the room, during the characterization the neutrons spectra, the total flow and the rapidity of environmental equivalent dose to 100 cm of the source were considered. The presence of the walls, floor and ceiling of the room is causing the most modification in the spectra and the integral values of the flow and the rapidity of environmental equivalent dose. (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.

High yield neutron generators using the DD reaction

Energy Technology Data Exchange (ETDEWEB)

Vainionpaa, J. H.; Harris, J. L.; Piestrup, M. A.; Gary, C. K.; Williams, D. L.; Apodaca, M. D.; Cremer, J. T. [Adelphi technology, 2003 E. Bayshore Rd. 94061, Redwood City, CA (United States); Ji, Qing; Ludewigt, B. A. [Lawrence Berkeley National Lab, 1 Cyclotron Road, Berkeley, CA 94720 (United States); Jones, G. [G and J Enterprise, 1258 Quary Ln, Suite F, Pleasanton California 94566 (United States)

2013-04-19

A product line of high yield neutron generators has been developed at Adelphi technology inc. The generators use the D-D fusion reaction and are driven by an ion beam supplied by a microwave ion source. Yields of up to 5 Multiplication-Sign 10{sup 9} n/s have been achieved, which are comparable to those obtained using the more efficient D-T reaction. The microwave-driven plasma uses the electron cyclotron resonance (ECR) to produce a high plasma density for high current and high atomic ion species. These generators have an actively pumped vacuum system that allows operation at reduced pressure in the target chamber, increasing the overall system reliability. Since no radioactive tritium is used, the generators can be easily serviced, and components can be easily replaced, providing essentially an unlimited lifetime. Fast neutron source size can be adjusted by selecting the aperture and target geometries according to customer specifications. Pulsed and continuous operation has been demonstrated. Minimum pulse lengths of 50 {mu}s have been achieved. Since the generators are easily serviceable, they offer a long lifetime neutron generator for laboratories and commercial systems requiring continuous operation. Several of the generators have been enclosed in radiation shielding/moderator structures designed for customer specifications. These generators have been proven to be useful for prompt gamma neutron activation analysis (PGNAA), neutron activation analysis (NAA) and fast neutron radiography. Thus these generators make excellent fast, epithermal and thermal neutron sources for laboratories and industrial applications that require neutrons with safe operation, small footprint, low cost and small regulatory burden.

Neutron spectrometry with organic scintillation detector; Espectrometria de nuetrones con cristales de centelleo organicos

Energy Technology Data Exchange (ETDEWEB)

Butragueno Casdo, J L

1972-07-01

This work describes a fast neutron spectrometer using a stilbene crystal as head detector with pulse shape discrimination (P.S.D.) to reject gamma background. Tre experimental procedure involves the P.S.D., the measurements to calibrate the spectrometer and the corrections for several factors, mainly the non-linear response of the stilbene. Results of the measurements with the reaction D{sup 2}(d,n)He{sup 3}, and with an Am-Be neutron source are presented. It is also presented the measurement of the spectrum of the fast reactor CCRAl-1. (Author) 17 refs.

Radioisotopic neutron transmission spectrometry: Quantitative analysis by using partial least-squares method

International Nuclear Information System (INIS)

Kim, Jong-Yun; Choi, Yong Suk; Park, Yong Joon; Jung, Sung-Hee

2009-01-01

Neutron spectrometry, based on the scattering of high energy fast neutrons from a radioisotope and slowing-down by the light hydrogen atoms, is a useful technique for non-destructive, quantitative measurement of hydrogen content because it has a large measuring volume, and is not affected by temperature, pressure, pH value and color. The most common choice for radioisotope neutron source is 252 Cf or 241 Am-Be. In this study, 252 Cf with a neutron flux of 6.3x10 6 n/s has been used as an attractive neutron source because of its high flux neutron and weak radioactivity. Pulse-height neutron spectra have been obtained by using in-house built radioisotopic neutron spectrometric system equipped with 3 He detector and multi-channel analyzer, including a neutron shield. As a preliminary study, polyethylene block (density of ∼0.947 g/cc and area of 40 cmx25 cm) was used for the determination of hydrogen content by using multivariate calibration models, depending on the thickness of the block. Compared with the results obtained from a simple linear calibration model, partial least-squares regression (PLSR) method offered a better performance in a quantitative data analysis. It also revealed that the PLSR method in a neutron spectrometric system can be promising in the real-time, online monitoring of the powder process to determine the content of any type of molecules containing hydrogen nuclei.