Pulsed neutron sources for epithermal neutrons

International Nuclear Information System (INIS)

Windsor, C.G.

1978-01-01

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

Polarized epithermal neutron spectrometer at KENS

International Nuclear Information System (INIS)

Kohgi, M.

1983-01-01

A spectrometer employing a white, epithermal, polarized neutron beam is under construction at KENS. The neutron polarization is achieved by passage through a dynamically polarized proton filter (DPPF). The results of the test experiments show that the DPPF method is promising in obtaining polarized epithermal neutron beam. The basic design of the spectrometer is described

Intraoperative boron neutron capture therapy for malignant gliomas. First clinical results of Tsukuba phase I/II trial using JAERI mixed thermal-epithermal beam

International Nuclear Information System (INIS)

Matsumura, A.; Yamamoto, T.; Shibata, Y.

2000-01-01

Since October 1999, a clinical trial of intraoperative boron neutron capture therapy (IOBNCT) is in progress at JRR-4 (Japan Research Reactor-4) in Japan Atomic Energy Research Institute (JAERI) using mixed thermal-epithermal beam (thermal neutron beam I: TNB-I). Compared to pure thermal beam (thermal neutron beam II: TNB-II), TNB-I has an improved neutron delivery into the deep region than TNB-II. The clinical protocol and the preliminary results will be discussed. (author)

Determination of silver using cyclic epithermal neutron activation analysis

International Nuclear Information System (INIS)

Pun, T.H.; Landsberger, S.

2012-01-01

A fast pneumatic transfer facility was installed in Nuclear Engineering Teaching Laboratory (NETL) of the University of Texas at Austin for the purpose of cyclic thermal and epithermal neutron activation analysis. In this study efforts were focused on the evaluation of cyclic epithermal neutron activation analysis (CENAA). Various NIST and CANMET certified materials were analyzed by the system. Experiment results showed 110 Ag with its 25 s half-life as one of the isotopes favored by the system. Thus, the system was put into practical application in identifying silver in metallic ores. Comparison of sliver concentrations as determined by CENAA in CANMET certified reference materials gave very good results. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Schmitz, Tobias

2016-01-22

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

Epithermal neutron activation analysis using a boron carbide irradiation filter

International Nuclear Information System (INIS)

Ehmann, W.D.; Brueckner, J.

1980-01-01

The use of boron carbide as a thermal neutron filter in epithermal neutron activation (ENAA) analysis has been investigated. As compared to the use of a cadmium filter, boron provides a greater reduction of activities from elements relatively abundant in terrestrial rocks and fossil fuels, such as Na, La, Sc and Fe. These elements have excitation functions which follow the 1/v law in the 1 to 10 eV lower epithermal region. This enhances the sensitivity of ENAA for elements such as U, Th, Ba and etc. which have strong resonances in the higher epithermal region above 10 eV. In addition, a boron carbide filter has the advantages over cadmium of acquiring a relatively low level of induced activity which poses minimal radiation safety problems, when used for ENAA. (author)

Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

International Nuclear Information System (INIS)

Chisela, F.

1986-01-01

The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

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

Science.gov (United States)

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

2015-01-01

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

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

2008-01-01

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

2009-01-01

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

Design of filtered epithermal neutron beams for BNC

International Nuclear Information System (INIS)

Greenwood, R.C.

1986-01-01

The design principles of filters (installed in nuclear reactors) to provide epithermal neutron beams suitable for use in 10 B Neutron Capture Therapy (BNCT) are reviewed. The goal of such filters is to provide epithermal neutron beams within an energy range of 1 keV to 30 keV with fluxes in excess of 5 x 10 8 neutrons/cm 2 .s, and having acceptably low contaminant fast neutron (> 30 keV) and gamma components. Filters considered for this application include 238 U, Sc, Fe/Al and Al/S. It is shown that in order to achieve a goal epithermal neutron flux of > 5 x 10 8 neutrons/cm 2 .s, such filters must be located in radial beam channels which view essentially the complete reactor core. Based on considerations of estimated epithermal fluxes, cost and availability of materials, and transmitted neutron energy spectrum, it is suggested that a filter consisting of elements of Al, S, Ti and V might prove to be an optimum design for BNCT applications. 13 references, 3 figures, 8 tables

Method and apparatus for epithermal neutron porosity well logging

International Nuclear Information System (INIS)

Hertzog, R.C.; Loomis, W.A.; Wraight, P.

1991-01-01

This patent describes a method for investigating the porosity of a subsurface earth formation surrounding a borehole. It comprises repetitively irradiating the borehole and earth formation with discrete bursts of high energy neutrons from a neutron source, which neutrons interact with nuclei of the materials in the borehole and the formation to produce therein populations of epithermal neutrons; detecting the populations of epithermal neutrons at near and far locations in the borehole spaced apart longitudinally by different distances from the neutron source; generating count signals indicative of the magnitudes of the detected epithermal neutron populations at the respective near and far locations; detecting the decay of the epithermal neutron populations following the neutron bursts at least at one location in the borehole and generating signals representative thereof; deriving from the decay signals a signal indicative of the slowing down time of epithermal neutrons in the formation of the at least one location; and deriving from the near and far count signals and the slowing down time signal a measurement signal representative of the porosity of the formation surrounding the borehole inherently compensated for the effects of tool standoff on the responses of the logging tool

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-23

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

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

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

1979-01-01

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

Temperature imaging using epithermal neutrons

International Nuclear Information System (INIS)

Fowler, P.H.; Taylor, A.D.

1987-08-01

The paper concerns the temperature measurement of suitable targets, both remotely and non-invasively, using epithermal neutrons. The text was presented at the Neutron Resonance Radiography Workshop, Los Alamos, U.S.A., 1987. The technique is demonstrated for tantalum foils at different temperatures, using a pulsed beam of epithermal neutrons, at both Los Alamos and ISIS (United Kingdom). Results on the measured time-of-flight spectra and the tantalum resonances are presented. Beam properties and fluxes at ISIS are discussed. Features of the proposed detectors suitable for the temperature technique are outlined, along with the data analysis, the moving targets, the cyclic temperature variations and transients, and the usefulness of the technique. (U.K.)

Theoretical and experimental study of collectrons for epithermal neutron flux in reactors

International Nuclear Information System (INIS)

Agu, M.N.

1986-01-01

A theoretical study of nuclear reactions and electric charge displacements arising in sensitivity to thermal and epithermal neutrons in collectrons allowed a computer code conception. Collectrons in Rhodium, Silver, Cobalt, Hafnium, Erbium, Gadolinium and Holmium have been tested in different radiation fields given by neutron or gamma filters irradiated in different places of Melusine and Siloe reactors. Some emitters were covered with different steel, nickel or zircaloy thicknesses. Theoretical and experimental results are consistent; that validate the computer code and show possibilities and necessity of covering collectron emitters to reduce or cancel the gamma sensitivity and to improve response instantaneity. A selective measurement of epithermal neutron flux can by this way, made by associating two types of collectrons [fr

Epithermal neutron flux characterization of the TRIGA Mark III reactor, Salazar, Mexico, for use in Internal Neutron Activation Analysis

International Nuclear Information System (INIS)

Diaz Rizo, O.; Herrera Peraza, E.

1996-01-01

The non ideality of the epithermal neutron flux distribution at a reactor site parameter (made, using Chloramine-T method. Radiochemical purity and stability of the labelled product were determined by radiochromatography. The labelled Melagenine-II showed two radioactive fractions thermal-to-epithermal neutron ratio (f) were determined in the 3 typical irradiations positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using the Cd-ratio for multi monitor and bare bi-isotopic monitor methods respectively. This characterization is of use in the K o - method of neutron activation analysis, recently introduced at the Institute

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...

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

International Nuclear Information System (INIS)

Mitchell, H.E.

1996-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-04-01

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

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

Science.gov (United States)

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

2016-10-01

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

Epithermal neutron beam interference with cardiac pacemakers

International Nuclear Information System (INIS)

Koivunoro, H.; Serén, T.; Hyvönen, H.; Kotiluoto, P.; Iivonen, P.; Auterinen, I.; Seppälä, T.; Kankaanranta, L.; Pakarinen, S.; Tenhunen, M.; Savolainen, S.

2011-01-01

In this paper, a phantom study was performed to evaluate the effect of an epithermal neutron beam irradiation on the cardiac pacemaker function. Severe malfunction occurred in the pacemakers after substantially lower dose from epithermal neutron irradiation than reported in the fast neutron or photon beams at the same dose rate level. In addition the pacemakers got activated, resulting in nuclides with half-lives from 25 min to 115 d. We suggest that BNCT should be administrated only after removal of the pacemaker from the vicinity of the tumor.

Epithermal neutron beam interference with cardiac pacemakers

Energy Technology Data Exchange (ETDEWEB)

Koivunoro, H., E-mail: hanna.koivunoro@helsinki.fi [Department of Physics, P.O.B. 64, FI-00014 University of Helsinki (Finland)] [Department of Oncology, Helsinki University Central Hospital, P.O.B. 180, FIN-00029 HUS (Finland)] [Boneca Corporation, Finland, Filnland (Finland); Seren, T. [VTT Technical Research Centre of Finland (Finland); Hyvoenen, H. [Boneca Corporation, Finland, Filnland (Finland); Kotiluoto, P. [VTT Technical Research Centre of Finland (Finland); Iivonen, P. [St. Jude Medical (Finland); Auterinen, I. [VTT Technical Research Centre of Finland (Finland); Seppaelae, T.; Kankaanranta, L. [Department of Oncology, Helsinki University Central Hospital, P.O.B. 180, FIN-00029 HUS (Finland); Pakarinen, S. [Department of Cardiology, Helsinki University Central Hospital (Finland); Tenhunen, M. [Department of Oncology, Helsinki University Central Hospital, P.O.B. 180, FIN-00029 HUS (Finland); Savolainen, S. [HUS Helsinki Medical Imaging Center, Helsinki University Central Hospital (Finland)

2011-12-15

In this paper, a phantom study was performed to evaluate the effect of an epithermal neutron beam irradiation on the cardiac pacemaker function. Severe malfunction occurred in the pacemakers after substantially lower dose from epithermal neutron irradiation than reported in the fast neutron or photon beams at the same dose rate level. In addition the pacemakers got activated, resulting in nuclides with half-lives from 25 min to 115 d. We suggest that BNCT should be administrated only after removal of the pacemaker from the vicinity of the tumor.

Measurement of thermal, epithermal and fast neutron flux in the IEA-R1 reactor by the foil activation method

International Nuclear Information System (INIS)

Koskinas, M.F.

1979-01-01

Experimental and theoretical details of the foil activation method applied to neutrons flux measurements at the IEA-R1 reactor are presented. The thermal - and epithermal - neutron flux were determined form activation measurements of gold, cobalt and manganese foils; and for the fast neutron flux determination, aluminum, iron and nickel foils were used. The measurements of the activity induced in the metal foils were performed using a Ge-Li gamma spectrometry system. In each energy range of the reactor neutron spectrum, the agreement among the experimental flux values obtained using the three kind of materials, indicates the consistency of the theoretical approach and of the nuclear parameters selected. (Author) [pt

Installation and testing of an optimized epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR)

Energy Technology Data Exchange (ETDEWEB)

Fairchild, R.G.; Kalef-Ezra, J.; Saraf, S.K.; Fiarman, S.; Ramsey, E.; Wielopolski, L.; Laster, B.; Wheeler, F. (Brookhaven National Lab., Upton, NY (USA); Ioannina Univ. (Greece); Brookhaven National Lab., Upton, NY (USA); State Univ. of New York, Stony Brook, NY (USA). Health Science Center; Brookhaven National Lab., Upton, NY (USA); EG and G Idaho, Inc., Idaho Falls, ID (USA))

1989-01-01

Various calculations indicate that an optimized epithermal neutron beam can be produced by moderating fission neutrons either with a combination of Al and D{sub 2}O, or with Al{sub 2}O{sub 3}. We have designed, installed and tested an Al{sub 2}O{sub 3} moderated epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR). The epithermal neutron fluence rate of 1.8 {times} 10{sup 9} n/cm{sup 2}-sec produces a peak thermal neutron fluence rate of 1.9 to 2.8 {times} 10{sup 9} n/cm{sup 2}-sec in a tissue equivalent (TE) phantom head, depending on the configuration. Thus a single therapy treatment of 5 {times} 10{sup 12} n/cm{sup 2} can be delivered in 30--45 minutes. All irradiation times are given for a BMRR power of 3 MW, which is the highest power which can be delivered continuously. 18 refs., 8 figs., 4 tabs.

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)

The Measurement of Epithermal-to-Thermal U-238 Neutron Capture Rate in Aagesta Power Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Bernander, G

1967-09-15

The epithermal-to-thermal neutron capture rate ratio {rho}{sub 28} in U-238 in Aagesta fuel has been measured by the chemical separation method. The method involves the isolation of Np-239 from uranium and fission products by reversed phase partition chromatography. Although somewhat elaborate, and in spite of difficulties with residual fission products, the method has yielded reasonably accurate results. Further development work on chemical procedures may lead to some improvement. A comparison with the coincidence method - electronic separation of activities - has not shown any large systematic differences between the two methods. The separation of the epithermal U-235 activation from the total has been achieved by means of the '1/v subtraction technique' using copper foils as the 1/v monitor. The complementary thermal column irradiations required have been performed in the research reactors TRIGA (Helsinki) and R1 (Stockholm). From the measured {rho}{sub 28} values the resonance escape probability (p) and the initial conversion ratio (ICR) may be calculated using cross-section data and other lattice parameters. Comparisons with theoretical values of {rho} and ICR as calculated with the BURNUP lattice parameter code are favourable.

Upgrades of the epithermal neutron beam at the Brookhaven Medical Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Liu, Hungyuan B.; Brugger, R.M.; Rorer, D.C.

1994-12-31

The first epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR) was installed in 1988 and produced a neutron beam that was satisfactory for the development of NCT with epithermal neutrons. This beam was used routinely until 1992 when the beam was upgraded by rearranging fuel elements in the reactor core to achieve a 50% increase in usable flux. Next, after computer modeling studies, it was proposed that the Al and Al{sub 2}O{sub 3} moderator material in the shutter that produced the epithermal neutrons could be rearranged to enhance the beam further. However, this modification was not started because a better option appeared, namely to use fission plates to move the source of fission neutrons closer to the moderator and the patient irradiation position to achieve more efficient moderation and production of epithermal neutrons. A fission plate converter (FPC) source has been designed recently and, to test the concept, implementation of this upgrade has started. The predicted beam parameters will be 12 x 10{sup 9} n{sub epi}/cm{sup 2}sec accompanying with doses from fast neutrons and gamma rays per epithermal neutron of 2.8 x 10{sup -11} and < 1 x 10{sup -11} cGycm{sup 2}/n, respectively, and a current-to-flux ratio of epithermal neutrons of 0.78. This conversion could be completed by late 1996.

Epithermal interrogation of fissile waste

International Nuclear Information System (INIS)

Coop, K.L.; Hollas, C.L.

1996-01-01

Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described

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

Science.gov (United States)

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

2011-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-12-15

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Boron thermal/epithermal neutron capture therapy

International Nuclear Information System (INIS)

Fairchild, R.G.

1982-01-01

The development of various particle beams for radiotherapy represents an attempt to improve dose distribution, and to provide high LET radiations which are less sensitive to ambient physical and radiobiological factors such as oxygen tension, cell cycle, and dose rate. In general, a compromise is necessary as effective RBE is reduced in order to spread the dose distribution over the anticipated tumor volume. The approach of delivering stable non-toxic isotopes to tumor, and then activating these atoms subsequently via an external radiation beam has mator advantages; problems associated with high uptake of these isotopes in competing cell pools are obviated, and the general tumor volume can be included in the treatment field of the activating beam. As long as the normal tissues supporting tumor show a low uptake of the isotope to be activated, and as long as the range of the reaction products is short, dose will be restricted to tumor, with a consequent high therapeutic ratio. Neutron Capture Therapy (NCT) is generally carried out by activating boron-10 with low energy neutrons. The range of the high LET, low OER particles from the 10 B(n, α) 7 Li reaction is approx. 10μ, or one cell diameter, a situation that is optimal for cell killing. Significant advantages may be gained by using the NCT procedure in conjunction with improved tissue penetration provided with epithermal or filtered beams, and new compounds showing physiological binding to tumor

Utilization of epithermal neutrons for the determination of molybdenum in the presence of uranium

International Nuclear Information System (INIS)

Oliveira Melo, M.A.M. de.

1984-05-01

Activation analysis by means of selective activation with epithermal neutrons is proposed for the determination of molybdenum in samples when uranium is present. Instrumental activation analysis with epithermal neutrons is advantageous for the determination of elements with large resonance integral, as compared to its thermal neutron activation cross section. The main reason for using this method is the serious interference caused by 99 Mo produced by fission of 235 U. This effect is strongly reduced by using the epicadmium irradiation technique. The filter efficiency has been investigated by irradiation experiments with bare and cadmium-covered samples. A solvent extraction process for uranium, before irradiation, is proposed to reduce sample background. The determination of Mo in leach samples is proposed in order to support the analytical needs of Figueira and Pocos de Caldas Mineral Prospection Programme of Departamento de Tecnologia Mineral from CDTN/NUCLEBRAS (MG,Brazil). The introduction of activation analysis with epithermal neutrons as a routine analytical tool in CDTN is our main goal. This method represents one more opportunity for exploring the analytical facilities available at TRIGA MARK I IPR-R1 nuclear reactor. (Author) [pt

Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities

International Nuclear Information System (INIS)

Mansy, M.S.; Bashter, I.I.; El-Mesiry, M.S.; Habib, N.; Adib, M.

2015-01-01

Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5–133 keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named “QMNB” was developed in the “MATLAB” programming language to perform the required calculations. - Highlights: • Quasi-monoenergetic neutron beams in energy range from (1.5–133) keV. • Interference between the resonance and potential scattering amplitudes. • Epithermal neutron beams used in BNCT

TRANSHEX, 2-D Thermal Neutron Flux Distribution from Epithermal Flux in Hexagonal Geometry

International Nuclear Information System (INIS)

Patrakka, E.

1994-01-01

1 - Description of program or function: TRANSHEX is a multigroup integral transport program that determines the thermal scalar flux distribution arising from a known epithermal flux in two- dimensional hexagonal geometry. 2 - Method of solution: The program solves the isotropic collision probability equations for a region-averaged scalar flux by an iterative method. Either a successive over-relaxation or an inner-outer iteration technique is applied. Flat flux collision probabilities between trigonal space regions with white boundary condition are utilized. The effect of epithermal flux is taken into consideration as a slowing-down source that is calculated for a given spatial distribution and 1/E energy dependence of the epithermal flux

Characteristic analysis on moderating material for obtaining epithermal neutron beam

International Nuclear Information System (INIS)

Jiang Xinbiao; Chen Da; Zhang Ying

2000-01-01

The one dimension discrete coordinates transport code ANISN was used to calculate three-group constants of 11 elements which could be used to consist moderating epithermal neutron material of beam. Moderating character of simple substances, compounds and mixtures consisted of the optimized elements analyzed three kinds of moderating materials were optimized for epithermal neutron beam

In vitro biological effectiveness of JRR-4 epithermal neutron beam. Experiment under free air beam and in water phantom. Cooperative research

CERN Document Server

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

2002-01-01

The surviving curve and the biological effectiveness factor of dose components generated in boron neutron capture therapy (BNCT) were separately determined in neutron beams at Japan Research Reactor No.4. Surviving fraction of V79 Chinese hamster cell with or without sup 1 sup 0 B was obtained using an epithermal neutron beam (ENB), a mixed thermal-epithermal neutron beam (TNB-1), and a thermal neutron beam (TNB-2), which were used or planned to use for BNCT clinical trial. The cell killing effect of these neutron beams with or without the presence of sup 1 sup 0 B depended highly on the neutron beam used, according to the epithermal and fast neutron content in the beam. The biological effectiveness factor values of the boron capture reaction for ENB, TNB-1 and TNB-2 were 3.99+-0.24, 3.04+-0.19 and 1.43+-0.08, respectively. The biological effectiveness factor values of the high-LET dose components based on the hydrogen recoils and the nitrogen capture reaction were 2.50+-0.32, 2.34+-0.30 and 2.17+-0.28 for EN...

Epithermal neutron beam adoption for lung and pancreatic cancer treatment by boron neutron capture therapy

International Nuclear Information System (INIS)

Matsumoto, Tetsuo; Fukushima, Yuji

2001-01-01

The depth-dose distributions were evaluated for possible treatment of both lung and pancreatic cancers using an epithermal neutron beam. The Monte Carlo Neutron Photon (MCNP) calculations showed that physical dose in tumors were 6 and 7 Gy/h, respectively, for lung and pancreas, attaining an epithermal neutron flux of 5 x 10 8 ncm -2 s -1 . The boron concentrations were assumed at 100 ppm and 30 ppm, respectively, for lung and pancreas tumors and normal tissues contains 1/10 tumor concentrations. The dose ratios of tumor to normal tissue were 2.5 and 2.4, respectively, for lung and pancreas. The dose evaluation suggests that BNCT using an epithermal neutron beam could be applied for both lung and pancreatic cancer treatment. (author)

A new position-sensitive transmission detector for epithermal neutron imaging

International Nuclear Information System (INIS)

Schooneveld, E M; Kockelmann, W; Rhodes, N; Tardocchi, M; Gorini, G; Perelli Cippo, E; Nakamura, T; Postma, H; Schillebeeckx, P

2009-01-01

A new neutron resonant transmission (NRT) detector for epithermal neutron imaging has been designed and built for the ANCIENT CHARM project, which is developing a set of complementary neutron imaging methods for analysis of cultural heritage objects. One of the techniques being exploited is NRT with the aim of performing bulk elemental analysis. The 16-pixel prototype NRT detector consists of independent crystals of 2 x 2 mm pixel size, which allow for 2D position-sensitive transmission measurements with epithermal neutrons. First results obtained at the ISIS pulsed spallation neutron source are presented. (fast track communication)

Mechanical design of epithermal neutron diagnostic for TFTR

International Nuclear Information System (INIS)

Groo, R.C.

1981-01-01

The mechanical design of the Epithermal Neutron Diagnostic for TFTR is described. This fission detector system measures the time resolution of the neutron flux for folding into the Neutron Activation system and also provides continuous, wide range coverage of all expected fusion reaction rates

Epithermal beam development at the BMRR [Brookhaven Medical Research Reactor]: Dosimetric evaluation

International Nuclear Information System (INIS)

Saraf, S.K.; Fairchild, R.G.; Kalef-Ezra, J.; Laster, B.H.; Fiarman, S.; Ramsey, E.; Ioannina Univ.; Brookhaven National Lab., Upton, NY; State Univ. of New York, Stony Brook, NY

1989-01-01

The utilization of an epithermal neutron beam for neutron capture therapy (NCT) is desirable because of the increased tissue penetration relative to a thermal neutron beam. Over the past few years, modifications have been and continue to be made at the Brookhaven Medical Research Reactor (BMRR) by changing its filter components to produce an optimal epithermal beam. An optimal epithermal beam should contain a low fast neutron contamination and no thermal neutrons in the incident beam. Recently a new moderator for the epithermal beam has been installed at the epithermal port of the BMRR and has accomplished this task. This new moderator is a combination of alumina (Al 2 O 3 ) bricks and aluminum (Al) plates. A 0.51 mm thick cadmium (Cd) sheet has reduced the thermal neutron intensity drastically. Furthermore, an 11.5 cm thick bismuth (Bi) plate installed at the port surface has reduced the gamma dose component to negligible levels. Foil activation techniques have been employed by using bare gold and cadmium-covered gold foil to determine thermal as well as epithermal neutron fluence. Fast neutron fluence has been determined by indium foil counting. Fast neutron and gamma dose in soft tissue, free in air, is being determined by the paired ionization chamber technique, using tissue equivalent (TE) and graphite chambers. Thermoluminescent dosimeters (TLD-700) have also been used to determine the gamma dose independently. This paper describes the methods involved in the measurements of the above mentioned parameters. Formulations have been developed and the various corrections involved have been detailed. 12 refs

Power Burst Reactor Facility as an epithermal neutron source for brain cancer therapy

International Nuclear Information System (INIS)

Wheeler, F.J.

1986-01-01

The Power Burst Facility (PBF) reactor is considered for modification to provide an intense, clean source of intermediate-energy (epithermal) neutrons desirable for clinical studies of neutron capture therapy (NCT) for malignant tumors. The modifications include partial replacement of the reflector, installation of a neutron-moderating, shifting region, additional shielding, and penetration of the present concrete shield with a collimating (and optionally) filtering region. The studies have indicated that the reactor, after these modifications, will be safely operable at full power (28 MW) within the acceptable limits of the plant protection systems. The neutron beam exiting from the collimator port is predicted to be of sufficient intensity (approx.10 10 neutrons/cm 2 -s) to provide therapeutic doses in very short irradiation times. The beam would be relatively free of undesirable fast neutrons, thermal neutrons and gamma rays. The calculated neutron energy spectrum and associated gamma rays in the beam were provided as input in simulation studies that used a computer model of a patient with a brain tumor to determine predicted dose rates to the tumor and healthy tissue. The results of this conceptual study indicate an intense, clean beam of epithermal neutrons for NCT clinical trials is attainable in the PBF facility with properly engineered design modifications. 9 refs., 11 figs., 3 tabs

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

International Nuclear Information System (INIS)

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

2004-01-01

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

The Measurement of Epithermal-to-Thermal U-238 Neutron Capture Rate (ρ28) in Aagesta Power Reactor Fuel

International Nuclear Information System (INIS)

Bernander, G.

1967-09-01

The epithermal-to-thermal neutron capture rate ratio ρ 28 in U-238 in Aagesta fuel has been measured by the chemical separation method. The method involves the isolation of Np-239 from uranium and fission products by reversed phase partition chromatography. Although somewhat elaborate, and in spite of difficulties with residual fission products, the method has yielded reasonably accurate results. Further development work on chemical procedures may lead to some improvement. A comparison with the coincidence method - electronic separation of activities - has not shown any large systematic differences between the two methods. The separation of the epithermal U-235 activation from the total has been achieved by means of the '1/v subtraction technique' using copper foils as the 1/v monitor. The complementary thermal column irradiations required have been performed in the research reactors TRIGA (Helsinki) and R1 (Stockholm). From the measured ρ 28 values the resonance escape probability (p) and the initial conversion ratio (ICR) may be calculated using cross-section data and other lattice parameters. Comparisons with theoretical values of ρ and ICR as calculated with the BURNUP lattice parameter code are favourable

Characterization of the Ljubljana TRIGA thermal column neutron radiographic facility

International Nuclear Information System (INIS)

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

1995-01-01

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

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

Shultz, W.E.

1980-01-01

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

Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

International Nuclear Information System (INIS)

Burns, T.D. Jr.

1995-05-01

The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 x 10 8 n/cm 2 · s. The fast neutron and gamma radiation KERMA factors are 10 x 10 -11 cGy·cm 2 /n epi and 20 x 10 -11 cGy·cm 2 /n epi , respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power

Dose planning with comparison to in vivo dosimetry for epithermal neutron irradiation of the dog brain

International Nuclear Information System (INIS)

Seppaelae, Tiina; Auterinen, Iiro; Aschan, Carita; Seren, Tom; Benczik, Judit; Snellman, Marjatta; Huiskamp, Rene; Ramadan, Usama Abo; Kankaanranta, Leena; Joensuu, Heikki; Savolainen, Sauli

2002-01-01

Boron neutron capture therapy (BNCT) is an experimental type of radiotherapy, presently being used to treat glioblastoma and melanoma. To improve patient safety and to determine the radiobiological characteristics of the epithermal neutron beam of Finnish BNCT facility (FiR 1) dose-response studies were carried on the brain of dogs before starting the clinical trials. A dose planning procedure was developed and uncertainties of the epithermal neutron-induced doses were estimated. The accuracy of the method of computing physical doses was assessed by comparing with in vivo dosimetry. Individual radiation dose plans were computed using magnetic resonance images of the heads of 15 Beagle dogs and the computational model of the FiR 1 epithermal neutron beam. For in vivo dosimetry, the thermal neutron fluences were measured using Mn activation foils and the gamma-ray doses with MCP-7s type thermoluminescent detectors placed both on the skin surface of the head and in the oral cavity. The degree of uncertainty of the reference doses at the thermal neutron maximum was estimated using a dose-planning program. The estimated uncertainty (±1 standard deviation) in the total physical reference dose was ±8.9%. The calculated and the measured dose values agreed within the uncertainties at the point of beam entry. The conclusion is that the dose delivery to the tissue can be verified in a practical and reliable fashion by placing an activation dosimeter and a TL detector at the beam entry point on the skin surface with homogeneous tissues below. However, the point doses cannot be calculated correctly in the inhomogeneous area near air cavities of the head model with this type of dose-planning program. This calls for attention in dose planning in human clinical trials in the corresponding areas

Optimization in Activation Analysis by Means of Epithermal Neutrons. Determination of Molybdenum in Steel

Energy Technology Data Exchange (ETDEWEB)

Brune, D; Jirlow, J

1963-12-15

Optimization in activation analysis by means of selective activation with epithermal neutrons is discussed. This method was applied to the determination of molybdenum in a steel alloy without recourse to radiochemical separations. The sensitivity for this determination is estimated to be 10 ppm. With the common form of activation by means of thermal neutrons, the sensitivity would be about one-tenth of this. The sensitivity estimations are based on evaluation of the photo peak ratios of Mo-99/Fe-59.

Experimental demonstration of a compact epithermal neutron source based on a high power laser

Science.gov (United States)

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

2017-07-01

Epithermal neutrons from pulsed-spallation sources have revolutionised neutron science allowing scientists to acquire new insight into the structure and properties of matter. Here, we demonstrate that laser driven fast (˜MeV) neutrons can be efficiently moderated to epithermal energies with intrinsically short burst durations. In a proof-of-principle experiment using a 100 TW laser, a significant epithermal neutron flux of the order of 105 n/sr/pulse in the energy range of 0.5-300 eV was measured, produced by a compact moderator deployed downstream of the laser-driven fast neutron source. The moderator used in the campaign was specifically designed, by the help of MCNPX simulations, for an efficient and directional moderation of the fast neutron spectrum produced by a laser driven source.

OPTIMIZATION OF THE EPITHERMAL NEUTRON BEAM FOR BORON NEUTRON CAPTURE THERAPY AT THE BROOKHAVEN MEDICAL RESEARCH REACTOR.

Energy Technology Data Exchange (ETDEWEB)

HU,J.P.; RORER,D.C.; RECINIELLO,R.N.; HOLDEN,N.E.

2002-08-18

Clinical trials of Boron Neutron Capture Therapy for patients with malignant brain tumor had been carried out for half a decade, using an epithermal neutron beam at the Brookhaven's Medical Reactor. The decision to permanently close this reactor in 2000 cut short the efforts to implement a new conceptual design to optimize this beam in preparation for use with possible new protocols. Details of the conceptual design to produce a higher intensity, more forward-directed neutron beam with less contamination from gamma rays, fast and thermal neutrons are presented here for their potential applicability to other reactor facilities. Monte Carlo calculations were used to predict the flux and absorbed dose produced by the proposed design. The results were benchmarked by the dose rate and flux measurements taken at the facility then in use.

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

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2001-01-01

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

Combination of epithermal and inelastic neutron scattering methods to locate coal and oil-shale zones

International Nuclear Information System (INIS)

Schultz, W.E.

1976-01-01

A pulsed neutron generator of the deuterium-tritium reaction type irradiates earth formations in the vicinity of a borehole with 14 MeV neutrons. Gamma rays produced by the inelastic scattering of the fast neutrons are observed in four energy regions of the gamma ray energy spectrum corresponding to the inelastic scattering of neutrons by carbon, oxygen, silicon, and calcium. The carbon/oxygen, calcium/silicon, and carbon plus oxygen gamma rays are found and combined with a separately derived hydrogen index log to determine the quality of coal-bearing formations or oil-shale regions. The hydrogen index curve is found preferably by a dual-spaced detector epithermal neutron porosity logging technique or from a conventional thermal neutron gamma ray log

In vitro biological effectiveness of JRR-4 epithermal neutron beam. Experiment under free air beam and in water phantom. Cooperative research

International Nuclear Information System (INIS)

Yamamoto, Tetsuya; Matsumura, Akira; Nose, Tadao; Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Hori, Naohiko; Torii, Yoshiya; Horiguchi, Yoji

2002-05-01

The surviving curve and the biological effectiveness factor of dose components generated in boron neutron capture therapy (BNCT) were separately determined in neutron beams at Japan Research Reactor No.4. Surviving fraction of V79 Chinese hamster cell with or without 10 B was obtained using an epithermal neutron beam (ENB), a mixed thermal-epithermal neutron beam (TNB-1), and a thermal neutron beam (TNB-2), which were used or planned to use for BNCT clinical trial. The cell killing effect of these neutron beams with or without the presence of 10 B depended highly on the neutron beam used, according to the epithermal and fast neutron content in the beam. The biological effectiveness factor values of the boron capture reaction for ENB, TNB-1 and TNB-2 were 3.99±0.24, 3.04±0.19 and 1.43±0.08, respectively. The biological effectiveness factor values of the high-LET dose components based on the hydrogen recoils and the nitrogen capture reaction were 2.50±0.32, 2.34±0.30 and 2.17±0.28 for ENB, TNB-1 and TNB-2, respectively. The biological effectiveness factor values of the neutron and photon components were 1.22±0.16, 1.23±0.16 and 1.21±0.16, respectively. The depth function of biological effectiveness factor in water phantom and the difference in biological effectiveness factor among boron compounds were also determined. The experimental determination of biological effectiveness factor outlined in this paper is applicable to the dose calculation for each dose component of the neutron beams and contribute to an accurate biological effectiveness factor as comparison with a neutron beam at a different facility employed in ongoing and planned BNCT clinical trials. (author)

Bromine and iodine in Chinese medical herbs determined via epithermal neutron activation analysis

International Nuclear Information System (INIS)

Chien-Yi Chen; Yuan-Yaw Wei; Sheng-Pin ChangLai; Lung-Kwang Pan

2003-01-01

Nineteen natural herbs and two prescriptions prepared from mixed herbs were analyzed via epithermal neutron activation analysis (ENAA) to evaluate their bromine and iodine concentration. Traditional medical doctors prescribed the samples presented in this work to most Taiwanese children for strengthening their immune systems. Empirical results indicated a wide diversity of bromine in the samples. Yet, the iodine concentration was only around one to tenth or twentieth of the bromine. The maximum daily intake (MDI) for various medical herbs was also widely diversified from one to tenfold on the basis of various criteria. The minimum detectable concentration (MDC) of bromine and iodine found was 0.42±0.14 ppm and 0.067±0.016 ppm, respectively. Compared to that from conventional thermal neutron activation analysis (NAA) for a similar evaluation, the extremely low MDC obtained here was attributed to the large amount of thermal neutron absorption during sample irradiation. (author)

Method and apparatus for dual-spaced fast/epithermal neutron porosity measurements

International Nuclear Information System (INIS)

Smith, H.D. Jr.

1986-01-01

A method is described for determining the porosity of earth formations in the vicinity of a well borehole, comprising: (a) irradiating the earth formations in the vicinity of the well borehole with a continuous chemical type source of fast neutrons, (b) detecting the fast neutron population at a first shorter spaced distance from the neutron source in the borehole and generating signals representative thereof, (c) detecting the epithermal neutron population at a second space distance from the neutron source in the borehole and generating signals representative thereof, the second spaced distance being greater than the first spaced distance from the neutron source, (d) forming a ratio of the signals representing the fast and epithermal neutron populations to derive a measurement signal functionally related to the porosity of the earth formations in the vicinity of the borehole, and (e) calibrating the measurement signal according to a predetermined functional relationship to derive a porosity signal quantitatively representative of the porosity of the earth formations in the vicinity of the borehole

Experimental characterization of semiconductor-based thermal neutron detectors

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-21

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

An in-situ check of the epithermal neutron log calibration

International Nuclear Information System (INIS)

Burkhard, N.R.

1993-01-01

The epithermal neutron log is used to measure the water content of the formation. The large hole epithermal neutron sonde (ENS) utilized at the Nevada Test Site (NTS) has been calibrated in the Hydrogen Content Test Facility (HCTF). These calibrations are used to correct the measured neutron count rate for the effects of tool stand-off and density. For some time, the suspicion has existed that the water contents that are calculated from the ENS data are too large. Hole U2gj represented a unique opportunity to check the validity of the ENS calibration under realistic logging conditions; a portion of the hole had been cemented and redrilled and then logged. The cements have a known water content and can be used as an in-situ calibration check. The author found that the water contents from the log data after processing with the existing calibrations are consistent with these known cement water contents. In addition, the study indicates that the raw neutron data might be more appropriately smoothed by using a median smoother rather than the currently utilized mean smoother

An in-situ check of the epithermal neutron log calibration

International Nuclear Information System (INIS)

Burkhard, N.R.

1993-09-01

The epithermal neutron log is used to measure the water content of the formation. The large hole epithermal neutron sonde (ENS) that we utilize at the Nevada Test Site (NTS) has been calibrated in the Hydrogen Content Test Facility (HCTF). These calibrations are used to correct the measured neutron count rate for the effects of tool stand-off and density. For sometime, the suspicion has existed that the water contents that are calculated from the ENS data are too large. Hole U2gj represented a unique opportunity to check the validity of the ENS calibration under realistic logging conditions; a portion of the hole had been cemented and re drilled and then logged. The cements have a known water content and can be used as an in situ calibration check. I found that the water contents from the log data after processing with the existing calibrations are consistent with these known cement water contents. In addition, the study indicates that the raw neutron data might be more appropriately smoothed by using a median smoother rather than the currently utilized mean smoother

The epithermal neutron-flux distribution in the reactor RA - Vinca

International Nuclear Information System (INIS)

Marinkov, V.; Bikit, I.; Martinc, R.; Veskovic, M.; Slivka, J.; Vaderna, S.

1987-01-01

The distribution of the epithermal neutron flux in the reactor RA - Vinca has been measured by means of Zr - activation detectors. In the channel VK-8 non-homogeneous flux distribution was observed (author) [sr

Advances on detectors for low-angle scattering of epithermal neutrons

International Nuclear Information System (INIS)

Perelli Cippo, E; Gorini, G; Tardocchi, M; Andreani, C; Pietropaolo, A; Senesi, R; Rhodes, N J; Schoonveld, E M

2008-01-01

The Very Low Angle Detector (VLAD) installed at the ISIS spallation neutron source is a novel instrument for epithermal neutron scattering with a range of applications in solid state physics. VLAD extends the kinematical space of the VESUVIO spectrometer to low momentum transfers at neutron energies above 1 eV. Measurements at scattering angles as low as 1° have been made with limitations due to the achievable signal/background ratio. (technical design note)

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

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

1980-01-01

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

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

International Nuclear Information System (INIS)

Harling, Otto K.

2009-01-01

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

Measurement of trace cadmium and elements in bone by epithermal neutron activation analysis

International Nuclear Information System (INIS)

Dowlati, R.; Jervis, R.E.

1991-01-01

Epithermal neutron activation analysis (ENAA) was applied to measure quantitatively Cd and other elements in bone samples from control and Cd-fed rats. This method was found to be non-destructive to the bone samples, with no sign of 'radiolytic charring' and was sensitive enough to detect and quantify Cd in bone samples at normal levels for mammals (viz. 0.5-1.0 μg/g) and higher. Two different thermal neutron shield materials were utilized, namely cadmium and boron. The boron shield resulted in a 27% improvement in the detection limit of Cd in bone. The accuracy of ENAA for Cd was assessed by intercomparison with electrothermal atomic absorption spectrophotometry (ETAAS), and the results in fair agreement (±23%) with those from ENAA

Comparison of different methods for activation analysis of geological and pedological samples: Reactor and epithermal neutron activation, relative and monostandard method

International Nuclear Information System (INIS)

Alian, A.; Sansoni, B.

1980-04-01

Using purely instrumental methods, a comparative study is presented on neutron activation analysis of rock and soil samples by whole reactor neutron spectrum and epithermal neutrons with both relative and monostandard procedures. The latter procedure used with epithermal neutron activation analysis of soil samples necessitated the use of the 'effective resonance integrals' which were determined experimentally. The incorporation of the #betta# factor, representing deviation of reactor epithermal neutron flux from 1/E law, is developed in the present work. The main criteria for the choice of one or more of the procedures studied for a given purpose are also indicated. Analysis of 15 trace elements, Ca and Fe in the standard Japanese granite JG-1 using monostandard epithermal neutron activation gave results in good agreement with the average literature values. (orig./RB) [de

8-group relative delayed neutron yields for epithermal neutron induced fission of 235U and 239Pu

International Nuclear Information System (INIS)

Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G

2002-01-01

An 8-group representation of relative delayed neutron yields was obtained for epithermal neutron induced fission of 235 U and 239 Pu. These data were compared with ENDF/B-VI data in terms of the average half- life of the delayed neutron precursors and on the basis of the dependence of reactivity on the asymptotic period. (author)

Personal neutron dosimeter using solid-state track detector

International Nuclear Information System (INIS)

Mettripan, S.

1980-01-01

A cellulose nitrate film coated on both sides with lithium tetraborate was used as a neutron dosimeter for surveillance of personnel exposed to thermal and epithermal neutron. It was found that the optimum etching conditions used were 10% solution of sodium hydroxide, 60 degrees C and 20 minutes etching time and the alpha track densities from the (n,α) reaction on the films were proportional to thermal and epithermal neutron fluxes. The response of the film was found to be 1.068 x 10 -3 tracks per thermal neutron and 3.438 x 10 -4 tracks per epithermal neutron

Characterization of the γ background in epithermal neutron scattering measurements at pulsed neutron sources

International Nuclear Information System (INIS)

Pietropaolo, A.; Tardocchi, M.; Schooneveld, E.M.; Senesi, R.

2006-01-01

This paper reports the characterization of the different components of the γ background in epithermal neutron scattering experiments at pulsed neutron sources. The measurements were performed on the VESUVIO spectrometer at ISIS spallation neutron source. These measurements, carried out with a high purity germanium detector, aim to provide detailed information for the investigation of the effect of the γ energy discrimination on the signal-to-background ratio. It is shown that the γ background is produced by different sources that can be identified with their relative time structure and relative weight

Microscopic cross-section measurements by thermal neutron activation

International Nuclear Information System (INIS)

Avila L, J.

1987-08-01

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

Pulsed neutron porosity logging system

International Nuclear Information System (INIS)

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

1978-01-01

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

Monte Carlo simulation of a coded-aperture thermal neutron camera

International Nuclear Information System (INIS)

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

2011-01-01

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

Neutron--neutron logging

International Nuclear Information System (INIS)

Allen, L.S.

1977-01-01

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

Thermal neutron standard fields with the KUR heavy water facility

International Nuclear Information System (INIS)

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

1978-01-01

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

Determination of trace cadmium and other elements in bone by epithermal neutron activation analysis

International Nuclear Information System (INIS)

Dowlati, R.; Jervis, R.E.

1991-01-01

Epithermal neutron activation analysis (ENAA) was applied to measure quantitatively Cd and other elements in bone samples from control and Cd-fed rats. This method was found to be non-destructive to the bone samples, with no sign of 'radiolytic charring' and was sensitive enough to detect and quantify Cd in bone samples at normal levels for mammals (viz. 0.5-1.0μg/g) and higher. Two different thermal neutron shield materials were utilized, namely cadmium and boron. The boron shield resulted in a 27% improvement in the detection limit of Cd in bone. The accuracy of ENAA for Cd was assessed by intercomparison with electrothermal atomic absorption spectrophotometry (ETAAS), and the results were in fair agreement (±23%) with those from ENAA. (author) 24 refs.; 5 tabs

Upgrade for the epithermal neutron beam at NRI Rez

International Nuclear Information System (INIS)

Marek, M.; Flibor, S.; Viererbl, L.; Burian, J.; Rejchrt, J.; Klupak, V.; Gambarini, G.; Vanossi, E.

2006-01-01

The epithermal neutron beam facility designed for pre-clinical neutron capture therapy research has been operated at LVR-15 reactor for more than ten years. The construction of the beam filter has been recently modified especially for the shielding quality of the beam shutter to be improved. The parameters of the upgraded beam were calculated with the MCNP code and a new source term for the NCTPLAN treatment planning software was evaluated. The calculated source term was consequently scaled according to the results of measurements in the free beam and in the 50x50x25 cm 3 water phantom. (author)

Errors of absolute methods of reactor neutron activation analysis caused by non-1/E epithermal neutron spectra

International Nuclear Information System (INIS)

Erdtmann, G.

1993-08-01

A sufficiently accurate characterization of the neutron flux and spectrum, i.e. the determination of the thermal flux, the flux ratio and the epithermal flux spectrum shape factor, α, is a prerequisite for all types of absolute and monostandard methods of reactor neutron activation analysis. A convenient method for these measurements is the bare triple monitor method. However, the results of this method, are very imprecise, because there are high error propagation factors form the counting errors of the monitor activities. Procedures are described to calculate the errors of the flux parameters, the α-dependent cross-section ratios, and of the analytical results from the errors of the activities of the monitor isotopes. They are included in FORTRAN programs which also allow a graphical representation of the results. A great number of examples were calculated for ten different irradiation facilities in four reactors and for 28 elements. Plots of the results are presented and discussed. (orig./HP) [de

Qualitative dose response of the normal canine head to epithermal neutron irradiation with and without boron capture

International Nuclear Information System (INIS)

DeHaan, C.E.; Gavin, P.R.; Kraft, S.L.; Wheeler, F.J.; Atkinson, C.A.

1992-01-01

Boron Neutron Capture Therapy is being re-evaluated for the treatment of intracranial tumors. Prior to human clinical trials, determination of normal tissue tolerance is critical. Dogs were chosen as a large animal model for the following reasons. Dogs can be evaluated with advanced imaging, diagnostic and therapeutic modalities. Dogs are amenable to detailed neurologic examination and subtle behavioral changes are easily detected. Specifically, Labrador retrievers were chosen for their large body and head size. The dogs received varying doses of epithermal neutron irradiation and boron neutron capture irradiation using an epithermal neutron source. The dogs were closely monitored for up to one year post irradiation

Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961); Pile AQUILON. Durcissement du spectre des neutrons dans les barreaux d'uranium et calcul des fissions epithermiques (1961)

Energy Technology Data Exchange (ETDEWEB)

Durand -Smet, R; Lourme, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

- Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott '{beta}' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors) [French] - Un certain nombre de mesures effectuees dans la pile Aquilon ont permis d'etablir la distribution fine des flux thermique et epithermique dans les barreaux d'uranium, et d'en deduire les variations du spectre des neutrons dans l'uranium, le terme {beta} du spectre de Wescott moyen dans le barreau et le nombre de fissions epithermiques. En annexe, il est propose une definition nouvelle du coefficient de multiplication infini, qui fait intervenir les parametres epithermiques. (auteurs)

Epithermal neutron beam design for neutron capture therapy at the Power Burst Facility and the Brookhaven Medical Research Reactor

International Nuclear Information System (INIS)

Wheeler, F.J.; Parsons, D.K.; Rushton, B.L.; Nigg, D.W.

1990-01-01

Nuclear design studies have been performed for two reactor-based epithermal neutron beams for cancer treatment by neutron capture therapy (NCT). An intermediate-intensity epithermal beam has been designed and implemented at the Brookhaven Medical Research Reactor (BMRR). Measurements show that the BMRR design predictions for the principal characteristics of this beam are accurate. A canine program for research into the biological effects of NCT is now under way at BMRR. The design for a high-intensity epithermal beam with minimal contamination from undesirable radiation components has been finalized for the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. This design will be implemented when it is determined that human NCT trials are advisable. The PBF beam will exhibit approximately an order of magnitude improvement in absolute epithermal flux intensity over that available in the BMRR, and its angular distribution and spectral characteristics will be more advantageous for NCT. The combined effects of beam intensity, angular distribution, spectrum, and contaminant level allow the desired tumor radiation dose to be delivered in much shorter times than are possible with the currently available BMRR beam, with a significant reduction (factor of 3 to 5) in collateral dose due to beam contaminants

Collimator optimization studies for the new MIT epithermal neutron beam

International Nuclear Information System (INIS)

Riley, K.J.; Ali, S.J.; Harling, O.K.

2000-01-01

A patient collimator has been designed for the epithermal neutron facility now being commissioned at MIT. Collimator performance both in and out of field was evaluated using the Monte Carlo code MCNP. A two piece design that can accommodate different circular field sizes will be manufactured using a composite lead, epoxy, boron and lithium mixture. (author)

YAP scintillators for resonant detection of epithermal neutrons at pulsed neutron sources

International Nuclear Information System (INIS)

Tardocchi, M.; Gorini, G.; Pietropaolo, A.; Andreani, C.; Senesi, R.; Rhodes, N.; Schooneveld, E. M.

2004-01-01

Recent studies indicate the resonance detector (RD) technique as an interesting approach for neutron spectroscopy in the electron volt energy region. This work summarizes the results of a series of experiments where RD consisting of YAlO 3 (YAP) scintillators were used to detect scattered neutrons with energy in the range 1-200 eV. The response of YAP scintillators to radiative capture γ emission from a 238 U analyzer foil was characterized in a series of experiments performed on the VESUVIO spectrometer at the ISIS pulsed neutron source. In these experiments a biparametric data acquisition allowed the simultaneous measurements of both neutron time-of-flight and γ pulse height (energy) spectra. The analysis of the γ pulse height and neutron time of flight spectra permitted to identify and distinguish the signal and background components. These measurements showed that a significant improvement in the signal-to-background ratio can be achieved by setting a lower level discrimination on the pulse height at about 600 keV equivalent photon energy. Present results strongly indicate YAP scintillators as the ideal candidate for neutron scattering studies with epithermal neutrons at both very low (<5 deg.) and intermediate scattering angles

Calculations of neutron spectra after neutron-neutron scattering

Energy Technology Data Exchange (ETDEWEB)

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

2004-09-01

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

Pulse neutron logging technique

International Nuclear Information System (INIS)

Bespalov, D.F.; Dylyuk, A.A.

1975-01-01

A new method of neutron-burst logging is proposed, residing in irradiating rocks with fast neutron bursts and registering the integrated flux burst of thermal and/or epithermal neutrons, from the moment of its initiation to that of full absorption. The obtaained value is representative of the rock properties (porosity, hydrogen content). The integrated flux in a burst of thermal and epithermal neutrons can be measured both by way of activation of a reference sample of a known chemical composition during the neutron burst and by recording the radiation of induced activity of the sample within an interval between two bursts. The proposed method features high informative value, accuracy and efficiency

Fundamental symmetry studies at Los Alamos using epithermal neutrons

International Nuclear Information System (INIS)

Bowman, C.D.; Bowman, J.D.; Yuan, V.W.

1988-01-01

Fundamental symmetry studies using intense polarized beams of epithermal neutrons are underway at the LANSCE facility of the Los Alamos National Laboratory. Three classes of symmetry experiments can be explored: parity violation, and time reversal invariance violation for both parity-violating and parity-conserved observables. The experimental apparatus is described and performance illustrated with examples of recent measurements. Possible improvements in the facilities and prospective experiments are discussed. 15 refs., 10 figs

Neutron reflector design with Californium 252 neutron for Boron neutron chapter therapy facility using MCNP5 simulation method

International Nuclear Information System (INIS)

Muhammad Fakhrurreza; Kusminanto; Y Sardjono

2014-01-01

In this research has made a reflector design to provide beams of Neutron for BNCT with Californium-252 radioactive source. This collimator is useful to obtain optimum epithermal neutron flux with the smallest impurity radiation (thermal neutron, fast neutron, and gamma). The design process is done using Monte Carlo N-Particle simulation version 5 (MCNP5) code to calculate the neutron flux tally form. The chosen reflector design is the reflectors which use material such as BeO ceramic with 13 cm thick. Moderator use sulfur material with the slope angle of the cone is 30°. From the calculation result, it is obtained that Reflector with 1 gram Californium-252 source can produce a neutron output thermal which has thermal neutron specification 2.23189 x 10 9 n/s.cm 2 , epithermal neutron 3.51548 x 10 9 n/s.cm 2 , and fast neutron 4.82241 x 10 9 n/s.cm 2 From the result, it needs additional collimator because the BNCT requirement. (author)

The secondary neutron sources for generation of particular neutron fluxes

International Nuclear Information System (INIS)

Tracz, G.

2007-07-01

The foregoing paper presents the doctor's thesis entitled '' The secondary neutron sources for generation of particular neutron fluxes ''. Two secondary neutron sources have been designed, which exploit already existing primary sources emitting neutrons of energies different from the desired ones. The first source is devoted to boron-neutron capture therapy (BNCT). The research reactor MARIA at the Institute of Atomic Energy in Swierk (Poland) is the primary source of the reactor thermal neutrons, while the secondary source should supply epithermal neutrons. The other secondary source is the pulsed source of thermal neutrons that uses fast 14 MeV neutrons from a pulsed generator at the Institute of Nuclear Physics PAN in Krakow (Poland). The physical problems to be solved in the two mentioned cases are different. Namely, in order to devise the BNCT source the initial energy of particles ought to be increased, whilst in the other case the fast neutrons have to be moderated. Slowing down of neutrons is relatively easy since these particles lose energy when they scatter in media; the most effective moderators are the materials which contain light elements (mostly hydrogen). In order to increase the energy of neutrons from thermal to epithermal (the BNCT case) the so-called neutron converter should be exploited. It contains a fissile material, 235 U. The thermal neutrons from the reactor cause fission of uranium and fast neutrons are emitted from the converter. Then fissile neutrons of energy of a few MeV are slowed down to the required epithermal energy range. The design of both secondary sources have been conducted by means of Monte Carlo simulations, which have been carried out using the MCNP code. In the case of the secondary pulsed thermal neutron source, some of the calculated results have been verified experimentally. (author)

Treatment planning figures of merit in thermal and epithermal boron capture therapy of brain tumours

Energy Technology Data Exchange (ETDEWEB)

Wallace, S.A.; Mathur, J.N. (Wollongong Univ., NSW (Australia)); Allen, B.J. (Ansto PMB 1 Menai, NSW (Australia). Biomedicine and Health)

1994-05-01

The boron neutron capture therapy (BNCT) figures of merit of advantage depth, therapeutic depth, modified advantage depth and maximum therapeutic depth have been studied as functions of [sup 10]B tumour to blood ratios and absolute levels. These relationships were examined using the Monte Carlo neutron photon transport code, MCNP, with an ideal 18.4 cm diameter neutron beam incident laterally upon an ellipsoidal neutron photon brain-equivalent model. Mono-energetic beams of 0.025 eV (thermal) and 35 eV (epithermal) were simulated. Increasing the tumour to blood [sup 10]B ratio predictably increases all figures of merit. [sup 10]B concentration was also shown to have a strong bearing on the figures of merit when low levels were present in the system. This is the result of a non-[sup 10]B dependent background dose. At higher levels however, the concentration of [sup 10]B has a diminishing influence. For boron sulphydryl (BSH), little advantage is gained by extending the blood [sup 10]B level beyond 30 ppm, whilst for D, L,-p-boronophenylalanine (BPA) this limit is 10 ppm. Applying the epithermal beam under identical conditions, the therapeutic depth reaches the brain mid-line with a tumour to blood [sup 10]B ratio of only 5.7 for BPA. For BSH, the maximum therapeutic depth reaches the brain mid-line with a tumour to blood ratio of only 1.9 with 30 ppm in the blood. Human data for these compounds are very close to these requirements. (author).

Uranium in coral skeletons determined by epithermal neutron activation analysis

International Nuclear Information System (INIS)

Ohde, S.; Hossain, M.M.M.; Ozaki, H.; Masuzawa, T.

2003-01-01

A simple and non-destructive method has been proposed for the routine determination of uranium by epithermal neutron activation analysis in coral skeletons. Using a cadmium capsule, about 0.1-0.2 g samples were irradiated for 6 hours in the Triga Mark II Reactor. Measurements of γ-ray ( 239 Np via 239 U) were performed with each sample and standard after cooling for about three days. Compared with a non-destructive thermal NAA, the present method was found to improve the sensitivity because it reduced the intense Compton background induced by 24 Na. Uranium in coral standards was determined within 2% of analytical precision. The data obtained for the carbonate standards are mostly consistent with reported values. The present method could be usefully applied to determine uranium contents in fossil corals from the Funafuti Atoll in the Pacific. The distribution of uranium between seawater and coral skeletons is also discussed in order to understand the environmental media in which the coral grew. (author)

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

International Nuclear Information System (INIS)

Molina Avila, J.; Carmolopes, M.

1989-01-01

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

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

Science.gov (United States)

Khattab, K; Sulieman, I

2009-04-01

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

A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR FOR IN VITRO AND IN VIVO TEST OF BORON NEUTRON CAPTURE THERAPY

Directory of Open Access Journals (Sweden)

Nina Fauziah

2015-03-01

Full Text Available Studies were carried out to design a collimator which results in epithermal neutron beam for IN VITRO and IN VIVO of Boron Neutron Capture Therapy (BNCT at the Kartini research reactor by means of Monte Carlo N-Particle (MCNP codes. Reactor within 100 kW of thermal power was used as the neutron source. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 2 cm thick of Bi as γ-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 1 to 5 cm varied aperture size, epithermal neutron beam with maximum flux of 7.65 x 108 n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.76 x 10-13 Gy.cm2.n-1 and 1.32 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.008, and maximum directionality of 0.73. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. When it was assumed that the graphite inside the thermal column was not discharged but only the part which was going to be replaced by the collimator, the performance of the collimator became better within the positive effect from the surrounding graphite that the beam resulted passed all criteria with epithermal neutron flux up to 1.68 x 109 n.cm-2.s-1. Keywords: design, collimator, epithermal neutron beam, BNCT, MCNP, criteria   Telah dilakukan penelitian tentang desain kolimator yang menghasilkan radiasi netron epitermal untuk uji in vitro dan in vivo pada Boron Neutron Capture Therapy (BNCT di Reaktor Riset Kartini dengan menggunakan program Monte

Conceptual design of 30 MeV magnet system used for BNCT epithermal neutron source

International Nuclear Information System (INIS)

Slamet Santosa; Taufik

2015-01-01

Conceptual design of 30 MeV Magnet System Used for BNCT Epithermal Neutron Source has been done based on methods of empirical model of basic equation, experiences of 13 MeV cyclotron magnet design and personal communications. In the field of health, cyclotron can be used as an epithermal neutron source for Boron Neutron Capture Therapy (BNCT). The development of cyclotron producing epithermal neutrons for BNCT has been performed at Kyoto University, of which it produces a proton beam current of 1.1 mA with energy of 30 MeV. With some experiences on 13 MeV cyclotron magnet design, to support BNCT research and development we performed the design studies of 30 MeV cyclotron magnet system, which is one of the main components of the cyclotron for deflecting proton beam into circular trajectory and serves as beam focusing. Results of this study are expected to define the parameters of particular cyclotron magnet. The scope of this study includes the study of the parameters component of the 30 MeV cyclotron and magnet initial parameters. The empirical method of basic equation model is then corroborated by a simulation using Superfish software. Based on the results, a 30 MeV cyclotron magnet for BNCT neutron source enables to be realized with the parameters of B 0 = 1.06 T, frequency RF = 64.733938 ≈ 65 MHz, the external radius of 0.73 m, the radius of the polar = 0.85 m, BH = 1.95 T and a gap hill of 4 cm. Because proton beam current that be needed for BNCT application is very large, then in the calculation it is chosen a great focusing axial νz = 0.630361 which can generate B V = 0.44 T. (author)

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

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

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

International Nuclear Information System (INIS)

Zalan, T.A.

1988-01-01

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

A spin-transport system for a longitudinally polarized epithermal neutron beam

International Nuclear Information System (INIS)

Crawford, B.E.; Bowman, J.D.; Penttilae, S.I.; Roberson, N.R.

2001-01-01

The TRIPLE (Time Reversal and Parity at Low Energies) collaboration uses a polarized epithermal neutron beam and a capture γ-ray detector to study parity violation in neutron-nucleus reactions. In order to preserve the spin polarization of the neutrons as they travel the 60-m path to the target, the beam pipes are wrapped with wire to produce a solenoidal magnetic field of about 10 G along the beam direction. The flanges and bellows between sections of the beam pipe cause gaps in the windings which in turn produce radial fields that can depolarize the neutron spins. A computer code has been developed that numerically evaluates the effect of these gaps on the polarization. A measurement of the neutron depolarization for neutrons in the actual spin-transport system agrees with a calculation of the neutron depolarization for the TRIPLE system. Features that will aid in designing similar spin-transport systems are discussed

VLAD for epithermal neutron scattering experiments at large energy transfers

International Nuclear Information System (INIS)

Tardocchi, M; Gorini, G; Perelli-Cippo, E; Andreani, C; Imberti, S; Pietropaolo, A; Senesi, R; Rhodes, N R; Schooneveld, E M

2006-01-01

The Very Low Angle Detector (VLAD) bank will extend the kinematical region covered by today's epithermal neutron scattering experiments to low momentum transfer ( -1 ) together with large energy transfer 0 -4 0 . In this paper the design of VLAD is presented together with Montecarlo simulations of the detector performances. The results of tests made with prototype VLAD detectors are also presented, confirming the usefulness of the Resonance Detector for measurements at very low scattering angles

Metrology and quality of radiation therapy dosimetry of electron, photon and epithermal neutron beams

Energy Technology Data Exchange (ETDEWEB)

Kosunen, A

1999-08-01

In radiation therapy using electron and photon beams the dosimetry chain consists of several sequential phases starting by the realisation of the dose quantity in the Primary Standard Dosimetry Laboratory and ending to the calculation of the dose to a patient. A similar procedure can be described for the dosimetry of epithermal neutron beams in boron neutron capture therapy (BNCT). To achieve the required accuracy of the dose delivered to a patient the quality of all steps in the dosimetry procedure has to be considered. This work is focused on two items in the dosimetry chains: the determination of the dose in the reference conditions and the evaluation of the accuracy of dose calculation methods. The issues investigated and discussed in detail are: a)the calibration methods of plane parallel ionisation chambers used in electron beam dosimetry, (b) the specification of the critical dosimetric parameter i.e. the ratio of stopping powers for water to air, (S I ?){sup water} {sub air}, in photon beams, (c) the feasibility of the twin ionization chamber technique for dosimetry in epithermal neutron beams applied to BNCT and (d) the determination accuracy of the calculated dose distributions in phantoms in electron, photon, and epithermal neutron beams. The results demonstrate that up to a 3% improvement in the consistency of dose determinations in electron beams is achieved by the calibration of plane parallel ionisation chambers in high energy electron beams instead of calibrations in {sup 60}Co gamma beams. In photon beam dosimetry (S I ?){sup water} {sub air} can be determined with an accuracy of 0.2% using the percentage dose at the 10 cm depth, %dd(10), as a beam specifier. The use of %odd(10) requires the elimination of the electron contamination in the photon beam. By a twin ionisation chamber technique the gamma dose can be determined with uncertainty of 6% (1 standard deviation) and the total neutron dose with an uncertainty of 15 to 20% (1 standard deviation

Metrology and quality of radiation therapy dosimetry of electron, photon and epithermal neutron beams

International Nuclear Information System (INIS)

Kosunen, A.

1999-08-01

In radiation therapy using electron and photon beams the dosimetry chain consists of several sequential phases starting by the realisation of the dose quantity in the Primary Standard Dosimetry Laboratory and ending to the calculation of the dose to a patient. A similar procedure can be described for the dosimetry of epithermal neutron beams in boron neutron capture therapy (BNCT). To achieve the required accuracy of the dose delivered to a patient the quality of all steps in the dosimetry procedure has to be considered. This work is focused on two items in the dosimetry chains: the determination of the dose in the reference conditions and the evaluation of the accuracy of dose calculation methods. The issues investigated and discussed in detail are: a)the calibration methods of plane parallel ionisation chambers used in electron beam dosimetry, (b) the specification of the critical dosimetric parameter i.e. the ratio of stopping powers for water to air, (S I ?) water air , in photon beams, (c) the feasibility of the twin ionization chamber technique for dosimetry in epithermal neutron beams applied to BNCT and (d) the determination accuracy of the calculated dose distributions in phantoms in electron, photon, and epithermal neutron beams. The results demonstrate that up to a 3% improvement in the consistency of dose determinations in electron beams is achieved by the calibration of plane parallel ionisation chambers in high energy electron beams instead of calibrations in 60 Co gamma beams. In photon beam dosimetry (S I ?) water air can be determined with an accuracy of 0.2% using the percentage dose at the 10 cm depth, %dd(10), as a beam specifier. The use of %odd(10) requires the elimination of the electron contamination in the photon beam. By a twin ionisation chamber technique the gamma dose can be determined with uncertainty of 6% (1 standard deviation) and the total neutron dose with an uncertainty of 15 to 20% (1 standard deviation). To improve the accuracy

Epithermal neutron beam for BNCT research at the Washington State University TRIGA research reactor

International Nuclear Information System (INIS)

Nigg, D.W.; Venhuizen, J.R.; Wheeler, F.J.; Wemple, C.A.; Tripard, G.E.; Gavin, P.R.

2000-01-01

A new epithermal-neutron beam facility for BNCT (Boron Neutron Capture Therapy) research and boronated agent screening in animal models is in the final stages of construction at Washington State University (WSU). A key distinguishing feature of the design is the incorporation of a new, high-efficiency, neutron moderating and filtering material, Fluental, developed by the Technical Research Centre of Finland. An additional key feature is the provision for adjustable filter-moderator thickness to systematically explore the radiobiological consequences of increasing the fast-neutron contamination above the nominal value associated with the baseline system. (author)

Characteristics of thermal neutron calibration fields using a graphite pile

International Nuclear Information System (INIS)

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

2005-03-01

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

Depth-dose evaluation for lung and pancreas cancer treatment by BNCT using an epithermal neutron beam

International Nuclear Information System (INIS)

Matsumoto, Tetsuo; Fukushima, Yuji

2000-01-01

The depth-dose distributions were evaluated for possible treatment of both lung and pancreas cancers using an epithermal neutron beam. The MCNP calculations showed that physical dose in tumors were 6 and 7 Gy/h, respectively, for lung and pancreas, attaining an epithermal neutron flux of 5x10 8 ncm -2 s -1 . The boron concentrations were assumed at 100 ppm and 30 ppm, respectively, for lung and pancreas tumors and normal tissues contains 1/10 tumor concentrations. The dose ratios of tumor to normal tissue were 2.5 and 2.4, respectively, for lung and pancreas. The dose evaluation suggests that BNCT could be applied for both lung and pancreas cancer treatment. (author)

Elemental analysis of brazing alloy samples by neutron activation technique

International Nuclear Information System (INIS)

Eissa, E.A.; Rofail, N.B.; Hassan, A.M.; El-Shershaby, A.; Walley El-Dine, N.

1996-01-01

Two brazing alloy samples (C P 2 and C P 3 ) have been investigated by Neutron activation analysis (NAA) technique in order to identify and estimate their constituent elements. The pneumatic irradiation rabbit system (PIRS), installed at the first egyptian research reactor (ETRR-1) was used for short-time irradiation (30 s) with a thermal neutron flux of 1.6 x 10 1 1 n/cm 2 /s in the reactor reflector, where the thermal to epithermal neutron flux ratio is 106. Long-time irradiation (48 hours) was performed at reactor core periphery with thermal neutron flux of 3.34 x 10 1 2 n/cm 2 /s, and thermal to epithermal neutron flux ratio of 79. Activation by epithermal neutrons was taken into account for the (1/v) and resonance neutron absorption in both methods. A hyper pure germanium detection system was used for gamma-ray acquisitions. The concentration values of Al, Cr, Fe, Co, Cu, Zn, Se, Ag and Sb were estimated as percentages of the sample weight and compared with reported values. 1 tab

Elemental analysis of brazing alloy samples by neutron activation technique

Energy Technology Data Exchange (ETDEWEB)

Eissa, E A; Rofail, N B; Hassan, A M [Reactor and Neutron physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); El-Shershaby, A; Walley El-Dine, N [Physics Department, Faculty of Girls, Ain Shams Universty, Cairo (Egypt)

1997-12-31

Two brazing alloy samples (C P{sup 2} and C P{sup 3}) have been investigated by Neutron activation analysis (NAA) technique in order to identify and estimate their constituent elements. The pneumatic irradiation rabbit system (PIRS), installed at the first egyptian research reactor (ETRR-1) was used for short-time irradiation (30 s) with a thermal neutron flux of 1.6 x 10{sup 1}1 n/cm{sup 2}/s in the reactor reflector, where the thermal to epithermal neutron flux ratio is 106. Long-time irradiation (48 hours) was performed at reactor core periphery with thermal neutron flux of 3.34 x 10{sup 1}2 n/cm{sup 2}/s, and thermal to epithermal neutron flux ratio of 79. Activation by epithermal neutrons was taken into account for the (1/v) and resonance neutron absorption in both methods. A hyper pure germanium detection system was used for gamma-ray acquisitions. The concentration values of Al, Cr, Fe, Co, Cu, Zn, Se, Ag and Sb were estimated as percentages of the sample weight and compared with reported values. 1 tab.

Development of the Very Low Angle Detector (VLAD) for detection of epithermal neutrons at low momentum transfers

International Nuclear Information System (INIS)

Tardocchi, M.; Andreani, C.; Cremonesi, O.; Gorini, G.; Perelli-Cippo, E.; Pietropaolo, A.; Rhodes, N.; Schooneveld, E.; Senesi, R.

2006-01-01

New perspectives for epithermal neutron spectroscopy are opened up by the recent development of new instrumentation for inverse geometry time of flight spectrometers at pulsed neutron sources. One example is the Very Low Angle Detector (VLAD) bank which will be installed as an upgrade of the VESUVIO neutron spectrometer, at the ISIS pulsed neutron source. VLAD is developed for detecting epithermal neutrons in the 1-100 eV energy range at very low scattering angles (l deg. - 5 deg.). VLAD will extend the kinematical region covered by today's neutron scattering experiments to the region of low wave vector transfers ( -1 ) and high energy transfers (>1 eV). Accessing such kinematical region will allow new experimental studies in condensed matter systems. The neutron detection is based on Resonance Detectors (RD), which consist of the combination of a resonance foil used as neutron-to-gamma converter and a photon detector. The results obtained with a prototype VLAD detector confirm the potential of this kind of experiments at scattering angles as low as 2 deg. - 5 deg. GEANT4 simulations are used to address issues, such as detector cross talk, which arise with the construction of compact RD arrays

Physical parameters and biological effects of the LVR-15 epithermal neutron beam

International Nuclear Information System (INIS)

Burian, J.; Marek, M.; Rejchrt, J.; Viererbl, L.; Gambarini, G.; Mares, V.; Vanossi, E.; Judas, L.

2006-01-01

Monitoring of the physical and biological properties of the epithermal neutron beam constructed at the multipurpose LVR-15 nuclear reactor for NCT therapy of brain tumors showed that its physical and biological properties are stable in time and independent on an ad hoc reconfiguration of the reactor core before its therapeutic use. Physical parameters were monitored by measurement of the neutron spectrum, neutron profile, fast neutron kerma rate in tissue and photon absorbed dose, the gel dosimetry was used with the group of standard measurement methods. The RBE of the beam, as evaluated by 3 different biological models, including mouse intestine crypt regeneration assay, germinative zones of the immature rat brain and C6 glioma cells in culture, ranged from 1.70 to 1.99. (author)

Method of assaying uranium with prompt fission and thermal neutron borehole logging adjusted by borehole physical characteristics

International Nuclear Information System (INIS)

Barnard, R.W.; Jensen, D.H.

1982-01-01

Uranium formations are assayed by prompt fission neutron logging techniques. The uranium in the formation is proportional to the ratio of epithermal counts to thermal or eqithermal dieaway. Various calibration factors enhance the accuracy of the measurement

Reactor beam calculations to determine optimum delivery of epithermal neutrons for treatment of brain tumors

International Nuclear Information System (INIS)

Wheeler, F.J.; Nigg, D.W.; Capala, J.

1997-01-01

Studies were performed to assess theoretical tumor control probability (TCP) for brain-tumor treatment with boron neutron capture therapy (BNCT) using epithermal neutron sources from reactors. The existing epithermal-neutron beams at the Brookhaven Medical Research Reactor Facility (BMRR), the Petten High Flux Reactor Facility (HWR) and the Finnish Research Reactor 1 (FIR1) have been analyzed and characterized using common analytical and measurement methods allowing for this inter-comparison. Each of these three facilities is unique and each offers an advantage in some aspect of BNCT, but none of these existing facilities excel in all neutron-beam attributes as related to BNCT. A comparison is therefore also shown for a near-optimum reactor beam which does not currently exist but which would be feasible with existing technology. This hypothetical beam is designated BNCT-1 and has a spectrum similar to the FIR-1, the mono-directionality of the HFR and the intensity of the BMRR. A beam very similar to the BNCT-1 could perhaps be achieved with modification of the BMRR, HFR, or FIR, and could certainly be realized in a new facility with today's technology

Photon detectors for epithermal neutron scattering at high-ω and low-q

International Nuclear Information System (INIS)

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

2004-01-01

Inelastic epithermal neutron scattering at high energy (ℎω≥1 eV) and low wave vector (q≤10 A -1 ) transfers is the unique technique for the investigation of high-energy excitations in a variety of systems, ranging from magnetic materials to semiconductors. The key issue in order to make these measurements feasible on inverse geometry spectrometers, is to develop suitable detection systems for neutrons in the energy range 1-100 eV. The Resonance Detector Spectrometer configuration has to be considered as the most promising approach for electron Volt neutron spectroscopy. This configuration will be employed in the new low angle detector bank, VLAD, planned for VESUVIO spectrometer operating at ISIS source

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-15

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

Determination of uranium in tree bark samples by epithermal neutron activation analysis

International Nuclear Information System (INIS)

Lima, Nicole Pereira de; Saiki, Mitiko

2017-01-01

In this study uranium (U) concentrations were determined in certified reference materials (CRMs) and in tree bark samples collected in 'Cidade Universitaria Armando de Salles Oliveira' (CUASO) USP, Sao Paulo, SP, Brazil). The barks were collected from different species namely Poincianella pluviosa and Tipuana tipu. These bark samples were cleaned, dried, grated and milled for the analyses by epithermal neutron activation analysis method (ENAA). This method consists on irradiating samples and U standard in IEAR1 nuclear reactor with thermal neutron flux of 1:9 x 10 12 n cm -2 s -1 during 40 to 60 seconds depending on the samples matrices. The samples and standard were measured by gamma ray spectroscopy. U was identified by the peak of 74.66 keV of 239 U with half life of 23.47 minutes. Concentration of U was calculated by comparative method. For analytical quality control of U results, certified reference materials were analysed. Results obtained for CRMs presented good precision and accuracy, with |Z score| <= 0.39. Uranium concentrations in tree barks varied from 83.1 to 627.6 ng g - 1 and the relative standard deviations of these results ranged from 1.8 to 10%. (author)

Determination of uranium in tree bark samples by epithermal neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Lima, Nicole Pereira de; Saiki, Mitiko, E-mail: mitiko@ipen.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

In this study uranium (U) concentrations were determined in certified reference materials (CRMs) and in tree bark samples collected in 'Cidade Universitaria Armando de Salles Oliveira' (CUASO) USP, Sao Paulo, SP, Brazil). The barks were collected from different species namely Poincianella pluviosa and Tipuana tipu. These bark samples were cleaned, dried, grated and milled for the analyses by epithermal neutron activation analysis method (ENAA). This method consists on irradiating samples and U standard in IEAR1 nuclear reactor with thermal neutron flux of 1:9 x 10{sup 12} n cm{sup -2} s{sup -1} during 40 to 60 seconds depending on the samples matrices. The samples and standard were measured by gamma ray spectroscopy. U was identified by the peak of 74.66 keV of {sup 239}U with half life of 23.47 minutes. Concentration of U was calculated by comparative method. For analytical quality control of U results, certified reference materials were analysed. Results obtained for CRMs presented good precision and accuracy, with |Z score| <= 0.39. Uranium concentrations in tree barks varied from 83.1 to 627.6 ng g{sup -} {sup 1} and the relative standard deviations of these results ranged from 1.8 to 10%. (author)

Boron neutron capture therapy (BNCT) for glioblastoma multiforme using the epithermal neutron beam at the Brookhaven Medical Research Reactor

International Nuclear Information System (INIS)

Capala, J.; Diaz, A.Z.; Chadha, M.

1997-01-01

The abstract describes evaluation of boron neutron capture therapy (BNCT) for two groups of glioblastoma multiforme patients. From September 1994 to February 1996 15 patients have been treated. In September 1997 another 34 patients were examined. Authors determined a safe starting dose for BNCT using epithermal neutrons and BPA-F. They have also evaluated adverse effects of BNCT at this starting dose. Therapeutic effectiveness of this starting dose has been evaluated. No significant side effects from BPA-F infusion or BNCT treatment were observed in normal brains

Boron neutron capture therapy (BNCT) for glioblastoma multiforme using the epithermal neutron beam at the Brookhaven Medical Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Capala, J. [Brookhaven National Lab., Upton, NY (United States); Diaz, A.Z.; Chadha, M. [Univ. Hospital, State Univ. of New York, NY (United States)] [and others

1997-12-31

The abstract describes evaluation of boron neutron capture therapy (BNCT) for two groups of glioblastoma multiforme patients. From September 1994 to February 1996 15 patients have been treated. In September 1997 another 34 patients were examined. Authors determined a safe starting dose for BNCT using epithermal neutrons and BPA-F. They have also evaluated adverse effects of BNCT at this starting dose. Therapeutic effectiveness of this starting dose has been evaluated. No significant side effects from BPA-F infusion or BNCT treatment were observed in normal brains.

Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

1987-02-01

Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

Method and apparatus for epithermal neutron decay logging

International Nuclear Information System (INIS)

Nelligan, W.B.

1998-01-01

The nature of hydrogenous fluids filling the pore spaces in formations surrounding a well bore are determined by irradiating the formations with bursts of high energy neutrons and using the ratio of time-dependent parameters related to the decay of epithermal neutron populations above two different energy levels to provide values indicative of the kind of fluid present, independent of porosity. The measurement above the higher of these energy levels, appr. 0.4eV, indicates the existence of hydrogenous fluid, water and hydrocarbons in the formations. The lower threshold measurement, above appr. 0.15eV, is indicative of the structure of the molecule in which the hydrogen molecule is bound and of the porosity. A pulsed neutron generator in a sonde irradiates the formations with 14meV neutrons and a pair of detectors, one shielded by cadmium, to establish the 0.4eV threshold energy level and the other, shielded by gadolinium, to establish the 0.15eV threshold energy level. Time related parameters of the count rate information, e.g. the decay constants, provided by each detector are derived. The ratio of the respective parameters is used to identify the type of fluid in the information, independent of porosity. The method and apparatus can determine water saturation, i.e. percentage of water in the formation fluid and can identify specific types of hydrocarbons under 100% hydrocarbon saturation conditions. 8 figs

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

International Nuclear Information System (INIS)

Shaaban, I.

2009-11-01

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

Cosmic-ray neutron transport at a forest field site: the sensitivity to various environmental conditions with focus on biomass and canopy interception

Science.gov (United States)

Andreasen, Mie; Jensen, Karsten H.; Desilets, Darin; Zreda, Marek; Bogena, Heye R.; Looms, Majken C.

2017-04-01

Cosmic-ray neutron intensity is inversely correlated to all hydrogen present in the upper decimeters of the subsurface and the first few hectometers of the atmosphere above the ground surface. This correlation forms the base of the cosmic-ray neutron soil moisture estimation method. The method is, however, complicated by the fact that several hydrogen pools other than soil moisture affect the neutron intensity. In order to improve the cosmic-ray neutron soil moisture estimation method and explore the potential for additional applications, knowledge about the environmental effect on cosmic-ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different parameters describing the subsurface to match measured height profiles and time series of thermal and epithermal neutron intensities at a field site in Denmark. Overall, modeled thermal and epithermal neutron intensities are in satisfactory agreement with measurements; however, the choice of forest canopy conceptualization is found to be significant. Modeling results show that the effect of canopy interception, soil chemistry and dry bulk density of litter and mineral soil on neutron intensity is small. On the other hand, the neutron intensity decreases significantly with added litter-layer thickness, especially for epithermal neutron energies. Forest biomass also has a significant influence on the neutron intensity height profiles at the examined field site, altering both the shape of the profiles and the ground-level thermal-to-epithermal neutron ratio. This ratio increases with increasing amounts of biomass, and was confirmed by measurements from three sites representing agricultural, heathland and forest land cover. A much smaller effect of canopy interception on the ground

The monostandard method in thermal neutron activation analysis of geological, biological and environmental materials

International Nuclear Information System (INIS)

Alian, A.; Djingova, R.G.; Kroener, B.; Sansoni, B.

1984-01-01

A simple method is described for instrumental multielement thermal neutron activation analysis using a monostandard. For geological and air dust samples, iron is used as a comparator, while sodium has advantages for biological materials. To test the capabilities of this method, the values of the effective cross sections of the 23 elements determined were evaluated in a reactor site with an almost pure thermal neutron flux of about 9x10 12 nxcm -2 xs -1 and an epithermal neutron contribution of less than 0.03%. The values obtained were found to agree mostly well with the best literature values of thermal neutron cross sections. The results of an analysis by activation in the same site agree well with the relative method using multielement standards and for several standard reference materials with certified element contents. A comparison of the element contents obtained by the monostandard and relative methods together with corresponding precisions and accuracies is given. (orig.) [de

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

Science.gov (United States)

Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

2016-12-01

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract.

Investigations on the comparator technique used in epithermal neutron activation analysis

International Nuclear Information System (INIS)

Bereznai, T.; Bodizs, D.; Keoemley, G.

1977-01-01

The possible extension of the comparator technique of reactor neutron activation analysis into the field of epithermal neutron activation has been investigated. Ruthenium was used for multi-isotopic comparator. Experiments show that conversion of the so-called reference k-factors - determined by irradiation with reactor neutrons - into ksup(epi)-factors usable at activation under cadmium filter, can be evaluated with fair accuracy. Sources and extent of errors and their contribution to the final error of analysis are discussed. For equal irradiation and counting times advantage of ENAA for several elements is obvious: the much lower background activity permitted the sample to be measured closer to the detector, under better geometry conditions, consequently, permitting several elements to be determined quantitatively. The number of elements determined and the sensitivity of the method are much dependent on the experimental conditions, especially on the composition of the sample, on the PHIsub(e) value, the irradiation time and the efficiency of the Ge(Li) detector. (T.G.)

Measurements of neutron flux in the RA reactor

International Nuclear Information System (INIS)

Raisic, N.

1961-12-01

This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire [sr

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

International Nuclear Information System (INIS)

Erben, O.

1980-01-01

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

Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

Energy Technology Data Exchange (ETDEWEB)

Gryzinski, M.A.; Maciak, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

2015-07-01

planned to create fully equipped complex facility possible to perform various experiments on the intensive neutron beam. Epithermal neutron beam enables development across the full spectrum of materials research for example shielding concrete tests or electronic devices construction improvement. Due to recent reports on the construction of the accelerator for the Boron Neutron Capture Therapy (BNCT) it has the opportunity to become useful and successful method in the fight against brain and other types of cancers not treated with well known medical methods. In Europe there is no such epithermal neutron source which could be used throughout the year for training and research for scientist working on BNCT what makes the stand unique in Europe. Also our research group which specializes in mixed radiation dosimetry around nuclear and medical facilities would be able to carry out research on new detectors and methods of measurements for radiological protection and in-beam (therapeutic) dosimetry. Another group of scientists from National Centre for Nuclear Research, where MARIA research reactor is located, is involved in research of gamma detector systems. There is an idea to develop Prompt-gamma Single Photon Emission Computed Tomography (Pg- SPECT). This method could be used as imaging system for compounds emitting gamma rays after nuclear reaction with thermal neutrons e.g. for boron concentration in BNCT. Inside the room, where H2 channel is located, there is another horizontal channel - H1 which is also unused. Simultaneously with the construction of the H2 stand it will be possible to create special pneumatic horizontal mail inside the H1 channel for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. It might expand the scope of research at the planned neutron station. Secondly it is planned to equip both stands with moveable positioning system, video system and facilities to perform animal experiments (anaesthesia, vital

Neutron beams for therapy

International Nuclear Information System (INIS)

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

2011-01-01

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

SPLET - A program for calculating the space-lethargy distribution of epithermal neutrons in a reactor lattice cell

International Nuclear Information System (INIS)

Matausek, M.V.; Zmijatevic, I.

1981-01-01

A procedure to solve the space-single-lethargy dependent transport equation for epithermal neutrons in a cylindricised multi-region reactor lattice cell has been developed and proposed in the earlier papers. Here, the computational algorithm is comprised and the computing program SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, as well as the related integral quantities as reaction rates and resonance integrals, is described. (author)

Development of a new electronic neutron imaging system

CERN Document Server

Brenizer, J S; Gibbs, K M; Mengers, P; Stebbings, C T; Polansky, D; Rogerson, D J

1999-01-01

An electronic neutron imaging camera system was developed for use with thermal, epithermal, and fast neutrons in applications that include nondestructive inspection of explosives, corrosion, turbine blades, electronics, low Z components, etc. The neutron images are expected to provide information to supplement that available from X-ray tests. The primary camera image area was a 30x30 cm field-of-view with a spatial resolution approaching 1.6 line pairs/mm (lp/mm). The camera had a remotely changeable second lens to limit the field-of-view to 7.6x7.6 cm for high spatial resolution (at least 4 lp/mm) thermal neutron imaging, but neutron and light scatter will limit resolution for fast neutrons to about 0.5 lp/mm. Remote focus capability enhanced camera set-up for optimum operation. The 75 dB dynamic range camera system included sup 6 Li-based screens for imaging of thermal and epithermal neutrons and ZnS(Ag)-based screens for fast neutron imaging. The fast optics was input to a Super S-25 Gen II image intensifi...

The effect of cadmium shielding on the spatial neutron flux distribution inside one of the outer irradiation sites

International Nuclear Information System (INIS)

Shaaban, I.

2009-06-01

A permanent epithermal neutron irradiation facility was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using the cadmium (cylindrical vial 1.0 mm in thickness, 38.50 mm in diameter and 180 mm in length) as thermal neutron shielding material, for a permanent epithermal neutron activation analysis (ENAA). This site was designed by shielding the internal surface of the aluminum tube of the first outer irradiation site in the MNSR reactor. I was used the activation detectors 0.1143% Au-Al alloy foils with 0.1 mm thickness and 2.0 mm diameter for measurement the thermal neutron flux, epithermal and R c d=A b are/A c over ratio in the outer irradiation site. Distribution of the thermal neutron flux in the outer irradiation capsule has been found numerically using MCNP-4C code with and without cadmium shield, and experimentally by irradiating five copper wires using the outer irradiation capsule. Good agreements were obtained between the calculated and the measured results. (author)

Spherical Harmonics Treatment of Epithermal Neutron Spectra in Reactor lattices

International Nuclear Information System (INIS)

Matausek, M.V.

1972-04-01

A procedure has been developed to solve the slowing down transport equation for neutrons in a cylindrized reactor lattice cell. Treating the anisotropy of the epithermal neutron flux by the spherical harmonics formalism, which reduces the space-angle-lethargy-dependent transport equation to the matrix integrodifferential equation in space and lethargy, and replacing the lethargy transfer integrals by finite-difference forms, a set of matrix ordinary differential equations, with lethargy and space dependent coefficients, is obtained. In the resonance region this set takes a lower block triangular form and can be directly solved by forward block substitution; in the lethargy range, where the fast fission effects have to be considered, the iterative procedure is introduced. A simple and efficient approximation is then proposed, making possible the analytical solution for the spatial dependence of the spherical harmonics flux moments. The proposed procedure has been numerically examined and approved. Some typical results are presented and discussed. (author)

Initial Experimental Verification of the Neutron Beam Modeling for the LBNL BNCT Facility

International Nuclear Information System (INIS)

Bleuel, D.L.; Chu, W.T.; Donahue, R.J.; Ludewigt, B.A.; McDonald, R.J.; Smith, A.R.; Stone, N.A.; Vuji, J.

1999-01-01

In preparation for future clinical BNCT trials, neutron production via the 7Li(p,n) reaction as well as subsequent moderation to produce epithermal neutrons have been studied. Proper design of a moderator and filter assembly is crucial in producing an optimal epithermal neutron spectrum for brain tumor treatments. Based on in-phantom figures-of-merit,desirable assemblies have been identified. Experiments were performed at the Lawrence Berkeley National Laboratory's 88-inch cyclotron to characterize epithermal neutron beams created using several microampere of 2.5 MeV protons on a lithium target. The neutron moderating assembly consisted of Al/AlF3 and Teflon, with a lead reflector to produce an epithermal spectrum strongly peaked at 10-20 keV. The thermal neutron fluence was measured as a function of depth in a cubic lucite head phantom by neutron activation in gold foils. Portions of the neutron spectrum were measured by in-air activation of six cadmium-covered materials (Au, Mn, In, Cu, Co, W) with high epithermal neutron absorption resonances. The results are reasonably reproduced in Monte Carlo computational models, confirming their validity

Characterisation of the TAPIRO BNCT epithermal facility

Energy Technology Data Exchange (ETDEWEB)

Burn, K. W. [FIS-NUC, ENEA, Via Martiri di Montesole 4, Bologna (Italy); Colli, V. [Dept. of Physics of Univ., INFN, Via Celoria 16, I-20133 Milano (Italy); Curzio, G.; D' Errico, F. [DIMNP, Univ. of Pisa, Via Diotisalvi 2, I-56126 Pisa (Italy); Gambarini, G. [Dept. of Physics of Univ., INFN, Via Celoria 16, I-20133 Milano (Italy); Rosi, G. [FIS-ION, ENEA, Casaccia, Via Anguillarese 301, I-00060 Santa Maria di Galeria, Roma (Italy); Scolari, L. [Dept. of Physics of Univ., INFN, Via Celoria 16, I-20133 Milano (Italy)

2004-07-01

A collimated epithermal beam for boron neutron capture therapy (BNCT) research has been designed and built at the TAPIRO fast research reactor. A complete experimental characterisation of the radiation field in the irradiation chamber has been performed, to verify agreement with IAEA requirements. Slow neutron fluxes have been measured by means of an activation technique and with thermoluminescent detectors (TLDs). The fast neutron dose has been determined with gel dosemeters, while the fast neutron spectrum has been acquired by means of a neutron spectrometer based on superheated drop detectors. The gamma-dose has been measured with gel dosemeters and TLDs. For an independent verification of the experimental results, fluxes, doses and neutron spectra have been calculated with Monte Carlo simulations using the codes MCNP4B and MCNPX 2.1.5 with the direct statistical approach (DSA). The results obtained confirm that the epithermal beams achievable at TAPIRO are of suitable quality for BNCT purposes. (authors)

An Emergency Dosimeter for Neutrons

Energy Technology Data Exchange (ETDEWEB)

Braun, J; Nilsson, R

1960-05-15

A neutron dosimeter suitable for single emergency exposures is described. The dosimeter is furnished with detectors for thermal, epi-thermal and fast neutrons. This means that three of the constants by which the spectrum of the incident neutron flux is approximated, can be determined. The dose calculated from these approximated spectra is compared to the dose from spectra obtained in different standard spectra of types which may be expected in a radiation accident.

The PTB thermal neutron reference field at GeNF

International Nuclear Information System (INIS)

Boettger, R.; Friedrich, H.; Janssen, H.

2004-01-01

The experimental setup and procedure for the characterization of the thermal neutron reference field established at the Geesthacht neutron facility (GeNF) of the GKSS is described. The neutron beam, free in air, with a maximum flux of 10 6 s -1 , can easily be reduced to less than 10 4 s -1 by using a diaphragm variable in size and without changing the beam divergence. Also, the spectral distribution with a mean energy of 45 meV, measured by time-of-flight over a 6.6 m long flight path, is independent of the beam current chosen. In the 2002/2003 experiments reported here, a 6 Li glass detector was employed to determine the absolute beam current and to calibrate the 3 He transmission beam monitor. In addition, activation measurements of gold foils were carried out at a reduced beam divergence. The results agree within ±2%. Furthermore, the beam is characterized by a low gamma background intensity and a negligible fraction of epithermal neutrons. Irradiations in combination with a scanner device to achieve a homogeneously illuminated scan field have shown that the thermal beam is well suited for dosemeter development and for the calibration of radiation protection instruments. (orig.)

The PTB thermal neutron reference field at GeNF

Energy Technology Data Exchange (ETDEWEB)

Boettger, R.; Friedrich, H.; Janssen, H.

2004-07-01

The experimental setup and procedure for the characterization of the thermal neutron reference field established at the Geesthacht neutron facility (GeNF) of the GKSS is described. The neutron beam, free in air, with a maximum flux of 10{sup 6} s{sup -1}, can easily be reduced to less than 10{sup 4} s{sup -1} by using a diaphragm variable in size and without changing the beam divergence. Also, the spectral distribution with a mean energy of 45 meV, measured by time-of-flight over a 6.6 m long flight path, is independent of the beam current chosen. In the 2002/2003 experiments reported here, a {sup 6}Li glass detector was employed to determine the absolute beam current and to calibrate the {sup 3}He transmission beam monitor. In addition, activation measurements of gold foils were carried out at a reduced beam divergence. The results agree within {+-}2%. Furthermore, the beam is characterized by a low gamma background intensity and a negligible fraction of epithermal neutrons. Irradiations in combination with a scanner device to achieve a homogeneously illuminated scan field have shown that the thermal beam is well suited for dosemeter development and for the calibration of radiation protection instruments. (orig.)

Radiobiology studies for the evaluation of epithermal neutron beams used for BNCT

International Nuclear Information System (INIS)

Green, S.; Jones, B.; Mill, A.J.

2006-01-01

This paper outlines our plans for a study to establish the radiobiological effectiveness of the various mixes of radiation components present in an epithermal neutron beam designed for BNCT and to incorporate these data into clinical protocols for the treatment of malignant glioma. This is a description of work which is funded and just now beginning in Birmingham so no results can be presented. Our project will involve a combination of experimental measurements carried out in Birmingham and in Boston and mathematical modelling carried out in Birmingham. Despite all the extant in-vitro and in-vivo work, there is no widely accepted method to determine biological effect by accounting for variations in beam component mix, dose rate and treatment fractionation for disparate from the various BNCT centres. The objectives of this study are: To develop a cell-based radiobiology protocol to provide essential data on safety and efficacy of beams for Boron Neutron Capture Therapy (BNCT) in advance of clinical trials. To exploit the facilities at Massachusetts Institute of Technology for variable dose-rate epithermal irradiations to validate the above protocol. To develop mathematical models of this radiobiological system that can be used to inform decisions on dose selection, fractionation schedules, BNCT use as supplementary boosts or for re-treatment of recurrent cancers. To provide fundamental data relevant to the understanding of the radiobiology of simultaneous mixed high-and low-LET radiations over a clinically relevant dose-range. (author)

The resonant detector and its application to epithermal neutron spectroscopy

International Nuclear Information System (INIS)

Gorini, G.; Perelli-Cippo, E.; Tardocchi, M.; Andreani, C.; D'Angelo, A.; Pietropaolo, A.; Senesi, R.; Imberti, S.; Bracco, A.; Previtali, E.; Pessina, G.; Rhodes, N.J.; Schooneveld, E.M.

2004-01-01

New perspectives for epithermal neutron spectroscopy are being opened by the development of the resonant detector (RD) and its use on inverse geometry time of flight spectrometers at spallation sources. The RD was first proposed in the 1980s and was recently brought to a performance level exceeding conventional neutron-sensitive Li-glass scintillator detectors. It features a photon counter coupled to a neutron analyzer foil. Resonant neutron absorption in the foil results in the emission of prompt gamma rays that are detected in the photon counter. The dimensions of the RD set the spatial resolution that can be achieved, ranging from a fraction of a cm to several cm. It can thus be tailored to the construction of detector arrays of different geometry. The main results of the research on this kind of detector are reported leading to the present optimized RD design based on a combination of YAP scintillation photon counter and uranium or gold analyzer foils. This detector has already been selected for application in the upgrade of the VESUVIO spectrometer on ISIS. A special application is the Very Low Angle Detector (VLAD) bank, which will extend the kinematical region for neutron scattering to low momentum transfer ( -1 ) whilst still keeping energy transfer >1 eV, thus allowing new experimental studies in condensed matter systems. The first results of tests made with prototype VLAD detectors are presented, confirming the usefulness of the RD for measurements at scattering angles as low as 2-5 deg

Neutron importance calculation in an equivalent cell using the age approximation and differential thermalization models. Determination of the cross section sensitivity to the parameters of a differential model in the thermal range

International Nuclear Information System (INIS)

Sidorenko, V.D.

1978-01-01

The equations are discussed for calculating the importance of neutron function in heterogeneous media obtained with the integral transport theory method. The thermalization effect in the thermal range is described using the differential model. The account of neutron slowing-down in the epithermal range is accomplished in the age approximation. The fast range is described in the 3-group approximation. On the basis of the equations derived the share of delayed neutrons and lifetimes of prompt neutrons are calculated and compared with available experimental data. In the thermal range the sensitivity of cross sections to some parameters of the differential model is analyzed for reactor cells typical for WWER type reactor cores. The models and approximations used are found to be adequate for the calculations

Determination of selenium in Ni + Co concentrates applying epithermal neutron activation analysis

International Nuclear Information System (INIS)

Capote Rodriguez, G.; Perez Sayaz, G.; Hernandez Rivero, A.; Moreno Bermudez, J.; Ribeiro Guevara, S.; Arribere, M.A.; Molina Insfran, J.

1996-01-01

Concentration of Se in Ni + Co concentrates obtained in nickel industry has to be determined as that is a quality control requirement for its commercialization. At present, analysis of Se, specially at a minor and trace levels is relatively complicated and destructive procedures are frequently required. In this work determination of Se by epithermal neutron activation analysis (ENAA) in 17 samples of nickel industry was investigated. Application of ENAA allowed nondestructive determination of Se concentration down to ppm level in spite of presence of high Co, Fe, Ni, and Cr contents in the samples

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

Energy Technology Data Exchange (ETDEWEB)

Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Trkov, Andrej [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); IAEA, Vienna International Centre, PO Box 100, A-1400 Vienna (Austria); Jaćimović, Radojko [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Gregoire, Gilles; Destouches, Christophe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St. Paul-Lez-Durance (France)

2016-12-21

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

International Nuclear Information System (INIS)

Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

2016-01-01

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

Development of an anthropomorfic simulator for simulation and measurements of neutron dose and flux the facility for BNCT studies

International Nuclear Information System (INIS)

Muniz, Rafael Oliveira Rondon

2010-01-01

IPEN facility for researches in BNCT (Boron Neutron Capture Therapy) uses IEA-R1 reactor's irradiation channel number 3, where there is a mixed radiation field - neutrons and gamma. The researches in progress require the radiation fields, in the position of the irradiation of sample, to have in its composition maximized thermal neutrons component and minimized, fast and epithermal neutron flux and gamma radiation. This work was developed with the objective of evaluating whether the present radiation field in the facility is suitable for BNCT researches. In order to achieve this objective, a methodology for the dosimetry of thermal neutrons and gamma radiation in mixed fields of high doses, which was not available in IPEN, was implemented in the Center of Nuclear Engineering of IPEN, by using thermoluminescent dosimeters - TLDs 400, 600 and 700. For the measurements of thermal and epithermal neutron flux, activation detectors of gold were used applying the cadmium ratio technique. A cylindrical phantom composed by acrylic discs was developed and tested in the facility and the DOT 3.5. computational code was used in order to obtain theoretical values of neutron flux and the dose along phantom. In the position corresponding to about half the length of the cylinder of the phantom, the following values were obtained: thermal neutron flux (2,52 ± 0,06).10 8 n/cm 2 s, epithermal neutron flux (6,17 ± 0,26).10 7 .10 6 n/cm 2 s, absorbed dose due to thermal neutrons (4,2 ± 1,8)Gy and (10,1 ± 1,3)Gy due to gamma radiation. The obtained values show that the fluxes of thermal and epithermal neutrons flux are appropriate for studies in BNCT, however, the dose due to gamma radiation is high, indicating that the facility should be improved. (author)

Comparison of Cadmium-Zinc-Telluride semiconductor and Yttrium-Aluminum-Perovskite scintillator as photon detectors for epithermal neutron spectroscopy

International Nuclear Information System (INIS)

Tardocchi, M.; Pietropaolo, A.; Andreani, C.; Gorini, G.; Imberti, S.; Perelli-Cippo, E.; Senesi, R.; Rhodes, N.; Schooneveld, E.M.

2006-01-01

The range of applications of epithermal neutron scattering experiments has been recently extended by the development of the Resonance Detector. In a Resonance Detector, resonant neutron absorption in an analyzer foil results in prompt emission of X- and γ-rays which are detected by a photon counter. Several combinations of analyzer foils and photon detectors have been studied and tested over the years and best results have been obtained with the combination of a natural uranium and (i) Cadmium-Zinc-Telluride (CZT) semiconductor (ii) Yttrium-Aluminum-Perovskite (YAP) scintillators. Here we compare the performance of the CZT semiconductor and YAP scintillator as Resonance Detector units. Two Resonance Detector prototypes made of natural uranium foil viewed by CZT and YAP were tested on the VESUVIO spectrometer at the ISIS spallation neutron source. The results show that both YAP and CZT can be used to detect epithermal neutrons in the energy range from 1 up to 66 eV. It was found that the signal-to-background ratio of the measurement can significantly be improved by raising the lower level discrimination threshold on the γ energy to about 600 keV. The advantages/disadvantages of the choice of a Resonance Detector based on YAP or CZT are discussed together with some potential applications

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

International Nuclear Information System (INIS)

Thiagu Supramaniam

2007-01-01

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

Dosimetry methods in boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D. [Universita degli Studi di Milano, Department of Physics, Via Festa del Patrono 7, 20122 Milano (Italy); Agosteo, S.; Barcaglioni, L. [Istituto Nazionale di Fisica Nucleare, Milano (Italy); Campi, F.; Garlati, L. [Politecnico di Milano, Energy Department, Piazza Leonardo Da Vinci 32, 20133 Milano (Italy); De Errico, F. [Universita degli Studi di Pisa, Department of Civil and Industrial Engineering, Lungamo Pacinotti 43, 56126 Pisa (Italy); Borroni, M.; Carrara, M. [Fondazione IRCCS Istituto Nazionale Tumori, Medical Physics Unit, Via Venezian 1, 20133 Milano (Italy); Burian, J.; Klupak, V.; Viererbl, L.; Marek, M. [Research Centre Rez, Department of Neutron Physics, 250-68 Husinec-Rez (Czech Republic)

2014-08-15

Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)

Dosimetry methods in boron neutron capture therapy

International Nuclear Information System (INIS)

Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D.; Agosteo, S.; Barcaglioni, L.; Campi, F.; Garlati, L.; De Errico, F.; Borroni, M.; Carrara, M.; Burian, J.; Klupak, V.; Viererbl, L.; Marek, M.

2014-08-01

Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)

Level gauge using neutron radiation

International Nuclear Information System (INIS)

Mathew, P.J.

1985-01-01

Apparatus for determining the level of a solid or liquid material in a container comprises: a vertical guide within or alongside the container; a sensor positioned within the guide; means for moving the sensor along the guide; and means for monitoring the position of the sensor. The sensor comprises a source of fast neutrons, a detector for thermal neutrons, and a body of a neutron moderating material in close proximity to the detector. Thermal neutrons produced by fast neutron irradiation of the solid or liquid material, or thermal neutrons produced by irradiation of the neutron-moderating material by fast or epithermal neutrons reflected by the solid or liquid material, are detected when the sensor is positioned at or below the level of the material in the container

Design of collimator in the radial piercing beam port of Kartini reactor for boron neutron capture therapy

International Nuclear Information System (INIS)

M Ilma Muslih A; Andang Widiharto; Yohannes Sardjono

2014-01-01

Studies were carried out to design a collimator which results in epithermal neutron beam for in vivo experiment of Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni (95%) as collimator wall, 15 cm thick of Al as moderator, 1 cm thick of Pb as γ-ray shielding, 1.5 cm thick of Boral as additional material, with 2 cm aperture diameter, epithermal neutron beam with maximum flux of 5.03 x 10 8 n.cm -2 .s -1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 2.17 x 10 -13 Gy.cm 2 .n -1 and 1.16 x 10 -13 Gy.cm 2 .n -l , minimum thermal neutron per epithermal neutron ratio of 0.12, and maximum directionality of 0.835 . It did not fully pass the IAEA's criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 10 9 n.cm -2 .s -l . Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 10 8 n.cm -2 .s -1 and fast neutron flux close to 2 x 10 -13 Gy.cm 2 .n -1 it is still feasible for BNCT in vivo experiment. (author)

The orientation effect in the activities of neutronic probes

International Nuclear Information System (INIS)

Vigon, M. A.

1956-01-01

The formulae relating activity and position of a neutron irradiated Indium foil, have been verified experimentally. Measurements with both thin and thick foils for epithermal neutrons and with thick foils for thermal neutrons have been carried out. The experimental results agree qualitatively with the theoretical predictions. (Author)

Neutron dosimetry in boron neutron capture therapy

International Nuclear Information System (INIS)

Fairchild, R.G.; Miola, U.J.; Ettinger, K.V.

1981-01-01

The recent development of various borated compounds and the utilization of one of these (Na 2 B 12 H 11 SH) to treat brain tumors in clinical studies in Japan has renewed interest in neutron capture therapy. In these procedures thermal neutrons interact with 10 B in boron containing cells through the 10 B(n,α) 7 Li reaction producing charged particles with a maximum range of approx. 10μm in tissue. Borated analogs of chlorpromazine, porphyrin, thiouracil and deoxyuridine promise improved tumor uptake and blood clearance. The therapy beam from the Medical Research Reactor in Brookhaven contains neutrons from a modified and filtered fission spectrum and dosimetric consequences of the use of the above mentioned compounds in conjunction with thermal and epithermal fluxes are discussed in the paper. One of the important problems of radiation dosimetry in capture therapy is determination of the flux profile and, hence, the dose profile in the brain. This has been achieved by constructing a brain phantom made of TE plastic. The lyoluminescence technique provides a convenient way of monitoring the neutron flux distributions; the detectors for this purpose utilize 6 Li and 10 B compounds. Such compounds have been synthesized specially for the purpose of dosimetry of thermal and epithermal beams. In addition, standard lyoluminescent phosphors, like glutamine, could be used to determine the collisional component of the dose as well as the contribution of the 14 N(n,p) 14 C reaction. Measurements of thermal flux were compared with calculations and with measurements done with activation foils

Nuclear polarization and neutrons

International Nuclear Information System (INIS)

Glaettli, H.

1985-01-01

Different possibilities for the use of polarized nuclei in thermal neutron scattering on condensed matter are reviewed. Highly polarized nuclei are the starting point for studying dipolar magnetic order. Systematic measurement of spin-dependent scattering lengths is possible on samples with polarized nuclei. Highly polarized hydrogen should help to unravel complicated structures in chemistry and biology. The use of polarized proton targets as an energy-independent neutron polarizer in the thermal and epithermal region should be considered afresh. (author)

Manual for the Epithermal Neutron Multiplicity Detector (ENMC) for Measurement of Impure MOX and Plutonium Samples

International Nuclear Information System (INIS)

Menlove, H. O.; Rael, C. D.; Kroncke, K. E.; DeAguero, K. J.

2004-01-01

We have designed a high-efficiency neutron detector for passive neutron coincidence and multiplicity counting of dirty scrap and bulk samples of plutonium. The counter will be used for the measurement of impure plutonium samples at the JNC MOX fabrication facility in Japan. The counter can also be used to create working standards from bulk process MOX. The detector uses advanced design "3He tubes to increase the efficiency and to shorten the neutron die-away time. The efficiency is 64% and the die-away time is 19.1 ?s. The Epithermal Neutron Multiplicity Counter (ENMC) is designed for high-precision measurements of bulk plutonium samples with diameters of less than 200 mm. The average neutron energy from the sample can be measured using the ratio of the inner ring of He-3 tubes to the outer ring. This report describes the hardware, performance, and calibration for the ENMC.

Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell.

Determination of Neutron Flux Parameter f and α and k0 Factor in Irradiation Facility of RSG GA Siwabessy reactor

International Nuclear Information System (INIS)

Amir Hamzah

2004-01-01

Determination of neutron flux thermal to epithermal ratio f and parameter α and k 0 factor has been done in irradiation facility of RSG G.A. Siwabessy reactor. Those parameters are needed to determine the concentration of an element in a sample using k 0 NAA method. Parameters f was measured using foil activation method and α parameter was obtained from power function fitting at epithermal neutron spectrum. Based on the fitting method the a parameter was determined of 0.0267,0.0255 and -0.0346 at system rabbit, IP2 and CIP irradiation position. The k 0 factor is depended on absolute gamma fraction. The neutron flux thermal to epithermal ratio f at all rabbit system is closed to 40. (author)

Neutron flux of 100kW in the irradiation terminals of the IPR-R1 Triga Reactor

International Nuclear Information System (INIS)

Zangirolami, Dante Marco

2009-01-01

In this work, it was carried out a study of the neutron flux in the IPR-R1 TRIGA reactor irradiation facilities: rotary specimen rack (RSR), pneumatic transfer tube two (PTT2) and the central thimble (CT). The objective was to obtain the neutron flux profile on the RSR, which has forty irradiation positions, and also values for the thermal and epithermal neutron fluxes of some RSR positions and also of the PTT2 and of the CT facility. It was applied the neutron activation analysis of a reference material, Al-Au (0.1%) alloy. Irradiations were performed on 16 different dates. It was concluded that for the RSR, the average value of thermal and epithermal neutron fluxes depends on the vertical position of the reactor control rods. Neutron flux variations along the RSR form a characteristic profile, whose values depend on the location of the irradiation position in the reactor core and on the control rods vertical position. In the RSR, the obtained values of thermal and epithermal neutron flux were (8.1 +- 0.3) x 10 11 n.cm -2 .s -1 , and (3.4 +- 0.2)x10 10 n.cm -2 .s -1 , respectively. For the PTT2 and the CT, the values for the epithermal neutron flux were respectively (3.3 +- 0.2) x 10 9 n.cm -2 .s -1 and (2.6 +- 0.1) x 10 11 n.cm -2 .s -1 . For these facilities, the thermal neutron flux was estimated, and the obtained values were (2.4 +- 0.2) x 10 11 n.cm -2 .s -1 and (2.8 +- 0.1)x10 12 n.cm -2 .s -1 for the PTT2 and the CT, respectively. (author)

The neutron porosity tool

International Nuclear Information System (INIS)

Oelgaard, P.L.

1988-01-01

The report contains a review of available information on neutron porosity tools with the emphasis on dual thermal-neutron-detector porosity tools and epithermal-neutron-detector porosity tools. The general principle of such tools is discussed and theoretical models are very briefly reviewed. Available data on tool designs are summarized with special regard to the source-detector distance. Tool operational data, porosity determination and correction of measurements are briefly discussed. (author) 15 refs

Epithermal Neutron Activation Analysis of the Asian Herbal Plants

International Nuclear Information System (INIS)

Baljinnyam, N.; Frontasyeva, M. V.; Ostrovnaya, T. M.; Pavlov, S. S.; Jugder, B.; Norov, N.

2011-01-01

Asian medicinal herbs Chrysanthemum (Spiraea aquilegifolia Pall.) and Red Sandalwood (Pterocarpus Santalinus) are widely used in folk and Ayurvedic medicine for healing and preventing some diseases. The modern medical science has proved that the Chrysanthemum (Spiraea aquilegifolia Pall.) possesses the following functions: reducing blood press, dispelling cancer cell, coronary artery's expanding and bacteriostating and Red Sandalwood (Pterocarpus Santalinus) is recommended against headache, toothache, skin diseases, vomiting and sometimes it is taken for treatment of diabetes. Species of Chrysanthemums were collected in the north-eastern and central Mongolia, and the Red Sandalwood powder was imported from India. Samples of Chrysanthemums (branches, flowers and leaves)(0.5 g) and red sandalwood powder (0.5 g) were subjected to the multi-element instrumental neutron activation analysis using epithermal neutrons (ENAA) at the IBR-2 reactor, Frank Laboratory of Neutron Physics (FLNP) JINR, Dubna. A total of 41 elements (Na, Mg, Al, Cl, K, Ca, Sc, V, Cr, Mn, Fe, Co, Ni, Zn, As, Se, Br, Rb, Sr, Zr, Mo, Cd, Cs, Ba, La, Hf, Ta, W, Sb, Au, Hg, Ce, Nd, Sm, Eu, Tb, Dy, Yb, Th, U, Lu) were determined. For the first time such a large group of elements was determined in the herbal plants used in Mongolia. The quality control of the analytical results was provided by using certified reference material Bowen Cabbage. The results obtained are compared to the ''Reference plant? data (B. Markert, 1992) and interpreted in terms of excess of such elements as Se, Cr, Ca, Fe, Ni, Mo, and rare earth elements.

Epithermal neutron activation analysis for studying the environment

International Nuclear Information System (INIS)

Frontas'eva, M.V.; Steinnes, E.

1997-01-01

Epithermal neutron activation analysis (ENAA) has certain advantages over the conventional instrumental analysis (INAA) in terms of improvement in precision and lowering of detection limits, reduction of high matrix activity and fission interferences if any. The current status and the applications of ENAA to environmental samples are reviewed. Experience in the use of ENAA in the monitoring of atmospheric depositions by means of moss-biomonitors at pulsed fast reactor IBR-2 in Dubna is summarized. INAA has shown to be useful for a number of sample types of interest in environmental studies, and should find more extensive use in this area. Analysis of airborne particulate matter is a case where ENAA should be particularly useful. A similar case where ENAA has shown strong performance is in the analysis of mosses used as biomonitors of atmospheric deposition, where 45 elements were determined. In this and other cases, however, induction-coupled plasma mass spectrometry is a very strong competitor, offering data for even more elements. A comparison of ENAA and ICP-MS for moss analysis is presented, and cases where ENAA is unique are discussed

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Development of new instrumentation for epithermal neutron scattering at very low angles

International Nuclear Information System (INIS)

Tardocchi, M.; Pietropaolo, A.; Andreani, C.; Gorini, G.; Perelli-Cippo, E.; Rhodes, N.J.; Schooneveld, E.M.; Senesi, R.

2004-01-01

New perspectives for epithermal neutron spectroscopy are opened up by the recent developments of Resonance Detectors (RD) for inverse geometry time-of-flight spectrometers at pulsed neutron sources. The RD is based on the combination of an analyser foil used as neutron-to-gamma converter and a suitable photon detector. Here, we report on the state of the RD which is based on a YAP scintillator viewing a natural uranium analyser foil. The response of the YAP detector to the radiative capture γ emission from the uranium analyser foil has been characterized with a bi-parametric measurement of a reference Pb sample, which allowed simultaneous measurements of both neutron time-of-flight and γ pulse height (energy) spectra. The analysis of the γ pulse height and neutron time-of-flight spectra permitted to identify the signal and background components. These measurements showed that a significant improvement in the signal-to-background ratio can be achieved by setting a lower level discrimination on the photon energy at about 600keV. The first application of RD is the Very Low Angle Detector Bank (VLAD) which is planned to be installed in the next three years as an upgrade of the VESUVIO spectrometer, at the ISIS pulsed neutron source. VLAD will extend the (q,ω) kinematical to low wave vector transfers (q10A-1) coupled to high-energy transfers (-bar ω>1eV), which is still unexplored by neutron scattering experiments. The first measurements obtained on an ice sample with a VLAD prototype consisting of four RD units are presented here

Development of new instrumentation for epithermal neutron scattering at very low angles

Energy Technology Data Exchange (ETDEWEB)

Tardocchi, M. [INFM, UdR Milano-Bicocca and Dipartimento di Fisica ' G.Occhialini' , Universita degli Studi di Milano-Bicocca, Pizza Della Scienza 3, Milan 20126 (Italy)]. E-mail: marco.tardocchi@mib.infn.it; Pietropaolo, A. [Dipartimento di Fisica, Universita degli Studi di Roma Tor Vergata and INFM, UdR Roma Tor Vergata, Rome (Italy); Andreani, C. [Dipartimento di Fisica, Universita degli Studi di Roma Tor Vergata and INFM, UdR Roma Tor Vergata, Rome (Italy); Gorini, G. [INFM, UdR Milano-Bicocca and Dipartimento di Fisica ' G.Occhialini' , Universita degli Studi di Milano-Bicocca, Pizza Della Scienza 3, Milan 20126 (Italy); Perelli-Cippo, E. [INFM, UdR Milano-Bicocca and Dipartimento di Fisica ' G.Occhialini' , Universita degli Studi di Milano-Bicocca, Pizza Della Scienza 3, Milan 20126 (Italy); Rhodes, N.J. [Isis Facility, Rutherford Appleton Laboratory, Chilton, Didcot (United Kingdom); Schooneveld, E.M. [Isis Facility, Rutherford Appleton Laboratory, Chilton, Didcot (United Kingdom); Senesi, R. [Dipartimento di Fisica, Universita degli Studi di Roma Tor Vergata and INFM, UdR Roma Tor Vergata, Rome (Italy)

2004-12-11

New perspectives for epithermal neutron spectroscopy are opened up by the recent developments of Resonance Detectors (RD) for inverse geometry time-of-flight spectrometers at pulsed neutron sources. The RD is based on the combination of an analyser foil used as neutron-to-gamma converter and a suitable photon detector. Here, we report on the state of the RD which is based on a YAP scintillator viewing a natural uranium analyser foil. The response of the YAP detector to the radiative capture {gamma} emission from the uranium analyser foil has been characterized with a bi-parametric measurement of a reference Pb sample, which allowed simultaneous measurements of both neutron time-of-flight and {gamma} pulse height (energy) spectra. The analysis of the {gamma} pulse height and neutron time-of-flight spectra permitted to identify the signal and background components. These measurements showed that a significant improvement in the signal-to-background ratio can be achieved by setting a lower level discrimination on the photon energy at about 600keV. The first application of RD is the Very Low Angle Detector Bank (VLAD) which is planned to be installed in the next three years as an upgrade of the VESUVIO spectrometer, at the ISIS pulsed neutron source. VLAD will extend the (q,{omega}) kinematical to low wave vector transfers (q10A-1) coupled to high-energy transfers (-bar {omega}>1eV), which is still unexplored by neutron scattering experiments. The first measurements obtained on an ice sample with a VLAD prototype consisting of four RD units are presented here.

Boron Neutron Capture Therapy activity of diffused tumors at TRIGA Mark II in Pavia

Energy Technology Data Exchange (ETDEWEB)

Bortolussi, S.; Stella, S.; De Bari, A.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy)]|[National Institute of Nuclear Physics (INFN), Pavia (Italy); Bruschi, P.; Bakeine, J.G. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy); Clerici, A.; Ferrari, C.; Zonta, C.; Zonta, A. [Department of Surgery, University of Pavia, Pavia (Italy); Nano, R. [Department of Animal Biology, University of Pavia, Pavia (Italy)

2008-10-29

The Boron neutron Capture Therapy research in Pavia has a long tradition: it begun more than 20 years ago at the TRIGA Mark II reactor of the University. A technique for the treatment of the hepatic metastases was developed, consisting in explanting the liver treated with {sup 10}B, irradiating it in the thermal column of the reactor, and re-implanting the organ in the patient. In the last years, the possibility of applying BNCT to the lung tumours using epithermal collimated neutron beams and without explanting the organ, is being explored. The principal obtained results of the BNCT research will be presented, with particular emphasis on the following aspects: a) the project of a new thermal column configuration to make the thermal neutron flux more uniform inside the explanted liver, b) the Monte Carlo study by means of the MCNP code of the thermal neutron flux distribution inside a patient's thorax irradiated with epithermal neutrons, and c) the measurement of the boron concentration in tissues by (n,{alpha}) spectroscopy and neutron autoradiography. (authors)

Boron Neutron Capture Therapy activity of diffused tumors at TRIGA Mark II in Pavia

International Nuclear Information System (INIS)

Bortolussi, S.; Stella, S.; De Bari, A.; Altieri, S.; Bruschi, P.; Bakeine, J.G.; Clerici, A.; Ferrari, C.; Zonta, C.; Zonta, A.; Nano, R.

2008-01-01

The Boron neutron Capture Therapy research in Pavia has a long tradition: it begun more than 20 years ago at the TRIGA Mark II reactor of the University. A technique for the treatment of the hepatic metastases was developed, consisting in explanting the liver treated with 10 B, irradiating it in the thermal column of the reactor, and re-implanting the organ in the patient. In the last years, the possibility of applying BNCT to the lung tumours using epithermal collimated neutron beams and without explanting the organ, is being explored. The principal obtained results of the BNCT research will be presented, with particular emphasis on the following aspects: a) the project of a new thermal column configuration to make the thermal neutron flux more uniform inside the explanted liver, b) the Monte Carlo study by means of the MCNP code of the thermal neutron flux distribution inside a patient's thorax irradiated with epithermal neutrons, and c) the measurement of the boron concentration in tissues by (n,α) spectroscopy and neutron autoradiography. (authors)

Fundamental of neutron radiography and the present of neutron radiography in Japan

International Nuclear Information System (INIS)

Sekita, Junichiro

1988-01-01

Neutron radiography refers to the application of transmitted neutrons to analysis. In general, thermal neutron is used for neutron radiography. Thermal neutron is easily absorbed by light atoms, including hydrogen, boron and lithium, while it is not easily absorbed by such heavy atoms as tungsten, lead and uranium, permitting detection of impurities in heavy metals. Other neutrons than thermal neutron can also be applied. Cold neutron is produced from fast neutron using a moderator to reduce its energy down to below that of thermal neutron. Cold neutron is usefull for analysis of thick material. Epithermal neutron can induce resonance characteristic of each substance. With a relatively small reaction area, fast neutron permits observation of thick samples. Being electrically neutral, neutrons are difficult to detect by direct means. Thus a substance that releases charged particles is put in the path of neutrons for indirect measurement. X-ray film combined with converter screen for conversion of neutrons to charge particles is placed behind the sample. Photographing is carried out by a procedure similar to X-ray photography. Major institues and laboratories in Japan provided with neutron radiography facilities are listed. (Nogami, K.)

Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT

International Nuclear Information System (INIS)

Evans, J.F.; Blue, T.E.

1996-01-01

Protecting the facility personnel and the general public from radiation exposure is a primary safety concern of an accelerator-based epithermal neutron irradiation facility. This work makes an attempt at answering the questions open-quotes How much?close quotes and open-quotes What kind?close quotes of shielding will meet the occupational limits of such a facility. Shielding effectiveness is compared for ordinary and barytes concretes in combination with and without borated polyethylene. A calculational model was developed of a treatment room, patient open-quotes scatterer,close quotes and the epithermal neutron beam. The Monte Carlo code, MCNP, was used to compute the total effective dose equivalent rates at specific points of interest outside of the treatment room. A conservative occupational effective dose rate limit of 0.01 mSv h -1 was the guideline for this study. Conservative Monte Carlo calculations show that constructing the treatment room walls with 1.5 m of ordinary concrete, 1.2 m of barytes concrete, 1.0 m of ordinary concrete preceded by 10 cm of 5% boron-polyethylene, or 0.8 m of barytes concrete preceded by 10 cm of 5% boron-polyethylene will adequately protect facility personnel. 20 refs., 8 figs., 2 tabs

Generating energy dependent neutron flux maps for effective ...

African Journals Online (AJOL)

For activation analysis and irradiation scheme of miniature neutron source reactor, designers or engineers usually require information on thermal neutron flux levels and other energy group flux levels (such as fast, resonance and epithermal). A methodology for readily generating such flux maps and flux profiles for any ...

Characterization of the TRIGA Mark III reactor for k0-neutron activation analysis

International Nuclear Information System (INIS)

Diaz R, O.; Herrera P, E.; Lopez R, M.C.

1997-01-01

The non-ideality of the epithermal neutron flux distribution in a a reactor site parameter (α), the thermal-to-epithermal neutron ratio (f), the irradiation channel neutron temperature (T n ) and the k 0 -factors for more than 20 isotopes were determined in the 3 typical irradiation positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using different experimental methods with conventional and non-conventional monitors. This characterization is used in the k 0 -method of NAA, recently introduced at the Institute. (author). 21 refs., 3 figs., 5 tabs

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

Energy Technology Data Exchange (ETDEWEB)

Van Do, Nguyen; Duc Khue, Pham; Tien Thanh, Kim [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Thi Hien, Nguyen [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Guinyun, E-mail: gnkim@knu.ac.kr [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Kwangsoo [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Shin, Sung-Gyun; Cho, Moo-Hyun [Department of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang 790-784 (Korea, Republic of); Lee, Manwoo [Research Center, Dongnam Institute of Radiological and Medical Science, Busan 619-953 (Korea, Republic of)

2015-11-01

The thermal neutron cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been measured relative to that of the {sup 197}Au(n,γ){sup 198}Au reaction by means of the activation method. High-purity natural scandium and gold foils without and with a cadmium cover of 0.5 mm thickness were irradiated with moderated pulsed neutrons produced from the Pohang Neutron Facility (PNF). The induced activities in the activated foils were measured with a high purity germanium (HPGe) detector. In order to improve the accuracy of the experimental results the counting losses caused by the thermal (G{sub th}) and resonance (G{sub epi}) neutron self-shielding, the γ-ray attenuation (F{sub g}) and the true γ-ray coincidence summing effects were made. In addition, the effect of non-ideal epithermal spectrum was also taken into account by determining the neutron spectrum shape factor (α). The thermal neutron cross-section and resonance integral of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been determined relative to the reference values of the {sup 197}Au(n,γ){sup 198}Au reaction, with σ{sub o,Au} = 98.65 ± 0.09 barn and I{sub o,Au} = 1550 ± 28 barn. The present thermal neutron cross section has been determined to be σ{sub o,Sc} = 27.5 ± 0.8 barn. According to the definition of cadmium cut-off energy at 0.55 eV, the present resonance integral cross section has been determined to be I{sub o,Sc} = 12.4 ± 0.7 barn. The present results are compared with literature values and discussed.

Volumetric Heat Generation and Consequence Raise in Temperature Due to Absorption of Neutrons from Thermal up to 14.9 MeV Energies

CERN Document Server

Massoud, E

2003-01-01

In this work, the heat generation rate and the consequence rise in temperature due to absorption of all neutrons from thermal energies (E<0.025) up to 14.9 MeV in water, paraffin wax, ordinary concrete and heavy concrete and heavy concrete as some selected hydrogenous materials are investigated. The neutron flux distributions are calculated by both ANISN-code and three group method in which the fast neutrons are expressed by the removal cross section concept while the other two groups (epithermal and thermal) are treated by the diffusion equation. The heat generation can be calculated from the neutron macroscopic absorption of each material or mixture multiplied by the corresponding neutron fluxes. The rise in temperature is then calculated by using both of the heat generation and the thermal conductivity of the selected materials. Some results are compared with the available experimental and theoretical data and a good agreement is achieved.

Epithermal neutron activation analysis in applied microbiology

International Nuclear Information System (INIS)

Marina Frontasyeva

2012-01-01

Some results from applying epithermal neutron activation analysis at FLNP JINR, Dubna, Russia, in medical biotechnology, environmental biotechnology and industrial biotechnology are reviewed. In the biomedical experiments biomass from the blue-green alga Spirulina platensis (S. platensis) has been used as a matrix for the development of pharmaceutical substances containing such essential trace elements as selenium, chromium and iodine. The feasibility of target-oriented introduction of these elements into S. platensis biocomplexes retaining its protein composition and natural beneficial properties was shown. The absorption of mercury on growth dynamics of S. platensis and other bacterial strains was observed. Detoxification of Cr and Hg by Arthrobacter globiformis 151B was demonstrated. Microbial synthesis of technologically important silver nanoparticles by the novel actinomycete strain Streptomyces glaucus 71 MD and blue-green alga S. platensis were characterized by a combined use of transmission electron microscopy, scanning electron microscopy and energy-dispersive analysis of X-rays. It was established that the tested actinomycete S. glaucus 71 MD produces silver nanoparticles extracellularly when acted upon by the silver nitrate solution, which offers a great advantage over an intracellular process of synthesis from the point of view of applications. The synthesis of silver nanoparticles by S. platensis proceeded differently under the short-term and long-term silver action. (author)

Microdosimetry for Boron Neutron Capture Therapy

International Nuclear Information System (INIS)

Maughan, R.L.; Kota, C.

2000-01-01

The specific aims of the research proposal were as follows: (1) To design and construct small volume tissue equivalent proportional counters for the dosimetry and microdosimetry of high intensity thermal and epithermal neutron beams used in BNCT, and of modified fast neutron beams designed for boron neutron capture enhanced fast neutron therapy (BNCEFNT). (2) To develop analytical methods for estimating the biological effectiveness of the absorbed dose in BNCT and BNCEFNT based on the measured microdosimetric spectra. (3) To develop an analytical framework for comparing the biological effectiveness of different epithermal neutron beams used in BNCT and BNCEFNT, based on correlated sets of measured microdosimetric spectra and radiobiological data. Specific aims (1) and (2) were achieved in their entirety and are comprehensively documented in Jay Burmeister's Ph.D. dissertation entitled ''Specification of physical and biologically effective absorbed dose in radiation therapies utilizing the boron neutron capture reaction'' (Wayne State University, 1999). Specific aim (3) proved difficult to accomplish because of a lack of sufficient radiobiological data

Absolute instrumental neutron activation analysis at Lawrence Livermore Laboratory

International Nuclear Information System (INIS)

Heft, R.E.

1977-01-01

The Environmental Science Division at Lawrence Livermore Laboratory has in use a system of absolute Instrumental Neutron Activation Analysis (INAA). Basically, absolute INAA is dependent upon the absolute measurement of the disintegration rates of the nuclides produced by neutron capture. From such disintegration rate data, the amount of the target element present in the irradiated sample is calculated by dividing the observed disintegration rate for each nuclide by the expected value for the disintegration rate per microgram of the target element that produced the nuclide. In absolute INAA, the expected value for disintegration rate per microgram is calculated from nuclear parameters and from measured values of both thermal and epithermal neutron fluxes which were present during irradiation. Absolute INAA does not depend on the concurrent irradiation of elemental standards but does depend on the values for thermal and epithermal neutron capture cross-sections for the target nuclides. A description of the analytical method is presented

Method and apparatus for formation logging using position sensitive neutron detectors

International Nuclear Information System (INIS)

Gadken, L.L.

1986-01-01

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

Application of reactors for testing neutron-induced upsets in commercial SRAMs

International Nuclear Information System (INIS)

Griffin, P.J.; Luera, T.F.; Sexton, F.W.; Cooper, P.J.; Karr, S.G.; Hash, G.L.; Fuller, E.

1997-01-01

Reactor neutron environments can be used to test/screen the sensitivity of unhardened commercial SRAMs to low-LET neutron-induced upset. Tests indicate both thermal/epithermal (< 1 keV) and fast neutrons can cause upsets in unhardened parts. Measured upset rates in reactor environments can be used to model the upset rate for arbitrary neutron spectra

Simulation study of the photon quality correction factors of ionization chambers for FiR 1 epithermal neutron beam

International Nuclear Information System (INIS)

Koivunoro, H.; Uusi-Simola, J.; Savolainen, S.; Kotiluoto, P.; Auterinen, I.; Kosunen, A.

2006-01-01

At FiR 1 BNCT facility in Finland, neutron-insensitive Mg(Ar) ionization chambers are used for photon dose measurements in an epithermal neutron beam. Previously, photon sensitivity factors for the chamber for the measurements in a water phantom in FiR 1 beam have been determined experimentally from measurements in 60 Co gamma and in a 6 MV clinical accelerator photon beams. However, the response of the ionization chamber in a water phantom depends on energy spectrum and angle of the photons and the secondary electrons created inside the phantom and may differ depending on type of the irradiation source (accelerator vs. an epithermal neutron beam). Also, the experimental sensitivity factor does not take into account the possible perturbations in the photon production in phantom caused by the ionization chamber materials. Therefore, it is necessary to determine the photon quality correction factors (k Qγ ) for the Mg(Ar) chamber at the FiR 1 beam through computer simulations. In this study, the k Qγ factors have been determined for Mg(Ar) chamber from Monte Carlo calculations of absorbed photon dose at two depths in a water phantom using MCNP code. The k qγ factors obtained with this method are compared to the sensitivity factors determined with measurements in an accelerator photon beam and to the k Qγ factors published previously. (author)

Improved Dose Targeting for a Clinical Epithermal Neutron Capture Beam Using Optional 6Li Filtration

International Nuclear Information System (INIS)

Binns, Peter J.; Riley, Kent J.; Ostrovsky, Yakov; Gao Wei; Albritton, J. Raymond; Kiger, W.S.; Harling, Otto K.

2007-01-01

Purpose: The aim of this study was to construct a 6 Li filter and to improve penetration of thermal neutrons produced by the fission converter-based epithermal neutron beam (FCB) for brain irradiation during boron neutron capture therapy (BNCT). Methods and Materials: Design of the 6 Li filter was evaluated using Monte Carlo simulations of the existing beam line and radiation transport through an ellipsoidal water phantom. Changes in beam performance were determined using three figures of merit: (1) advantage depth (AD), the depth at which the total biologically weighted dose to tumor equals the maximum weighted dose to normal tissue; (2) advantage ratio (AR), the ratio of the integral tumor dose to that of normal tissue averaged from the surface to the AD; and (3) advantage depth dose rate (ADDR), the therapeutic dose rate at the AD. Dosimetry performed with the new filter installed provided calibration data for treatment planning. Past treatment plans were recalculated to illustrate the clinical potential of the filter. Results: The 8-mm-thick Li filter is more effective for smaller field sizes, increasing the AD from 9.3 to 9.9 cm, leaving the AR unchanged at 5.7 but decreasing the ADDR from 114 to 55 cGy min -1 for the 12 cm diameter aperture. Using the filter increases the minimum deliverable dose to deep seated tumors by up to 9% for the same maximum dose to normal tissue. Conclusions: Optional 6 Li filtration provides an incremental improvement in clinical beam performance of the FCB that could help to establish a therapeutic window in the future treatment of deep-seated tumors

An epithermal neutron source for BNCT based on an ESQ-accelerator

International Nuclear Information System (INIS)

Ludewigt, B.A.; Chu, W.T.; Donahue, R.J.; Kwan, J.; Phillips, T.L.; Reginato, L.L.; Wells, R.P.

1997-07-01

An accelerator-based BNCT facility is under development at the Lawrence Berkeley National Laboratory. Neutrons will be produced via the 7 Li(p,n) reaction at proton energies of about 2.5 MeV with subsequent moderation and filtering for shaping epithermal neutron beams for BNCT. Moderator, filter, and shielding assemblies have been modeled using MCNP. Head-phantom dose distributions have been calculated using the treatment planning software BNCT RTPE. The simulation studies have shown that a proton beam current of ∼ 20 mA is required to deliver high quality brain treatments in about 40 minutes. The results also indicate that significantly higher doses can be delivered to deep-seated tumors in comparison to the Brookhaven Medical Research Reactor beam. An electrostatic quadrupole (ESQ) accelerator is ideally suited to provide the high beam currents desired. A novel power supply utilizing the air-coupled transformer concept is under development. It will enable the ESQ-accelerator to deliver proton beam currents exceeding 50 mA. A lithium target has been designed which consists of a thin layer of lithium on an aluminum backing. Closely spaced, narrow coolant passages cut into the aluminum allow the removal of a 50kW heat-load by convective water cooling. The system under development is suitable for hospital installation and has the potential for providing neutron beams superior to reactor sources

Neutron metrology in the HFR

International Nuclear Information System (INIS)

Voorbraak, W.P.; Freudenreich, W.E.; Stecher-Rasmussen, F.; Verhagen, H.W.

1991-10-01

Neutron fluence rate and gamma dose data are presented for the first series of experiments at the filtered HFR beam HB11 at full reactor power. Measurements were performed on two beagle dogs and one cylindrical phantom. The main results for thermal and epithermal fluence rates, physical neutron dose and gamma dose are presented in the tables 1 and 2. (author). 10 refs.; 9 figs.; 8 tabs

Characterisation of the epithermal neutron irradiation facility at the Portuguese research reactor using MCNP.

Science.gov (United States)

Beasley, D G; Fernandes, A C; Santos, J P; Ramos, A R; Marques, J G; King, A

2015-05-01

The radiation field at the epithermal beamline and irradiation chamber installed at the Portuguese Research Reactor (RPI) at the Campus Tecnológico e Nuclear of Instituto Superior Técnico was characterised in the context of Prompt Gamma Neutron Activation Analysis (PGNAA) applications. Radiographic films, activation foils and thermoluminescence dosimeters were used to measure the neutron fluence and photon dose rates in the irradiation chamber. A fixed-source MCNPX model of the beamline and chamber was developed and compared to measurements in the first step towards planning a new irradiation chamber. The high photon background from the reactor results in the saturation of the detector and the current facility configuration yields an intrinsic insensitivity to various elements of interest for PGNAA. These will be addressed in future developments. Copyright © 2015 Elsevier Ltd. All rights reserved.

Study of the RP-10 reactor neutron beam applied to the neutron radiography

International Nuclear Information System (INIS)

Zegarra, Manuel; Lopez, Alcides

2013-01-01

We have studied the RP-10 reactor radial neutron beam No. 3, which is used for neutron radiographies, by comparing radiograph's with and without the inner duct, and neutron flux determination with in flakes along the external duct, being the presence of photons creating signals at comparable levels of neutron effects, which reduce the quality of the analysis, values around 10 6 and 10 4 n/cm 2 s for thermal and epithermal flux were obtained respectively. It is recommended evaluate the design of the internal duct which presents strong photon emission. (authors).

Detection of explosives and illicit drugs using neutrons

Energy Technology Data Exchange (ETDEWEB)

Kiraly, B. E-mail: kiralyb@tigris.klte.hu; Sanami, T.; Doczi, R.; Csikai, J

2004-01-01

A procedure developed for the determination of the flux perturbation factor required for the thermal neutron activation analysis of bulky samples of unknown composition has been extended for epithermal neutrons using hydrogenous and graphite moderators. Measurements on the diffusion and backscattering of thermal neutrons in soil components were carried out for the development of novel nuclear methods in order to speed up the humanitarian demining process. Results obtained for the diffusion length were checked by MCNP-4C calculations. In addition, the effect of the weight and density of the explosives on the observation of the anomaly in the reflected thermal neutrons was examined by using different dummy landmines.

Study of spectral response of a neutron filter. Design of a method to adjust spectra

International Nuclear Information System (INIS)

Colomb-Dolci, F.

1999-02-01

The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (author)

A preliminary investigation on the epithermal flux depression effect due to cadmium box in a multiplying medium

International Nuclear Information System (INIS)

Ahmad, A.

1983-01-01

Cadmium boxes are widely used as filter in Reactor Neutron Activation Analysis (RNAA) for the irradiation of samples in epithermal neutrons. By virtue of being an absorber the cadmium boxes produce epithermal flux depression in the medium surrounding them. A preliminary study of this effect was carried out (author)

Neutron capture cross section standards for BNL 325, Fourth Edition

International Nuclear Information System (INIS)

Holden, N.E.

1981-01-01

This report evaluates the experimental data and recommends values for the thermal neutron cross sections and resonance integrals for the neutron capture reactions: 55 Mn(n,γ), 59 Co(n,γ) and 197 Au(n,γ). The failure of lithium and boron as standards due to the natural variation of the absorption cross sections of these elements is discussed. The Westcott convention, which describes the neutron spectrum as a thermal Maxwellian distribution with an epithermal component, is also discussed

SU-F-T-183: Design of a Beam Shaping Assembly of a Compact DD-Based Boron Neutron Capture Therapy System

Energy Technology Data Exchange (ETDEWEB)

Hsieh, M; Liu, Y; Nie, L [Purdue University, West Lafayette, Indiana (United States)

2016-06-15

Purpose: To design a beam shaping assembly (BSA) to shape the 2.45-MeV neutrons produced by a deuterium-deuterium (DD) neutron generator and to optimize the beam output for boron neutron capture therapy of brain tumors Methods: MCNP is used for this simulation study. The simulation model consists of a neutron surface source that resembles an actual DD source and is surrounded by a BSA. The neutron source emits 2.45-MeV neutrons isotropically. The BSA is composed of a moderator, reflector, collimator and filter. Various types of materials and geometries are tested for each component to optimize the neutron output. Neutron characteristics are measured with an 2×2×2-cm{sup 3} air-equivalent cylinder at the beam exit. The ideal BSA is determined by evaluating the in-air parameters, which include epithermal neutron per source neutron, fast neutron dose per epithermal neutron, and photon dose per epithermal neutron. The parameter values are compared to those recommended by the IAEA. Results: The ideal materials for reflector and thermal neutron filter were lead and cadmium, respectively. The thickness for reflector was 43 cm and for filter was 0.5 mm. At present, the best-performing moderator has 25 cm of AlF{sub 3} and 5 cm of MgF{sub 2}. This layout creates a neutron spectrum that has a peak at approximately 10 keV and produces 1.35E-4 epithermal neutrons per source neutron per cm{sup 2}. Additional neutron characteristics, fast neutrons per epithermal neutron and photon per epithermal neutron, are still under investigation. Conclusion: Working is ongoing to optimize the final layout of the BSA. The neutron spectrum at the beam exit window of the final configuration will have the maximum number of epithermal neutrons and limited photon and fast neutron contaminations within the recommended values by IAEA. Future studies will also include phantom experiments to validate the simulation results.

Thermal neutron moderating device

International Nuclear Information System (INIS)

Takigami, Hiroyuki.

1995-01-01

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

Neutron spectrum for neutron capture therapy in boron; Espectro de neutrones para terapia por captura de neutrones en boro

Energy Technology Data Exchange (ETDEWEB)

Medina C, D.; Soto B, T. G. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Programa de Doctorado en Ciencias Basicas, 98068 Zacatecas, Zac. (Mexico); Baltazar R, A. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Programa de Doctorado en Ingenieria y Tecnologia Aplicada, 98068 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

2016-10-15

Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with {sup 10}B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the {sup 10}B and produce a nucleus of {sup 7}Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10{sup 9} n/cm{sup 2}-sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

Measurement of thermal neutron cross-section and resonance integral for the 165Ho(n,γ) 166gHo reaction using electron linac-based neutron source

Science.gov (United States)

Nguyen, Van Do; Pham, Duc Khue; Kim, Tien Thanh; Kim, Guinyun; Lee, Manwoo; Kim, Kyung Sook; Kang, Heung-Sik; Cho, Moo-Hyun; Ko, In Soo; Namkung, Won

2011-01-01

The thermal neutron cross-section and the resonance integral of the 165Ho(n,γ) 166gHo reaction have been measured by the activation method using a 197Au(n,γ) 198Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation ( Fg), the thermal neutron self-shielding ( Gth), the resonance neutron self-shielding ( Gepi) effects, and the epithermal neutron spectrum shape factor ( α) were taken into account. The thermal neutron cross-section for the 165Ho(n,γ) 166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the 197Au(n,γ) 198Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the 165Ho(n,γ) 166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the 197Au(n,γ) 198Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.

Secondary neutron production from thick Pb target by light particle irradiation

CERN Document Server

Adloff, J C; Debeauvais, M; Fernández, F; Krivopustov, M; Kulakov, B A; Sosnin, A; Zamani, M

1999-01-01

Neutron multiplicities from spallation neutron sources were measured by Solid State Nuclear Track Detectors. Light particles as protons, deuterons and alphas in the GeV range were used on Pb targets. For neutron thermalization the targets were covered by 6 cm paraffin moderator. Neutron multiplicity distributions were studied inside and on the moderator surface. Comparison of SSNTDs results were made for thermal-epithermal neutrons with sup 1 sup 3 sup 9 La activation method as well as with Dubna DCM/CEM code. Discussion including previous sup 1 sup 2 C results are given.

Development of a new electronic neutron imaging system

Energy Technology Data Exchange (ETDEWEB)

Brenizer, J.S. [Department of Mechanical, Aerospace and Nuclear Engineering, Thornton Hall, University of Virginia, Charlottesville, VA 22903-2442 (United States); Berger, H. [Industrial Quality, Inc., Gaithersburg, MD (United States); Gibbs, K.M. [Industrial Quality, Inc., Gaithersburg, MD (United States); Mengers, P. [Paultek Systems, Inc., Nevada City, CA (United States); Stebbings, C.T. [Department of Mechanical, Aerospace and Nuclear Engineering, Thornton Hall, University of Virginia, Charlottesville, VA 22903-2442 (United States); Polansky, D. [Industrial Quality, Inc., Gaithersburg, MD (United States); Rogerson, D.J. [Naval Air Warfare Center, China Lake, CA (United States)

1999-11-03

An electronic neutron imaging camera system was developed for use with thermal, epithermal, and fast neutrons in applications that include nondestructive inspection of explosives, corrosion, turbine blades, electronics, low Z components, etc. The neutron images are expected to provide information to supplement that available from X-ray tests. The primary camera image area was a 30x30 cm field-of-view with a spatial resolution approaching 1.6 line pairs/mm (lp/mm). The camera had a remotely changeable second lens to limit the field-of-view to 7.6x7.6 cm for high spatial resolution (at least 4 lp/mm) thermal neutron imaging, but neutron and light scatter will limit resolution for fast neutrons to about 0.5 lp/mm. Remote focus capability enhanced camera set-up for optimum operation. The 75 dB dynamic range camera system included {sup 6}Li-based screens for imaging of thermal and epithermal neutrons and ZnS(Ag)-based screens for fast neutron imaging. The fast optics was input to a Super S-25 Gen II image intensifier, fiber optically coupled to a 1134 (h)x486 (v) frame transfer CCD camera. The camera system was designed to be compatible with a Navy-sponsored accelerator neutron source. The planned neutron source is an RF quadrupole accelerator that will provide a fast neutron flux of 10{sup 7} n/cm{sup 2}-s (at a source distance of 1 m) at an energy of about 2.2 MeV and a thermal neutron flux of 10{sup 6} n/cm{sup 2}-s at a source L/D ratio of 30. The electronic camera produced good quality real-time images at these neutron levels. On-chip integration could be used to improve image quality for low flux situations. The camera and accelerator combination provided a useful non-reactor neutron inspection system.

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

Science.gov (United States)

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

2017-02-01

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

Epithermal neutron flux in the experimental channels of the RA reactor; Fluks epitermalnih neutrona u eksperimentalnim kanalima reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Dobrosavljevic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Epithermal neutron flux was determined by measuring the cadmium ratio from activation analysis of gold and indium foils. Irradiation was done in experimental channels VK-5, VK-2, VK-7, VK-0, VK-HS next to the core, and next to the fuels elements. Activation of bare foils and foils covered by 0.8 and 1.0 mm thick cadmium foils was done simultaneously. Activity was measured by GM counter. Corrections were done for resonant neutrons self-absorption since the foils used were too thick concerning most important resonances of gold and indium. Final results were presented as spectrum factor r for their direct use in determining the effective neutron cross sections.

Single Crystal Filters for Neutron Spectrometry

International Nuclear Information System (INIS)

Habib, N.

2008-01-01

A study of neutron transmission properties trough a large single crystals specimens of Si, Ge, Pb, Bi and sapphire at 300 K and 80 K have been made for a wide range of neutron energies. The effectiveness of such filters is given by the ratio of the total cross-section of unwanted epithermal neutrons to that the desired thermal neutron beam and by the optimum choice of the crystal orientation, its mosaic spread, thickness and temperature.Our study indicates that sapphire is significantly more effective than the others for a wide range of neutron energies

A neutron beam polarizer for study of parity violation in neutron-nucleus interactions

International Nuclear Information System (INIS)

Penttilae, S.I.; Bowman, J.D.; Delheij, P.P.; Frankle, C.M.; Haase, D.G.; Postma, H.; Seestrom, S.J.; Yen, Y.

1995-01-01

A dynamically-polarized proton target operating at 5 Tesla and 1 K has been built to polarize an epithermal neutron beam for studies of parity violation in compound-nuclear resonances. Nearly 0.9 proton polarization was obtained in an electron-beam irradiated ammonia target. This was used to produce a neutron beam polarization of 0.7 at epithermal energies. The combination of the polarized proton target and the LANSCE spallation neutron source produces the most intense pulsed polarized epithermal neutron beam in the world. The neutron-beam polarizer is described and methods to determine neutron beam polarization are presented. copyright 1995 American Institute of Physics

Neutron spectrum for neutron capture therapy in boron

International Nuclear Information System (INIS)

Medina C, D.; Soto B, T. G.; Baltazar R, A.; Vega C, H. R.

2016-10-01

Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with 10 B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the 10 B and produce a nucleus of 7 Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10 9 n/cm 2 -sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

Application of semiconductor MOSFET and pin diode dosimeters to epithermal neutron beam dose distribution measurements in phantoms

International Nuclear Information System (INIS)

Carolan, M.G.; Wallace, S.A.; Allen, B.J.; Rosenfeld, A.B.; Mathur, J.N.

1996-01-01

For any clinical application of Boron Neutron Capture Therapy (BNCT) fast and accurate dose calculations will be required for treatment planning. Such calculations are also necessary for the planning and interpretation of results from pre-clinical and clinical trials where the speed of calculation is not so critical. A dose calculation system based on the MCNP Monte Carlo Neutron transport code has been developed by Wallace. This system takes image data from CT scans and constructs a voxel based geometrical model for input into MCNP. To validate the calculations, a number of phantoms were constructed and exposed in the HB11 epithermal neutron beam at the HFR of the CEC Joint Research Centre in Petten. The doses recorded by arrays of PIN diode neutron dosimeters and MOSFET gamma dosimeters in these phantoms were compared with the calculated results from the MCNP dose planning system. Initial results have been reported elsewhere. Poster 197. (author)

DAWN GRAND MAP CERES TPE NEUTRON COUNTS V1.0

Data.gov (United States)

National Aeronautics and Space Administration — A global map thermal+epithermal neutron counting rates binned on twenty-degree quasi-equal-area pixels is provided. The map was determined from a time series of the...

An optimum source neutron spectrum and holder shape for extra-corporal treatment of liver cancer by BNCT

International Nuclear Information System (INIS)

Nievaart, Sander; Moss, Ray; Sauerwein, Wolfgang; Malago, Massimo; Kloosterman, Jan Leen; Hagen, Tim van der; Dam, Hugo van

2006-01-01

In extra-corporal treatment of liver cancer by BNCT, it is desired to have an as homogeneous as possible thermal neutron field throughout the organ. Previous work has shown that when using an epithermal neutron beam, the shape of the holder in which the liver is placed is the critical factor. This study develops the notion further as to what is the optimum neutron spectrum to perform such treatments. In the design calculations, when using Monte Carlo techniques, it is shown that when the expected contributions of the source neutrons in every part of the liver is calculated, a linear optimization scheme such as the Simplex method results in a mix of thermal and epithermal source neutrons to get the highest homogeneity for the thermal neutron field. This optimisation method is demonstrated in 3 holder shapes: cuboid, cylindrical and spherical with each 3 volumes of 2, 4 and 6 litres. A 10 cm thick cuboid model, irradiated from both sides gives the highest homogeneity. The spherical (rotating) holder has the lowest homogeneity but the highest contribution of every source neutron to the thermal neutrons in the liver. This can be advantageous when using a relatively small sized neutron beam with a low strength. (author)

Study on the determination of uranium by activation analysis with epithermal neutrons

International Nuclear Information System (INIS)

Atalla, L.T.

1977-01-01

A method is described that is applied to the determination of uranium in different types of materials, either by an entirely instrumental method or with the chemical separation of uranium-239, when the presence of interferences does not allow the instrumental analysis. The advantages and disadvantages in the use of epithermal neutrons in the activation of samples for a more selective activation of uranium-238 also presented. The instrumental method is tested through standart materials, accepted internationally. The possibility of uranium extraction with di-etil-hexilphosphoric acid is also presented and the choice of the former technique is justified. The sensitibility of the method is discussed as well as precision and accuracy through results obtained in the analysis of the standards and the calibration curve of uranium [pt

Neutron flux measurements in PUSPATI Triga Reactor

International Nuclear Information System (INIS)

Gui Ah Auu; Mohamad Amin Sharifuldin Salleh; Mohamad Ali Sufi.

1983-01-01

Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm 2 sec and 1.10E+12n/cm 2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

Measurement of the epithermal neutron flux of the Argonauta reactor by the Sandwich method

International Nuclear Information System (INIS)

Nascimento, H.M.

1973-01-01

A common method of obtaining information about the neutron spectrum in the energy range of 1 eV to a few keV is by using resonance sandwich detectors. A sandwich detector is usually made up of three foils placed one on top of the other, each having the same thickness and being made of the same material which has a pronounced absorption resonance. To make an adequate evaluation, the sandwich method was compared with one using an isolated detector. The results obtained from approximate theoretical calculations were checked experimentally, using In, Au and Mn foils, in an isotropic 1/E flux in the Argonaut Reactor at I.E.N. As practical application of this method, the deviation from a 1/E spectrum of the epithermal neutron flux in the core and external graphite reflector of the Argonaut Reactor has been measured with the sandwich foils previously calibrated in a 1/E spectrum. (author)

Study of spectral response of a neutron filter. Design of a method to adjust spectra; Etude des moyens de conditionnement de la reponse spectrale d'un filtre a neutrons. Mise au point d'une methode d'ajustement rapide de spectre

Energy Technology Data Exchange (ETDEWEB)

Colomb-Dolci, F. [Universite Louis Pasteur, 67 - Strasbourg (France)

1999-02-01

The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (author)

Clinical considerations for neutron capture therapy of brain tumors

International Nuclear Information System (INIS)

Madoc-Jones, H.; Wazer, D.E.; Zamenhof, R.G.; Harling, O.K.; Bernard, J.A. Jr.

1990-01-01

The radiotherapeutic management of primary brain tumors and metastatic melanoma in brain has had disappointing clinical results for many years. Although neutron capture therapy was tried in the US in the 1950s and 1960s, the results were not as hoped. However, with the newly developed capability to measure boron concentrations in blood and tissue both quickly and accurately, and with the advent of epithermal neutron beams obviating the need for scalp and skull reflection, it should not be possible to mount such a clinical trial of NCT again and avoid serious complications. As a prerequisite, it will be important to demonstrate the differential uptake of boron compound in brain tumor as compared with normal brain and its blood supply. If this can be done, then a trial of boron neutron capture therapy for brain tumors should be feasible. Because boronated phenylalanine has been demonstrated to be preferentially taken up by melanoma cells through the biosynthetic pathway for melanin, there is special interest in a trial of boron neutron capture therapy for metastatic melanoma in brain. Again, the use of an epithermal beam would make this a practical possibility. However, because any epithermal (or thermal) beam must contain a certain contaminating level of gamma rays, and because even a pure neutron beam cases gamma rays to be generated when it interacts with tissue, they think that it is essential to deliver treatments with an epithermal beam for boron neutron capture therapy in fractions in order to minimize the late-effects of low-LET gamma rays in the normal tissue

Borehole logging system

International Nuclear Information System (INIS)

Allen, L.S.

1988-01-01

A radioactive borehole logging tool employs an epithermal neutron detector having a neutron counter surrounded by an inner thermal neutron filter and an outer thermal neutron filter. Located between the inner and outer filters is a neutron moderating material for extending the lifetime of epithermal neutrons to enhance the counting rate of such epithermal neutrons by the neutron counter

Effective dose evaluation for BNCT treatment in the epithermal neutron beam at THOR

Energy Technology Data Exchange (ETDEWEB)

Wang, J.N. [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)] [Division of Health Physics, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Huang, C.K. [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Tsai, W.C. [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Liu, Y.H. [Nuclear Science and Technol. Develop. Center, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Jiang, S.H., E-mail: shjiang@mx.nthu.edu.tw [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)] [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)

2011-12-15

This paper aims to evaluate the effective dose as well as equivalent doses of several organs of an adult hermaphrodite mathematical phantom according to the definition of ICRP Publication 60 for BNCT treatments of brain tumors in the epithermal neutron beam at THOR. The MCNP5 Monte Carlo code was used for the calculation of the average absorbed dose of each organ. The effective doses for a typical brain tumor treatment with a tumor treatment dose of 20 Gy-eq were evaluated to be 0.59 and 0.35 Sv for the LLAT and TOP irradiation geometries, respectively. In addition to the stochastic effect, it was found that it is also likely to produce deterministic effects, such as cataracts and depression of haematopoiesis.

The orientation effect in the activities of neutronic probes; Efecto de orientacion en las actividades de sondas neutronicas

Energy Technology Data Exchange (ETDEWEB)

Vigon, M A

1956-07-01

The formulae relating activity and position of a neutron irradiated Indium foil, have been verified experimentally. Measurements with both thin and thick foils for epithermal neutrons and with thick foils for thermal neutrons have been carried out. The experimental results agree qualitatively with the theoretical predictions. (Author)

Effect of head size on 10B dose distribution

International Nuclear Information System (INIS)

Gupta, N.; Blue, T.E.; Gahbauer, R.

1992-01-01

Boron neutron capture therapy (BNCT) for treatment of brain tumors is based on the utilization of large epithermal-neutron fields. Epithermal neutrons thermalize at depths of ∼2.5 cm inside the head and provide a maximum thermal fluence at deep-seated tumor sites with minimum damage to normal tissue. Brain tissue is a highly scattering medium for epithermal and thermal neutrons; therefore, a broad treatment field enables epithermal neutrons to enter the head over a large area. These neutrons slow down as they undergo scattering collisions and contribute to the thermal-neutron fluence at the tumor location. With the use of large neutron fields, the size of the head affects the thermal-neutron distribution and thereby the 10 B absorbed dose distribution inside the head. In this paper, the authors describe measurements using a boron trifluoride (BF 3 )-filled proportional counter to determine the effect of head size on 10 B absorbed dose distributions for a broad field accelerator epithermal-neutron source

Application of epithermal neutron activation in multielement analysis of silicate rocks employing both coaxial Ge(Li) and low energy photon detector systems

Science.gov (United States)

Baedecker, P.A.; Rowe, J.J.; Steinnes, E.

1977-01-01

The instrumental activation analysis of silicate rocks using epithermal neutrons has been studied using both high resolution coaxial Ge(Li) detectors and low energy photon detectors, and applied to the determination of 23 elements in eight new U.S.G.S. standard rocks. The analytical use X-ray peaks associated with electron capture or internal conversion processes has been evaluated. Of 28 elements which can be considered to be determinable by instrumental means, the epithermal activation approach is capable of giving improved sensitivity and precision in 16 cases, over the normal INAA procedure. In eleven cases the use of the low energy photon detector is thought to show advantages over convertional coaxial Ge(Li) spectroscopy. ?? 1977 Akade??miai Kiado??.

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)

A physical and engineering study on the irradiation techniques in neutron capture therapy aiming for wider application

International Nuclear Information System (INIS)

Sakurai, Y.; Ono, K.; Suzuki, M.; Katoh, I.; Miyatake, S.-I.; Yanagie, H.

2003-01-01

The solo-irradiation of thermal neutrons has been applied for brain cancer and malignant melanoma in the boron neutron capture therapy (BNCT) at the medical irradiation facility of Kyoto University Reactor (KUR), from the first clinical trial in 1974. In 1997, after the facility remodeling, the application of the mix-irradiation of thermal and epi-thermal neutrons was started, and the depth dose distribution for brain cancer has been improved in some degree. In 2001, the solo-irradiation of epi-thermal neutrons also started. It is specially mentioned that the application to oral cancers started at the same time. The BNCT clinical trial using epi-thermal neutron irradiation at KUR, amounts to twelve as of March 2003. The seven trials; more than a half of the total trials, are for oral cancers. From this fact, we think that the wider application to the other cancers is required for the future prosperity of BNCT. The cancers applied for BNCT in KUR at the present time, are brain cancer, melanoma and oral cancers, as mentioned above. The cancers, expected to be applied in near future, are liver cancer, pancreas cancer, lung cancer, tongue cancer, breast cancer, etc.. Any cancer is almost incurable by the other therapy including the other radiation therapy. In the wider application of BNCT to these cancers, the dose-distribution control suitable to each cancer and/or each part, is important. The introduction of multi-directional and/or multi-divisional irradiation is also needed. Here, a physical and engineering study using two-dimensional transport calculation and three-dimensional Monte-Carlo simulation for the irradiation techniques in BNCT aiming for wider application is reported

Remarks on some rock neutron parameters

International Nuclear Information System (INIS)

Czubek, J.A.

1984-01-01

A method to calculate the thermal neutron parameters of rocks is given in the paper. It is based on a proper energy averaging of cross-sections for all rock matrix and rock saturating liquid constituents. The diffusion lengths in different lithologies in function of the variable rock porosity have been calculated. An influence of the thermal neutron spectrum on the shape of the porosity calibration curves for the dual spacing neutron method is shown. Magmatic rocks as a possible source of geothermal energy are now becoming a target of neutron loggings for the porosity determination. Here the knowledge of the slowing-down lengths is of great importance in the problem of the estimation of the calibration curves. A semi-analytical approach to get this parameter is given in the paper. It was found, as far as concerns the slowing-down of fast neutrons, that all magmatic rocks behave as sandstone with, however, different content of bound water in the rock matrix and different rock matrix density. Some neutron methods are based on the detection of epithermal neutrons. For theoretical considerations it is important to know the physical meaning of the registered signal. From the discussion of experimental data reported in the literature it seems that it is the slowing-down density that is the physical quantity being measured. This conclusion has a very important practical implication - the porosity calibration curves depend upon the slowing-down length alone and are independent of the slowing-down cross-section for epithermal neutrons

Use of the Power Burst Facility for boron neutron capture therapy

International Nuclear Information System (INIS)

Crocker, J.G.; Griebenow, M.L.; Leatham, J.

1990-01-01

A program is under development at the Idaho National Engineering Laboratory (INEL) that involves using the Power Burst Facility (PBF) for research into boron neutron capture therapy (BNCT). BNCT utilizes the ionizing energy from boron-neutron capture to stop reproduction of or destroy cells in cancerous tissue in a two-step process. The first step is to selectively concentrate a boron isotope within the tumor cell, that when activated by neutron capture emits highly ionizing, short range particles. The second step involves activation of the isotope only in the vicinity of the tumor with a narrow neutron beam. The ( 10 B[n, 4 He] 7 Li) reaction with thermal neutrons produces fission products with track lengths approximately equal to a cell diameter. The INEL program includes the modification of the PBF by the addition of a filter and treatment area. The filter will down-scatter high energy neutrons into the epithermal range and remove thermal neutrons and excessively damaging gamma components. The intense source of epithermal neutrons from PBF is considered necessary to achieve optimum therapy for deep-seated tumors with minimum damage to surface tissue. THe neutron filter conceptualized for PBF utilizes aluminum and heavy water to down-scatter neutrons into the proper energy range. Bismuth will be used for gamma shielding and cadmium will remove the thermal neutron contaminant from the beam. The INEL program leads to human clinical trials at PBF which are intended to prove that brain tumors can be successfully treated through noninvasive techniques. Further research into BNCT at PBF for other cancer types is also anticipated

Thermal neutron dose calculations in a brain phantom from 7Li(p,n) reaction based BNCT setup

International Nuclear Information System (INIS)

Elshahat, B.A.; Naqvi, A.A.; Maalej, N.; Abdallah, Khalid

2006-01-01

Monte Carlo simulations were carried out to calculate neutron dose in a brain phantom from a 7 Li(p,n) reaction based setup utilizing a high density polyethylene moderator with graphite reflector. The dimensions of the moderator and the reflector were optimized through optimization of epithermal /(fast +thermal) neutron intensity ratio as a function of geometric parameters of the setup. Results of our calculation showed the capability of our setup to treat the tumor within 4 cm of the head surface. The calculated Peak Therapeutic Ratio for the setup was found to be 2.15. With further improvement in the moderator design and brain phantom irradiation arrangement, the setup capabilities can be improved to reach further deep-seated tumor. (author)

Analysis of neutron flux increase in the horizontal experimental channels of Ra reactor - masters thesis

International Nuclear Information System (INIS)

Strugar, P.

1964-12-01

Calculation and experimental results shown in this paper show that higher thermal neutron flux is obtained in the reactor core with central horizontal reflector at the same power level. The flux is increased when the moderation capability of the core is decreased. Apart from increase of the thermal component of the neutron flux in the experimental channels, the central reflector causes decrease of the epithermal neutron flux and gamma radiation intensity. This is very useful for studying (n, γ) reaction, neutron diffraction, etc. [sr

Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities.

Science.gov (United States)

Mansy, M S; Bashter, I I; El-Mesiry, M S; Habib, N; Adib, M

2015-03-01

Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5-133keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named "QMNB" was developed in the "MATLAB" programming language to perform the required calculations. Copyright © 2014 Elsevier Ltd. All rights reserved.

Design of incoming neutron-beam for detecting oil dirt

International Nuclear Information System (INIS)

Zhao Jingwu; Chen Xiaocheng; Alimujiang Naimaiti; Aierken Abuliemu

2012-01-01

For the technique of neutron back-scattering, the neutron counts are non-linear and have a tendency toward saturation because of the neutron self-shielding. As a result, the measurement accuracy is reduced and the measurement range is limited. Using a simply model and comparing with experimental data, it is shown that, in the measurement of the thickness of oil dirt, by adjusting the ratio of thermal to epithermal neutrons, the neutron self: shielding is weakened. As a result, the non-linearity can be reduced and the measurement accuracy and range can be improved. (authors)

Determination of Uranium and Thorium in Brazilian coals by epithermal neutron activation analysis

International Nuclear Information System (INIS)

Bernedo, L.F.B.

1981-08-01

An experimental technique for the determination of uranium and thorium in coal by epithermal neutron activation was developed and systemized. Seventeen different coal samples, six copper monitors for neutron flux corrections and three NBS standard coal samples were irradiated together in a cadmium cylinder. Uranium and thorium were determined by measuring the 239 N sub(p) and 233 P sub(a) activities respectively, being both produced in (n,γ) reactions and subsequent β - decay. The 239 N sub(p) was measured by counting the 106.4 KeV γ-ray in a LEPS detector and the 233 P sub(a) by counting the 311.8 KeV γ-ray, but in a Ge(Li) detector. A 4096 multichannel analizer and a PDP-11 computer complemented the basic measuring equipment. An average precision of 3% was obtained in the analysis of seventeen coal samples coming from different strata and heights of Charqueadas and Morungava mines in Rio Grande do Sul State. The sensitivity of the method is around 100 ppb. This technique will allow determinations of up to twenty elements, besides uranium and thorium, and it can be applied in routine analysis. (Author) [pt

Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

Science.gov (United States)

Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

2018-01-01

University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

Fuel assembly inspection by three-dimensional neutron radiography

International Nuclear Information System (INIS)

Lapinski, N.P.; Reimann, K.J.; Berger, H.

1979-01-01

Radiographic inspection of complex objects such as fuel subassemblies often presents problems because superimposition of images at different depths in the object complicates interpretation. One method for obtaining and displaying three-dimensional neutron radiographic images in multiple-film laminagraphy; a series of radiographs generated at different angular orientations are superimposed to provide focussed images of any object plane. In the present work multiple-film neutron laminagraphs were generated using direct and indirect exposure techniques, with neutrons in thermal, epithermal, and fast energy ranges

Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes

International Nuclear Information System (INIS)

Bringer, O.

2007-10-01

This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of 241 Am and 237 Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the 241 Am isomeric branching ratio were precisely measured at thermal energy point. (author)

Status report on the development of a prompt fission neutron uranium borehole logging technique

International Nuclear Information System (INIS)

Smith, G.W.

1977-05-01

The prompt fission neutron (PFN) method of direct uranium measurement was studied. The PFN uranium logging technique measures the enhanced epithermal neutron population created by the prompt thermal fission of 235 U to assay uranium mineralization around a borehole. This neutron population exists for several hundred microseconds after a pulsed neutron source produces a burst of high energy (14 MeV) neutrons. A feasibility study established the basic relationship between the uranium concentration and the enhanced epithermal neutron count, and defined the major measurement perturbing factors. Following the feasibility study, development of a PFN prototype field probe was undertaken. A laboratory type neutron generator, the Controlatron, was modified for use in the probe. Field evaluation of the prototype system began in January 1976. Comparisons of neutron logs and natural gamma logs taken during this evaluation period clearly define many disequilibrium conditions as verified by ore grade estimates from core samples. The feasibility of the PFN logging technique to detect uranium in-situ has now been demonstrated

Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

Energy Technology Data Exchange (ETDEWEB)

Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

2011-02-15

The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

Evaluation of mixed energy neutron doses using TLD NG-67 type

International Nuclear Information System (INIS)

Akhadi, Mukhlis; Thoyib Thamrin, M; Usmiyati Dewi, K.

2000-01-01

A research has been carried out to develop dose evaluation method of mixed neutron source with its neutron doses can be classified to two groups, I.e neutron doses with energy ≥ 0.5 eV and thermal neutron doses with energy less than 0.5 e V consist of epithermal and fast neutron, but in this research they were classified as fast neutron. Development of this dose evaluation method was carried out by sensitivity (S) intercomparison of TLD-600 to fast neutron, mixed energy neutron of nuclear rectors, and thermal neutron. From the experiment it was obtained that the value of Sfast : Sreactor : Sthermal = 0.005 : 0.010 : 1. Calibration factor (CF) of TLD is defined as 1/S. from the sensitivity data it can be obtained that the value of Cffast : Cfreactor : Cfthermal = 200 :100 : 1. The value of Cfreactor can be applied for mixed energy neutron doses evaluation of TLD-600. Key word : dosemeter, neutron dose, calibration factor, fast neutron, thermal neutron, nuclear reactor

Method and apparatus for determination of temperature, neutron absorption cross section and neutron moderating power

Science.gov (United States)

Vagelatos, Nicholas; Steinman, Donald K.; John, Joseph; Young, Jack C.

1981-01-01

A nuclear method and apparatus determines the temperature of a medium by injecting fast neutrons into the medium and detecting returning slow neutrons in three first energy ranges by producing three respective detection signals. The detection signals are combined to produce three derived indicia each systematically related to the population of slow neutrons returning from the medium in a respective one of three second energy ranges, specifically exclusively epithermal neutrons, exclusively substantially all thermal neutrons and exclusively a portion of the thermal neutron spectrum. The derived indicia are compared with calibration indicia similarly systematically related to the population of slow neutrons in the same three second energy ranges returning from similarly irradiated calibration media for which the relationships temperature, neutron absorption cross section and neutron moderating power to such calibration indicia are known. The comparison indicates the temperature at which the calibration indicia correspond to the derived indicia and consequently the temperature of the medium. The neutron absorption cross section and moderating power of the medium can be identified at the same time.

Characterization of a fast to thermal neutron spectrum converter on PROSPERO reactor

Energy Technology Data Exchange (ETDEWEB)

Jacquet, X.; Authier, N.; Casoli, P.; Combacon, S. [CEA, Valduc Center, 21120 Is sur Tille (France); Calzavarra, Y. [ILL, Institut Laue Langevin, 38000 Grenoble (France)

2009-07-01

The PROSPERO reactor is located at CEA Valduc Center in France. The reactor is composed of an internal core made of High Enriched Uranium metal alloy surrounded by a reflector of depleted uranium. The reactor is used as a fast neutron spectrum source and is operated in delayed critical state with a continuous and steady power for several hours, which can vary from 3 mW to 3 kW, which is the nominal power. The flux at nominal power varies from 5.10{sup +10} n.cm{sup -2}/s at the reflector surface to 10{sup +7} n.cm{sup -2}/s at 5 meters from reactor axis. It has been decided to build a neutron energy converter allowing the production of a neutron thermal spectrum. As the core produces fast neutrons spectrum, we built a hollow cubic box of 50 cm x 50 cm x 50 cm with 10-cm-thick polyethylene bricks and placed one meter away from central reactor axis to moderate as much as possible neutrons to lower energies (E<0.6 eV). Analysis of the moderated flux inside the converter was performed using different activation foils such as indium or gold. We have developed a model of the experiment in the Monte Carlo neutron transport code TRIPOLI-4. A non-analogous transport calculation scheme was necessary to reproduce properly the experimental activities. The results of the calculated activations are within 4% of the experimental measurements given with 10% uncertainty (2 sigma). We show that the converter realizes thermalization of 80 % of the PROSPERO reactor fast neutrons below the cadmium threshold of 0.6 eV. Epithermal neutrons represent 15% of the spectrum and only 5% are in the fast neutron range above 1 MeV. The total flux at the center of the converter is 1.4 10{sup +9} n.cm{sup -2}/s at 3000 W

Marine Gradients of Halogens in Moss Studied by Epithermal Neutron Activation Analysis

CERN Document Server

Frontasyeva, M V

2002-01-01

Epithermal neutron activation analysis is known to be a powerful technique for the simultaneous study of chlorine, bromine and iodine in environmental samples. In this paper it is shown to be useful to elucidate marine gradients of these elements. Examples are from a transect study in northern Norway where samples of the feather moss Hylocomium splendens were collected at distances 0-300 km from the coastline. All three elements decreased exponentially as a function of distance from the ocean in the moss samples, strongly indicating that atmospheric supply from the marine environment is the predominant source of these elements to the terrestrial ecosystem. These results are compared with similar data for surface soils along the same gradients. Comparison is also made with previous data for halogens in moss in Norway obtained by conventional NAA and covering similar transects in other geographical regions. The Cl/Br and Br/I ratios in moss showed a regular change distance from the ocean in all transects, and h...

Epithermal Neutron Activation Analysis (ENAA) of Cr(VI)-reducer Basalt-inhabiting Bacteria

CERN Document Server

Tsibakhashvili, N Ya; Kirkesali, E I; Aksenova, N G; Kalabegishvili, T L; Murusidze, I G; Mosulishvili, L M; Holman, H Y N

2005-01-01

Epithermal neutron activation analysis (ENAA) has been applied to studying elemental composition of Cr(VI)-reducer bacteria isolated from polluted basalts from the Republic of Georgia. Cr(VI)-reducing ability of the bacteria was examined by electron spin resonance (ESR) demonstrating that the bacteria differ in the rates of Cr(VI) reduction. A well-pronounced correlation between the ability of the bacteria to accumulate Cr(V) and their ability to reduce Cr(V) to Cr(III) observed in our experiments is discussed. Elemental analysis of these bacteria also revealed that basalt-inhabiting bacteria are distinguished by relative contents of essential elements such as K, Na, Mg, Fe, Mn, Zn, and Co. A high rate of Cr(III) formation correlates with a high concentration of Co in the bacterium. ENAA detected some similarity in the elemental composition of the bacteria. The relatively high contents of Fe detected in the bacteria (140-340 $\\mu $g/g of dry weight) indicate bacterial adaptation to the environmental condition...

Application of epithermal neutron activation analysis to investigate accumulation and adsorption of mercury by Spirulina platensis biomass

International Nuclear Information System (INIS)

Mosulishvili, L.M.; Belokobyl'skij, A.I.; Khizanishvili, A.I.; Frontas'eva, M.V.; Kirkesali, E.I.; Aksenova, N.G.

2004-01-01

Epithermal neutron activation analysis was used to study interaction of blue-green alga Spirulina platensis with toxic metal mercury. Various concentrations of Hg(II) were added to cell cultures in a nutrient medium. The dynamics of accumulation of Hg was investigated over several days in relation to Spirulina biomass growth. The process of Hg adsorption by Spirulina biomass was studied in short-time experiments. The isotherm of adsorption was carried out in Freindlich coordinates. Natural Spirulina biomass has potential to be used in the remediation of sewage waters at Hg concentrations ∼100 μg/1

Application of Epithermal Neutron Activation Analysis to Investigate Accumulation and Adsorption of Mercury by Spirulina platensis Biomass

CERN Document Server

Mosulishvili, L M; Khizanishvili, A I; Frontasyeva, M V; Kirkesali, E I; Aksenova, N G

2004-01-01

Epithermal neutron activation analysis was used to study interaction of blue-green alga Spirulina platensis with toxic metal mercury. Various concentrations of Hg(II) were added to cell cultures in a nutrient medium. The dynamics of accumulation of Hg was investigated over several days in relation to Spirulina biomass growth. The process of Hg adsorption by Spirulina biomass was studied in short-time experiments. The isotherm of adsorption was carried out in Freindlich coordinates. Natural Spirulina biomass has potential to be used in the remediation of sewage waters at Hg concentrations \\sim 100 {\\mu}g/l.

A 'hybrid' neutron area survey instrument for the determination of neutron dose quantities in the workplace

International Nuclear Information System (INIS)

Tanner, R.J.; Jenkins, R.; Lowe, T.; Silvie, J.; Joyce, M.J.; Winsby, A.; Molinos, C.

2005-01-01

Full text: Neutron survey instruments are used routinely to determine the dose rates in areas where persons may be occupationally exposed. With a few exceptions, these instruments generally use a proportional counter with a high thermal neutron response located in a moderating sphere of CH 2 . The moderating sphere in such designs contains a thermal neutron absorber to reduce the over-response to thermal and intermediate energy neutrons. However, the commercially available examples of such instruments tend to have strongly energy dependent ambient dose equivalent response characteristics. In particular, they often over-respond in the energy range between 1 eV and 10 keV. A prototype of a novel design has been produced that uses seven detectors located in a moderating sphere of CH 2 , six near the surface to detect thermal and epithermal neutrons, and one in the centre to detect fast neutrons. This has been characterized using a combination of MCNP modelling and measurements to produce an instrument that has improved energy dependence of response characteristics. Additionally, the use of seven detectors offers direction and field hardness information. The design and calibration of the instrument are described and its response in workplaces calculated. (author)

Fast neutron (14.5 MeV) radiography: a comparative study

International Nuclear Information System (INIS)

Klann, R.T.

1996-01-01

Fast neutron (14.5 MeV) radiography is a type of non-destructive analysis tool that offers its own benefits and drawbacks. Because cross-sections vary with energy, a different range of materials can be examined with fast neutrons than can be studied with thermal neutrons, epithermal neutrons, or x-rays. This paper details these differences through a comparative study of fast neutron radiography to the other types of radiography available. The most obvious difference among the different types of radiography is in the penetrability of the sources. Fast neutrons can probe much deeper and can therefore obtain details of the internals of thick objects. Good images have been obtained through as much as 15 cm of steel, 10 cm of water, and 15 cm of borated polyethylene. In addition, some objects were identifiable through as much as 25 cm of water or 30 cm of borated polyethylene. The most notable benefit of fast neutron radiography is in the types of materials that can be tested. Fast neutron radiography can view through materials that simply cannot be viewed by X rays, thermal neutrons, or epithermal neutrons due to the high cross-sections or linear attenuation coefficients involved. Cadmium was totally transparent to the fast neutron source. Fast neutron radiography is not without drawbacks. The most pronounced drawback has been in the quality of radiograph produced. The image resolution is only about 0.8 mm for a 1.25 cm thick object, whereas, other forms of radiography have much better resolution

Action plan for the task: Physical measurements at the RA reactor related to VISA-2 project, '0' program, Reactor start-up and measurement of basic parameters of the new core; Plan rada po zadatku: Fizicka merenja na reaktoru RA u vezi projekta VISA-2, '0' program, Pustanje u rad reaktora RA i merenje osnovnih parametara novog jezgra

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-07-01

This report consists of two parts. Part one describes the RA reactor start-up, measurements of thermal neutron flux distribution, measurements of epithermal flux, fast neutron flux distribution, absolute values of both thermal and fast neutron fluxes, calibration of regulating rods, and measurements of neutron flux inside the fuel elements. All the mentioned measurements were done at low power level. Part two includes description of the reactor power increase up to nominal value of 6.5 MW, and measurements of thermal neutron flux distribution under xenon poisoning conditions, measurements of epithermal neutrons, absolute values of both thermal and fast neutron fluxes, and measurements of thermal and epithermal neutron fluxes at the exit of the horizontal experimental channel HK-d.

Hyper-thermal neutron irradiation field for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1994-01-01

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

Neutron flux determination at the IPR-R1 Triga Mark I neutron beam extractor

International Nuclear Information System (INIS)

Zangirolami, Dante Marco; Maretti Junior, Fausto; Ferreira, Andrea Vidal

2009-01-01

The IPR-R1 Triga Mark I Reactor located at the CDTN/CNEN, Belo Horizonte, Brazil, has been operating since November of 1960. In this work, measurements of thermal and epithermal neutron flux along the IPR-R1 neutron beam extractor were performed by neutron activation of reference materials using the two foils method. The obtained results were compared with results from two previous works: an experimental measurement done in a previous reactor core configuration and a numerical work made by Monte Carlo simulation using the actual reactor core configuration. The main purpose of this work is to update the measured data to the actual reactor core configuration. (author)

Neutron activation analysis methodology of marine sediments of the Cuban shelf

International Nuclear Information System (INIS)

Garcia, G.

1988-01-01

It is described the methodology followed for the neutron activation analysis of marine sediments of the Cuban shelf. A total of 35 elements were determined by means of the employment of thermal and epithermal fluxes of a nuclear reactor as well as from the flux of a generator of 14 MeV neutrons. It is signaled the particularities of detection and measurement of some elements and their interferences, with and special mention of the correction over 153 Sm, 113m In(Sn), 65 Zn, 160 Th, 169 Yb, 75 Sc and 177 Lu. It is concluded that 82% of elements may be determined with epithermal neutrons and 41%, with thermal neutrons. Only Si is determined with 14 MeV neutrons. Moreover, it is recommended the optimal periods of decaying (''cooling'') by groups of elements. The determined elements were: Ag, As, Au, Ba, Br, Ce, Co, Cr, Cs, Eu, Fe, Hf, K, La, Lu, Mo, Na, Nd, Ni, Rb, Sb, Sc, Se, Si, Sm, Sn, Sr, Ta, Tb, Th, U, W, Yb, Zn and Zr. The accuracy varied between 1% and 5% to 19 elements, between 6% and 10% to 5 elements, between 11% and 30% to 6 elements and it was greater than 30% to 5 elements

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

Science.gov (United States)

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

2017-04-01

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

A neutron beam polarizer for study of parity violation in neutron-nucleus interactions

International Nuclear Information System (INIS)

Penttilae, S.I.; Bowman, J.D.; Frankle, C.M.; Seestrom, S.J.; Yen, Yi-Fen; Delheij, P.P.J.; Haase, D.G.; Postma, H.

1994-01-01

A dynamically-polarized proton target operating at 5 Tesla and 1 K has been built to, neutron beam for studies of parity violation in compound-nuclear resonances. Nearly 0.9 proton polarization was obtained in an electron-beam irradiated ammonia target. This was used to produce a neutron beam polarization of 0.7 at epithermal energies. The combination of the polarized proton target and the LANSCE spallation neutron source produces the most intense pulsed polarized epithermal neutron beam in the world. The neutron-beam polarizer is described and methods to determine neutron beam polarization are presented

Multi-element neutron activation analysis of Brazilian coal samples

International Nuclear Information System (INIS)

Atalla, L.T.; Requejo, C.S.

1982-09-01

The elements U, Th, La, Ce, Nd, Sm, Eu, Dy, Tb, Yb, Lu, Sc, Ta, Hf, Co, Ni, Cr, Mo, Ti, V, W, In, Ga, Mn, Ba, Sr, Mg, Rb, Cs, K, Cl, Br, As, Sb, Au, Ca, Al and Fe were determined in coal samples by instrumental neutron activation analysis, by using both thermal and epithermal neutron irradiations. The irradiation times were 10 minutes and 8 or 16 hours in a position where the thermal neutron flux was about 10 12 n.cm - 2 .s - 1 and 72 non-consecutive hours for epithermal irradiation at a flux of about 10 11 n.Cm - 2 .s - 1 . After the instrumental analysis of the above mentioned elements, Zn and Se were determined with chemical separation. The relative standard deviation of, at least, 4 determinations was about + - 10% for the majority of the results. The coal samples analysed were supplied by: Cia. Estadual da Tecnologia e Saneamento Basico (CETESB-SP), Cia. de Pesquisas e Lavras Minerais (COPELMI-RS), Cia. Carbonifera Urussunga (SC), Cia. Carbonifera Prospera (SC), Cia. Carbonifera Treviso (SC), Cia. Nacional de Mineracao de Carvao do Barro Branco (SC) and Comissao Nacional de Energia Nuclear (CNEN-RJ). (Author) [pt

Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

Science.gov (United States)

El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

2008-09-01

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

A large animal model for boron neutron capture therapy

International Nuclear Information System (INIS)

Gavin, P.R.; Kraft, S.L.; DeHaan, C.E.; Moore, M.P.; Griebenow, M.L.

1992-01-01

An epithermal neutron beam is needed to treat relatively deep seated tumors. The scattering characteristics of neutrons in this energy range dictate that in vivo experiments be conducted in a large animal to prevent unacceptable total body irradiation. The canine species has proven an excellent model to evaluate the various problems of boron neutron capture utilizing an epithermal neutron beam. This paper discusses three major components of the authors study: (1) the pharmacokinetics of borocaptate sodium (NA 2 B 12 H 11 SH or BSH) in dogs with spontaneously occurring brain tumors, (2) the radiation tolerance of normal tissues in the dog using an epithermal beam alone and in combination with borocaptate sodium, and (3) initial treatment of dogs with spontaneously occurring brain tumors utilizing borocaptate sodium and an epithermal neutron beam

TREAT neutron-radiography facility

International Nuclear Information System (INIS)

Harrison, L.J.

1981-01-01

The TREAT reactor was built as a transient irradiation test reactor. By taking advantage of built-in system features, it was possible to add a neutron-radiography facility. This facility has been used over the years to radiograph a wide variety and large number of preirradiated fuel pins in many different configurations. Eight different specimen handling casks weighing up to 54.4 t (60 T) can be accommodated. Thermal, epithermal, and track-etch radiographs have been taken. Neutron-radiography service can be provided for specimens from other reactor facilities, and the capacity for storing preirradiated specimens also exists

Design of an irradiation facility with thermal, epithermal and fast neutron beams

International Nuclear Information System (INIS)

Pfister, G.; Bernnat, W.; Seidel, R.; Schatz, A.K.; Wagner, F.M.; Waschkowski, W.; Schraube, H.

1992-01-01

The main features of a neutron irradiation facility to be installed at the planned research reactor FRM-II are presented. In addition to the operational possibilities of the existing facility at the reactor FRM-I, the new facility will produce quasi-monoenergetic neutron fields and a neutron beam in the keV region whose spectrum can be modified by application of suitable filters and scatterers. For this beam, which is well suited for boron capture therapy, calculated boron reaction rates inside a phantom and an experimental verification of the calculations at the existing facility are presented. (orig.) [de

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Elimination of eight interfering radioisotopes in the determination of uranium by activation analysis with epithermic neutrons

International Nuclear Information System (INIS)

Requejo, C.S.

1977-01-01

The total or parcial elimination interfering radioisotopes in activation analysis of uranium by epithermic neutrons, has been made. It was possible to determine uranium, after chemical separation, from samples of organic and mineral matrixes, which had mercury, selenium, bromine, antimony, gold, barium, molybden and tungsten. Mineral samples were analysed giving results between 0.2 to 5.0 ppm of uranium. The same mineral were ground in agate mortar and in tungsten carbide mill. In the first sample is has been found 0.2277 +- -+ 0.0474 ppm U. The second which had tungsten, at level of 150 ppm, after radiochemical separation, it has been found 0.2465+- -+0.0326 ppm U. These results are considered statistically the same [pt

Spallation target-moderator-reflector studies at the Weapons Neutron Research facility

International Nuclear Information System (INIS)

Russell, G.J.; Gilmore, J.S.; Prael, S.D.; Robinson, H.; Howe, S.D.

1980-01-01

Basic neutronics data, initiated by 800-MeV proton spallation reactions, are important to spallation neutron source development and electronuclear fuel production. Angle-dependent and energy-dependent neutron production cross sections, energy-dependent and total neutron yields, thermal and epithermal neutron surface and beam fluxes, and fertile-to-fissile conversion ratios are being measured. The measurements are being done at the Weapons Neutron Research facility on a variety of targets and target-moderator-reflector configurations. The experiments are relevant to the above applications, and provide data to validate computer codes. Preliminary results are presented and compared to calculated predictions. 13 figures

Neutron activation determination of impurities in molybdenum

International Nuclear Information System (INIS)

Usmanova, M.M.; Mukhamedshina, N.M.; Obraztsova, T.V.; Saidakhmedov, K.Kh.

1984-01-01

Instrumental neutron-activation techniques of impurity element determination in molybdenum and MoO 3 (solid and powdered samples) have been developed. When determining impurities of Na, K, Mn, Cu, W, Re molybdenum has been irradiated by thermal neutrons in reactor for 20 min, the sample mass constituted 200-300 mg, sample cooling time after irradiation - 2.5-3.5 h. It is shown that in the process of Cr, Fe, Co, Zn determination the samples should be irradiated with thermal neutrons, and in the process of Sb, Ta and Ni determination - with resonance and fast neutrons. Simultaneous determination of the elements during irradiation with neutrons with reactor spectrum is possible. When determining P and S the samples are irradiated with thermal and epithermal neutrons and β-activity of samples and comparison samples are measured using β-spectrometer with anthracene crystal. The techniques developed permit to determine impurities in Mo with a relative standard deviation 0.07-0.15 and lower boundaries of contents determined - 10 -4 - 10 -7 %

neutron radiography

International Nuclear Information System (INIS)

Barton, J.P.

1993-01-01

Neutron radiography (or radiology) is a diverse filed that uses neutrons of various energies, subthermal, thermal, epithermal or fast in either steady state or pulsed mode to examine objects for industrial, medical, or other purposes, both microscopic and macroscopic. The applications include engineering design, biological studies, nondestructive inspection and materials evaluation. In the past decade, over 100 different centers in some 30 countries have published reports of pioneering activities using reactors, accelerators and isotopic neutron sources. While film transparency and electronic video are most common imaging methods for static or in motion objects respectively, there are other important data gathering techniques, including track etch, digital gauging and computed tomography. A survey of the world-wide progress shows the field to be gaining steadily in its diversity, its sophistication and its importance. (author)

Boron Neutron Capture Therapy at the TRIGA Mark II of Pavia, Italy - The BNCT of the diffuse tumours

Energy Technology Data Exchange (ETDEWEB)

Altieri, S.; Bortolussi, S.; Stella, S.; Bruschi, P.; Gadan, M.A. [University of Pavia (Italy); INFN - National Institute for Nuclear Physics, of Pavia (Italy)

2008-10-29

The selectivity based on the B distribution rather than on the irradiation field makes Boron neutron Capture Therapy (BNCT) a valid option for the treatment of the disseminated tumours. As the range of the high LET particles is shorter than a cell diameter, the normal cells around the tumour are not damaged by the reactions occurring in the tumoral cells. PAVIA 2001: first treatment of multiple hepatic metastases from colon ca by BNCT and auto-transplantation technique: TAOrMINA project. The liver was extracted after BPA infusion, irradiated in the Thermal Column of the Pavia TRIGA Mark II reactor, and re-implanted in the patient. Two patients were treated, demonstrating the feasibility of the therapy and the efficacy in destroying the tumoral nodules sparing the healthy tissues. In the last years, the possibility of applying BNCT to the lung tumours using epithermal collimated neutron beams and without explanting the organ, is being explored. The principal obtained results of the BNCT research are presented, with particular emphasis on the following aspects: a) the project of a new thermal column configuration to make the thermal neutron flux more uniform inside the explanted liver, b) the Monte Carlo study by means of the MCNP code of the thermal neutron flux distribution inside a patient's thorax irradiated with epithermal neutrons, and c) the measurement of the boron concentration in tissues by (n,{alpha}) spectroscopy and neutron autoradiography. The dose distribution in the thorax are simulated using MCNP and the anthropomorphic model ADAM. To have a good thermal flux distribution inside the lung epithermal neutrons must be used, which thermalize crossing the first tissue layers. Thermal neutrons do not penetrate and the obtained uniformity is poor. In the future, the construction of a PGNAA facility using a horizontal channel of the TRIGA Mark II is planned. With this method the B concentration can be measured also in liquid samples (blood, urine) and

Feasibility of sealed D-T neutron generator as neutron source for liver BNCT and its beam shaping assembly.

Science.gov (United States)

Liu, Zheng; Li, Gang; Liu, Linmao

2014-04-01

This paper involves the feasibility of boron neutron capture therapy (BNCT) for liver tumor with four sealed neutron generators as neutron source. Two generators are placed on each side of the liver. The high energy of these emitted neutrons should be reduced by designing a beam shaping assembly (BSA) to make them useable for BNCT. However, the neutron flux decreases as neutrons pass through different materials of BSA. Therefore, it is essential to find ways to increase the neutron flux. In this paper, the feasibility of using low enrichment uranium as a neutron multiplier is investigated to increase the number of neutrons emitted from D-T neutron generators. The neutron spectrum related to our system has a proper epithermal flux, and the fast and thermal neutron fluxes comply with the IAEA recommended values. Copyright © 2014 Elsevier Ltd. All rights reserved.

Uses of reactor neutrons for studying the microcomposition of materials

International Nuclear Information System (INIS)

Jervis, R.E.

1993-01-01

Reactor neutrons constitute excellents 'probes' for exploring and measuring a wide range both of minor and trace constituents in solids and liquids with high sensitivity because of their transparency in materials. Nondestructive neutron prompt-gamma analysis (PGA) utilizing either cold or thermal neutrons, such as at JRR-3M, is compared and contrasted to the more common (delayed) instrumental neutron activation analysis (INAA) and epithermal NAA. Clearly PGA offers high sensitivity for selected elements: B, H, Cd and REE's in suitable matrices, and is therefore, complementary to INAA which is not as useful for them, or for Ni, Sn, Fe, C or N. Recent INAA applications in our laboratory that demonstrate some of the uniqueness of neutron methods include use of epithermal neutrons for small biological specimens to measure Cd, K, As, Zn and, multielemental INAA for environmental pollution studies. The latter involves large data sets of multielemental concentrations which are subjected to statistical multivariant factor analysis to reveal unknown or unsuspected quantitative relationships among groups of trace constituents. These patterns, or 'factors' are shown to be uniquely related to pollution sources and can be utilized to compute the relative source contributions at a given receptor site. (author)

The determination of the space distribution, energy spectrum and dose parameters of thermal column beam resulting from swimming pool reactor

International Nuclear Information System (INIS)

Chen Changmao; Xie Jianlun; Leng Ruiping; Song Shushou; Su Jingling

1991-01-01

The axial and radial distribution, epithermal energy spectrum and dose equivalent rate of thermal column beam resulting from SPR have been determined in the Institute of Atomic Energy. The results show that the neutron fluence rate along the axial direction decreases as the distance increases outside the thermal column channel, and the trend of fluence rate attenuation follows approximately the inverse square law of a point source. When the reactor thermal power rate is 3 MW, at a distance of 50 cm to the channel, the thermal and epithermal neutron fluence rate are about 1.61 x 10 7 and 6.1 x 10 4 n/cm 2 · s respectively; dose equivalent rates are some 62 and 2.9 cSv/h respectively. At the end of the chennal, γ dose equivalent rate is 60 cSv/h or so

Notes on neutron flux measurement

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs

Pulsed thermal neutron source at the fast neutron generator.

Science.gov (United States)

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

2009-06-01

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

Fail-safe neutron shutter used for thermal neutron radiography

International Nuclear Information System (INIS)

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

1976-11-01

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

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

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

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

Deriving profiles of incident and scattered neutrons for TOF experiments with the spallation sources

International Nuclear Information System (INIS)

Watanabe, Hidehiro

1993-01-01

A formula that closely matches the incident profile of epi-thermal and thermal neutrons for time of flight experiments carried out with a spallation neutron source and moderator scheme is derived based on the slowing-down and diffusing-out processes in a moderator. This analytical description also enables us to predict burst-function profiles; these profiles are verified by a comparison with a diffraction pattern. The limits of the analytical model are discussed through the predictable peak position shift brought about by the slowing-down process. (orig.)

Design study of a medical reactor for Boron Neutron Capture Therapy

International Nuclear Information System (INIS)

Sasaki, M.; Hirota, J.; Tamao, S.; Kanda, K.; Mishima, Y.

1992-01-01

A new design study of a medical reactor for Boron Neutron Capture Therapy (BNCT) has been carried out. The reactor is to be used exclusively for the treatment of malignant melanoma and other cancers as well as for the further biomedical research. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of triangular lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam hole are to be provided to deliver thermal and epithermal neutrons. N-γ coupling Sn transport calculations indicate that the patient treatment period will be about 30 minutes with minimal fast neutron and gamma contaminants. (author)

Beam shaping assembly of a D-T neutron source for BNCT and its dosimetry simulation in deeply-seated tumor

Science.gov (United States)

Faghihi, F.; Khalili, S.

2013-08-01

This article involves two aims for BNCT. First case includes a beam shaping assembly estimation for a D-T neutron source to find epi-thermal neutrons which are the goal in the BNCT. Second issue is the percent depth dose calculation in the adult Snyder head phantom. Monte-Carlo simulations and verification of a suggested beam shaping assembly (including internal neutron multiplier, moderator, filter, external neutron multiplier, collimator, and reflector dimensions) for thermalizing a D-T neutron source as well as increasing neutron flux are carried out and our results are given herein. Finally, we have simulated its corresponding doses for treatment planning of a deeply-seated tumor.

The Swedish facility for boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Skoeld, K.; Capala, J. [Studsvik Medical AB (Sweden); Kierkegaard, J.; Haakansson, R. [Studsvik Nuclear AB (Sweden); Gudowska, I. [Karolinska Institute (Sweden)

2000-10-01

A BNCT (Boron Neutron Capture Therapy) facility has been constructed at the R2-0 reactor at Studsvik, Sweden. R2-0 is a 1 MW, open core, pool reactor. The reactor core is suspended on a movable tower and can be positioned anywhere in the pool. The BNCT facility includes two adjacent, parallel filter/moderator configurations and the reactor core is positioned in front of any of them as appropriate. One of the resulting neutron beams has been optimized for clinical irradiations with a filter/moderator system that allows easy variation of the neutron spectrum from the thermal to the epithermal energy range and with an extended collimator for convenient patient positioning. The other beam has been designed for radiobiological research and is equipped with a heavy water moderator and a large irradiation cavity with a uniform field of thermal neutrons. (author)

The Swedish facility for boron neutron capture therapy

International Nuclear Information System (INIS)

Skoeld, K.; Capala, J.; Kierkegaard, J.; Haakansson, R.; Gudowska, I.

2000-01-01

A BNCT (Boron Neutron Capture Therapy) facility has been constructed at the R2-0 reactor at Studsvik, Sweden. R2-0 is a 1 MW, open core, pool reactor. The reactor core is suspended on a movable tower and can be positioned anywhere in the pool. The BNCT facility includes two adjacent, parallel filter/moderator configurations and the reactor core is positioned in front of any of them as appropriate. One of the resulting neutron beams has been optimized for clinical irradiations with a filter/moderator system that allows easy variation of the neutron spectrum from the thermal to the epithermal energy range and with an extended collimator for convenient patient positioning. The other beam has been designed for radiobiological research and is equipped with a heavy water moderator and a large irradiation cavity with a uniform field of thermal neutrons. (author)

Epithermal neutron activation analysis of Spirulina platensis biomass, of the C-phycocianin and of DNA extracted from it

International Nuclear Information System (INIS)

Mosulishvili, L.M.; Belokobyl'skij, A.I.; Kirkesali, E.I.; Khizanishvili, A.I.; Frontas'eva, M.V.; Pomyakushina, E.V.

2002-01-01

The epithermal neutron activation analysis (ENAA) was used for study of the biomass of Spirulina platensis. The background levels of concentration of 27 macro-, micro- and trace elements ranging from 10 -3 up to 10 4 ppm were determined. It was found that the biomass of Spirulina does not contain toxic elements above the tolerance levels and can be utilized as a matrix of pharmaceuticals based on it. The concentrations of basic elements in C-phycocianin and DNA extracted from Spirulina platensis were determined by ENAA. A comparison of the element content of a whole Spirulina biomass with that of a refined C-phycocianin preparation was made

Epithermal Neutron Activation Analysis of Spirulina platensis Biomass, of the C-Phycocianin and of DNA Extracted from It

CERN Document Server

Mosulishvili, L M; Belokobylsky, A I; Kirkesali, E I; Khizanishvili, A I; Pomyakushina, E V

2002-01-01

The epithermal neutron activation analysis (ENAA) was used for study of the biomass of Spirulina platensis. The background levels of concentration 27 macro-, micro- and trace elements ranging from 10^{-3} up to 10^{4} ppm were determined. It was found that the biomass of spirulina does not contain toxic elements above the tolerance levels and can be utilized as a matrix of pharmaceuticals based on it. The concentrations of basic elements in C-phycocianin and DNA extracted from Spirulina platensis were determined by ENAA. A comparison of the element content of a whole spirulina biomass with that of a refined C-phycocianin preparation was made.

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

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1993-01-01

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

Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire. [Serbo-Croat] Ovaj izvestaj sadrzi sledece referate: Fluks termalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks epitermalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA; Fluks termalnih neurona u termalnoj koloni i bioloskom eksperimentalnom kanalu; Neutronska merenja u elementarnoj celiji reaktora RA; Temperaturni koeficijent reaktivnosti reaktora RA; Projekat uredjaja za merenje radioaktivnosti zice.

Determination of hydrogen content by neutron techniques

International Nuclear Information System (INIS)

Santisteban, J.R.; Granada, J.R.; Mayer, R.E.

1997-01-01

The commonly available techniques for the determination of hydrogen dissolved in solids are usually destructive from the point of view of the sample. A new, nondestructive method for this kind of measurements has been developed at our laboratory, with the requirement of improved sensitivity for massive samples. This scattering method is based on the use of epithermal neutrons, and has been implemented through the design and construction of a spectrometer dedicated to that task. In addition, the traditional transmission method has been employed to determine hydrogen content in metals, using the full sub thermal and thermal neutron energy ranges. A pulsed neutron source based on an electron LINAC is employed, together with time-of-flight techniques. In this work we will present some results illustrative of the sensitivity achieved by these neutron techniques in different systems and for a wide range of hydrogen concentrations. (author) [es

About neutron capture therapy method development at WWR-SM reactor in institute of Nuclear Physics of Uzbekistan Academy of Sciences

International Nuclear Information System (INIS)

Abdullaeva, G.A.; Baytelesov, S.A.; Dosimbaev, A.A.; Koblik, Yu.N.; Gritsay, O.O.

2006-01-01

Full text: Neutron capture therapy (NCT) is developing method of swellings treatment, on which specialists set one's serious hopes, as at its realization the practical possibilities of the effect on any swellings open. The essence of method is simple and lies in the fact that to the swelling enter preparation containing boron or gadolinium, which one have a large capture cross-section of the thermal and slow neutrons. Then the swelling is irradiated once with the slow (epithermal) neutron beam with fluency about 10 9 neutrons /sm 2 s for a short time and single. As a result of thermal neutrons capture by the boron (or gadolinium) nuclei secondary radiation which affecting swelling cells is emitted. NCT of oncologic diseases makes the specific demands to physical parameters of neutron beams. Now research reactors are often used for NCT. However, research reactor WWR-SM (INP, Uzbekistan AS, Tashkent) doesn't provide with the epithermal neutron beams and to develop this technique the reactor, first of all, needs for obtaining the epithermal neutron beams with energy spectrum in range from 1 eV up to 10 keV and with intensity ∼ 10 9 neutron /sm 2 s. Practically it is connected with upgrade of at least one of existed reactor channels, namely with equipping with the special equipment (filters), forming from the reactor spectrum the beam of necessary energy neutrons. It requires realization of preliminary model calculations, including calculations of capture cross-sections, of filters types and their geometrical parameters on the basis of optimal selected materials. Such calculations, as a rule, are carried out on the basis of Monte-Carlo method and designed software for calculation of nuclear reactor physical and technical characteristics [1]. In this work the calculation results of devices variants and problems discussion, related with possibility of WWR-SM reactor using for NCT are presented. (author)

Development of high flux thermal neutron generator for neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-01

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

Evaluation of JRR-4 neutron beam using tumor cells

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-01

For preparation of irradiation plan of boron-neutron capture therapy (BNCT), not only the physical dose is important, but also weighted factors or RBE are also necessary on the evaluation of the effect on the organism. Physical dose calculated by dose evaluation system (JCDS : JAERI Computational Dosimetry System) must appropriately carry out the weighting by various cells like tumor, central nerve, glia, and the vascular in proportion to JRR-4 each irradiation mode. In-vitro biological experiment which used 9L gliosarcoma and C6 glioma in the head water phantom was carried out in order to evaluate these effect. Neutron beam characteristics of JRR-4 were also evaluated from the functions of survival fraction of these cells. As a result of the evaluation, it became clear that the dose evaluation calculated from physical dose of the boron and nitrogen carried out in traditional BNCT of Japan using thermal neutron is applicable for thermal and epi-thermal mixed neutron beam. (author)

Epithermal neutron activation analysis of CR(VI)-reducer basalt-inhabiting bacteria

International Nuclear Information System (INIS)

Tsibakhashvili, N.Ya.; Kalabegishvili, T.L.; Murusidze, I.G.; Mosulishvili, L.M.; Frontas'eva, M.V.; Kirkesali, E.I.; Aksenova, N.G.; Holman, H.Y.

2005-01-01

Epithermal neutron activation analysis (ENAA) has been applied to studying elemental composition of Cr(VI)-reducer bacteria isolated from polluted basalts from the Republic of Georgia. Cr(VI)-reducing ability of the bacteria was examined by electron spin resonance (ESR) demonstrating that the bacteria differ in the rates of Cr(VI) reduction. A well-pronounced correlation between the ability of the bacteria to accumulate Cr(VI) and their ability to reduce Cr(V) to Cr(III) observed in our experiments is discussed. Elemental analysis of these bacteria also revealed that basalt-inhabiting bacteria are distinguished by relative contents of essential elements such as K, Na, Mg, Fe, Mn, Zn, and Co. A high rate of Cr(III) formation correlates with a high concentration of Co in the bacterium. ENAA detected some similarity in the elemental composition of the bacteria. The relatively high contents of Fe detected in the bacteria (140-340 μg/g of dry weight) indicate bacterial adaptation to the environmental conditions typical of the basalts. The concentrations of at least 12-19 different elements ranging from major- to ultratrace ones were determined in each type of bacteria simultaneously. The range of concentrations spans over 8 orders of magnitude

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

International Nuclear Information System (INIS)

Shaikh, A.M.

2013-01-01

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

Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of neutron flux distribution and its reactivity ratio

Energy Technology Data Exchange (ETDEWEB)

Bitelli, Ulysses d' Utra; Aredes, Vitor O.G.; Mura, Luiz E.C.; Santos, Diogo F. dos; Silva, Alexandre P. da, E-mail: ubitelli@ipen.br, E-mail: vitoraredes@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

When compared to a rectangular parallelepiped configuration the cylindrical configuration of a nuclear reactor core has a better neutron economy because in this configuration the probability of the neutron leakage is smaller, causing an increase in overall reactivity in the system to the same amount of fuel used. In this work we obtained a critical cylindrical configuration with the control rods 89.50% withdraw from the active region of the IPEN/MB-01 core. This is the cylindrical configuration minimum possible excess of reactivity. Thus we obtained a cylindrical configuration with a diameter of only 28 fuel rods with lowest possible excess of reactivity. For this purpose, 112 peripheral fuel rods are removed from standard reactor core (rectangular parallelepiped of 28x28 fuel rods). In this configuration the excesses of reactivity is approximated 279 pcm. From there, we characterize the neutron field by measuring the spatial distribution of the thermal and epithermal neutron flux for the reactor operating power of 83 watts measured by neutron noise analysis technique and 92.08± 0.07 watts measured by activation technique [10]. The values of thermal and epithermal neutron flux in different directions, axial, radial north-south and radial east-west, are obtained in the asymptotic region of the reactor core, away from the disturbances caused by the reflector and control bar, by irradiating thin gold foils infinitely diluted (1% Au - 99% Al) with and without (bare) cadmium cover. In addition to the distribution of neutron flux, the moderator temperature coefficient, the void coefficient, calibration of the control rods were measured. (author)

Earth formation porosity log using measurement of neutron energy spectrum

International Nuclear Information System (INIS)

1981-01-01

Methods and apparatus are described for measuring the porosity of subsurface earth formations in the vicinity of a well borehole by means of neutron well logging techniques. All the commercial techniques for measuring porosity currently available are not as accurate as desirable due to variations in the borehole wall diameter, in the borehole fluids (e.g. with chlorine content) in the casings of the borehole etc. This invention seeks to improve accuracy by using a measurement of the epithermal neutron population at one detector and the fast neutron population at a second detector, spaced approximately the same distance from a neutron source. The latter can be detected either by a fast neutron detector or indirectly by an inelastic gamma ray detector. Background correction can be made, and special detectors used, to discriminate against the detection of thermal neutrons or their resultant capture gamma rays. These fluctuations affect the measurement of thermal neutron populations. (U.K.)

Comparison between different flux traps assembled in the core of the nuclear reactor IPEN/MB-01 by measuring of the thermal and epithermal neutron fluxes using activation foils

International Nuclear Information System (INIS)

Mura, Luiz Ernesto Credidio; Bitelli, Ulysses d'Utra; Mura, Luis Felipe Liambos; Carluccio, Thiago; Andrade, Graciete Simoes de

2011-01-01

The production of radioisotopes is one of the most important applications of nuclear research reactors. This study investigated a method called Flux Trap, which is used to increase the yield of production of radioisotopes in nuclear reactors. The method consists in the rearrangement of the fuel rods to allow the increase of the thermal neutron flux in the irradiation region inside the reactor core, without changing the standard reactor power level. Various configurations were assembled with the objective of finding the configuration with the highest thermal neutron flux in the region of irradiation. The method of activation analysis was used to measure the thermal neutron flux and determine the most efficient reactor core configuration . It was found that there was an increase in the thermal neutron flux of 337% in the most efficient configuration, which demonstrates the effectiveness of the method. (author)

Thermal neutron polarisation

International Nuclear Information System (INIS)

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

1984-01-01

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

Neutron flux characterization using Lr-115 NTD and binary glass metal as converter

International Nuclear Information System (INIS)

Alvarado, R.; Palacios, D.; Sajo B, L.; Greaves, E.; Barros, H.; Nemeth, P.; Goncalves, I. F.

2010-01-01

The installed neutron irradiation cell at the Simon Bolivar University Laboratory is employed frequently for applications where the neutron field should be monitored. The ratio of epithermal and thermal neutron flux were determined by the cadmium difference method in three different irradiation locations of a graphite thermalized 252 Cf source. Monte Carlo simulation provided the expected thermal neutron field inside the irradiation cell and the nuclear trace detectors sensitivity. The passive device is a Kodak-Pathe Lr-115 externally coated with a thin (0,1 mm) binary glass metal strip Fe 75 B 25 where the 10 B(n,α) 7 Li reaction converts free neutrons into alpha particles that are registered as etched tracks. The results allow to determine the neutron flux in an easy, reproducible and effective way, with the advantage of a low self-absorption effect due to the converter material. Improved etching process of the Lr-115 nuclear track detectors and the software parameters for track analysis are included. (Author)

Study on iodine levels in thyroids of iodine-supplemented rats by epithermal neutron activation analysis

International Nuclear Information System (INIS)

Wang Xuefei; Zhang Fang; Xu Qing; Liu Nianqing; Chai Zhifang; Zhao Xueqin; Zuo Aijun

2003-01-01

The second generation female Wistar rats that have been treated with iodine-deficient food, after their delivery, are divided into three groups i.e. excessive-iodine (EI), adequate-iodine (AI) and iodine-deficient (ID) according to the KIO 3 concentration in the drinking water (3.0, 0.4, 0 mg/L). In addition, the normal rats with low iodine food and 0.4 mg/L KIO 3 water are used as the control group (C). The iodine content in thyroid and the serum thyroid hormone levels of the third generation rats are measured by means of epithermal neutron activation analysis (ENAA), and the method of enzyme-linked immunosorbent assay (ELISA), respectively. The results indicate that the total thyroxine (TT 4 ) and the free thyroxine (FT 4 ) of the EI, compared with those of the controls, are significantly decreased (p 3 ) evidently increased (p 4 , FT 4 and goiter

A neutron irradiator applied to cancer treatment

International Nuclear Information System (INIS)

Campos, Tarcisio P.R.; Andrade, Ana P. de

2000-01-01

Cancer and the way of treating it with neutron capture therapy are addressed. This paper discusses also the type of neutron facilities used to treat cancer around the world, as follow: discrete neutron sources, accelerators, and nuclear reactors. The major features of an epithermal neutron irradiation facility applied to BNCT treatment are addressed. The main goal is to give another choice of neutron irradiators to be set in a hospital. The irradiation facility embeds a set of 252 Cf neutron source coupled with a homogeneous mixture of uranium-zirconium hydride alloy containing 8.4 wt % uranium enriched to 20% U 235 . The facility delivers an epithermal neutron beam with low background of fast neutron and gamma rays. The N particle transport code (MCNP-4A) has been used during the simulation in order to achieve the desired configurations and to estimate the multiplication factor, k eff . The present facility loaded with 30 mg of 252 Cf neutron source generates an external beam with an intensity of 10 7 n/cm 2 .s on the spectrum of 4 eV to 40 KeV. The 252 Cf - facility coupled with fissile material was able to amplify the epithermal flux to 10 8 n/cm 2 .s, maintaining the figure-of-merits represented by the ratios of the fast dose and gamma dose in air per epithermal neutron flux closed to those values presented by BMRR, MITR-II and Petten Reactor. The medical irradiation facility loaded with 252 Cf- 235 U can be a choice for BNCT. (author)

Determination of elemental concentrations in airborne particulate matter in the City of Santiago de Chile, through neutron activation analysis using epithermal neutrons and Compton suppression system

International Nuclear Information System (INIS)

Rojas S, Ximena

1995-01-01

In order to optimize the Neutron-Activation Analysis (NAA) technique currently carried out in our country, the present work was carried out in the United States where irradiations with epithermal neutrons and a Compton suppression system were used, which allowed the characterization of aerosols of the city of Santiago de Chile. With this purpose, 54 filters of polycarbonate membranes were analysed with aerosols collected in an area of the capital during Spring 1993 and Winter of 1994. As a result, an improvement in the detection limits was observed, specially in elements such as Ni and Zn, which are not easily detectable through NAA. The application of both systems also permits the usage of this technique in geological and biological samples, where the presence of Na, Al and Cl obstruct the determination of some elements. The determined elements in both fractions were Mn, V, Cu, As, Sb, Co, Br, Cl, Ni, Zn, Ca, Al, Na and Fe. (author). 8 refs., 7 figs., 9 tabs

Calculations of neutron source at the KYIV research reactor for the boron neutron capture therapy aims

International Nuclear Information System (INIS)

Gritzay, O.; Kalchenko, O.; Klimova, N.; Razbudey, V.; Sanzhur, A.

2006-01-01

Calculation results of an epithermal neutron source which can be created at the Kyiv Research Reactor (KRR) by means of placing of specially selected moderators, filters, collimators, and shielding into the 10-th horizontal experimental tube (so-called thermal column) are presented. The general Monte-Carlo radiation transport code MCNP4C [1], the Oak Ridge isotope generation code ORIGEN2 [2] and the NJOY99 [3] nuclear data processing system have been used for these calculations

The instrumental neutron-activation analysis of granites from the Bushveld Complex

International Nuclear Information System (INIS)

Watterson, J.I.W.

1978-01-01

Three methods of instrumental neutron-activation analysis, 14MeV, reactor thermal, and reactor epithermal, are compared for the analysis of granites form the Bushveld Complex. A total of 34 elements can be determined in the granites by these methods. Samples from the Zaaiplaats area were analysed by thermal neutron activation, and 22 elements were determined in all of them. These elements were used to distinguish between the mineralized Bobbejaankop and Lease granites and the Main granite by the use of multivariate statistics. The Bobbejaankop granite appears as a more-differentaited rock carrying greater amounts of the incompatible elements than does the Main granite [af

Studsvik thermal neutron facility

International Nuclear Information System (INIS)

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

1992-01-01

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

The epithermal critical experiments; Experiences critiques avec des neutrons epitliermiques; Nadteplovye kriticheskie ehksperimenty; Experimentos criticos con neutrones epitermicos

Energy Technology Data Exchange (ETDEWEB)

Morewitz, H A; Carpenter, S O [Atomics International, Canoga Park, CA (United States)

1962-03-15

The epithermal critical experiments. The present phase of the advanced-epithermal-thorium-reactor programme consists of integral-reactor-physic s experiments designed to provide neutron-cross-section information in the 10-MeV to 1-keV range. A series of nine, multi-region, slow-fast, pseudospherica l critical assemblies of the honey- comb, split-table type are being studied. So far, three assemblies have' been run. The outer driver-decouple r region drives an interior U{sup 233}-Th fuelled spherical test region whose neutron-flux spectrum is successively degraded by increasing the graphite moderator to fuel ratio. A square-wave oscillator experiment defines the central reactivity worths of forty small samples of different materials to 10{sup -8} {Delta}k for each assembly. Additionally, intercalibrated artificial neutron sources are oscillated to determine the various central neutron importance functions. The spectra are obtained by fission-counter measurements with calibrated foils of different thresholds and by a Li{sup 6}-solid-state- counter sandwich spectrometer. A digital computer routine will be used to compile all measurements into a self-consistent library of spectrum averaged cross-sections. (author) [French] La phase actuelle du programme de reacteur au thorium a neutrons epithermiques comprend des experiences integrales de physique des reacteurs pour obtenir des renseignements sur les sections efficaces neutroniques pour la gamme d'energie comprise entre 1 keV et 10 MeV. Les auteurs etudient une serie de neuf ensembles critiques pseudospheriques, a plusieurs regions, a couplage neutrons lents et neutrons rapides du type a alveoles et a coeur divise. A ce jour, trois de ces ensembles ont ete mis en service. La region exterieure, mettant en service ou hors service, commande une zone d'essai interieure de forme spherique ou le combustible est constitue de {sup 233}U-Th, dont le spectre du flux de neutrons est degrade progressivement par augmentation du

Medical and biological requirements for boron neutron capture therapy

International Nuclear Information System (INIS)

Gahbauer, R.; Goodman, J.H.; Kanellitsas, C.; Clendenon, N.; Blue, J.

1986-01-01

In conventional radiation therapy, tumor doses applied to most solid tumors are limited by the tolerance of normal tissues. The promise of Boron Neutron Capture Therapy lies in its potential to deposit high doses of radiation very specifically to tumor tissue. Theoretically ratios of tumor to normal tissue doses can be achieved significantly higher than conventional radiotherapeutic techniques would allow. Effective dose distributions obtainable are a complex function of the neutron beam characteristics and the macro and micro distributions of boron in tumor and normal tissues. Effective RBE doses are calculated in tumors and normal tissue for thermal, epithermal and 2 keV neutrons

Simulation study of accelerator based quasi-mono-energetic epithermal neutron beams for BNCT.

Science.gov (United States)

Adib, M; Habib, N; Bashter, I I; El-Mesiry, M S; Mansy, M S

2016-01-01

Filtered neutron techniques were applied to produce quasi-mono-energetic neutron beams in the energy range of 1.5-7.5 keV at the accelerator port using the generated neutron spectrum from a Li (p, n) Be reaction. A simulation study was performed to characterize the filter components and transmitted beam lines. The feature of the filtered beams is detailed in terms of optimal thickness of the primary and additive components. A computer code named "QMNB-AS" was developed to carry out the required calculations. The filtered neutron beams had high purity and intensity with low contamination from the accompanying thermal, fast neutrons and γ-rays. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

Directory of Open Access Journals (Sweden)

Kaiser Krista

2016-01-01

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

Neutron spectra and H*(10) around and 18 MV Linac by Ann's

International Nuclear Information System (INIS)

Banuelos F, A.; Valero L, C.; Borja H, C. G.; Hernandez D, V. M.; Vega C, H. R.

2011-10-01

Neutron spectra and ambient dose equivalent H*(10) were calculated for a radiotherapy room in 16 point-like detectors, 15 located inside the vault room and 1 located outside the bunker. The calculation was carried out using Monte Carlo Methods with the MCNP5 code for a generic radiotherapy room model operating with a 18 MV Linac, obtaining 16 neutron spectra with 47 energy bins, the H*(10) values were calculated from the neutron spectra by the use of the fluence-dose conversion factors. An artificial neural network were designed and trained to determine the neutron H*(10) in 15 different locations inside the vault room from the H*(10) dose calculated for the detector located outside the room, using the calculated dose values as training set, using the scaled conjugated gradient training algorithm. The mean squared error set for the network training was 1E(-14), adjusting the data in 99.992 %. In the treatment hall, as the distance respect to the isocenter is increased, the amount of neutrons and the H*(10) are reduced, neutrons in the high-energy region are shifted to lower region peaking around 0.1 MeV, however the epithermal and thermal neutrons remain constant due to the room-return effect. In the maze the spectra are dominated by epithermal and thermal neutrons that contributes to produce activation and the production of prompt gamma-rays. The results shows the using this artificial intelligence technic as a useful tool for the neutron spectrometry and dosimetry by the simplification on the neutronic fields characterization inside radiotherapy rooms avoiding the use of traditional spectrometric systems. And once the H*(10) doses have been calculated, to take the appropriated actions to reduce or prevent the patient and working staff exposure to this undesirable neutron radiation. (Author)

Internal-standard method for the determination of uranium, thorium, lanthanum and europium in carbonaceous shale and monazite by epithermal neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Chen, Shuenn-Gang; Tsai, Hui-Tuh; Wu, Shaw-Chii [Institute of Nuclear Energy Research, Lung-Tan (Taiwan, Republic of China)

1981-10-03

An internal-standard method was applied for the determination of uranium, thorium, lanthanum and europium is carbonaceous shale samples and monazite sand by epithermal neutron activation analysis using gold as an internal standard element. The samples were irradiated in a zero-power reactor at the Institute of Nuclear Energy Research and measured with a high-resolution Ge(Li) detector. The detection limit is 0.1 ppm for uranium and europium, 1 ppm for thorium, 5 ppm for lanthanum, and the realative error of all elements is within +-2.6%.

Development of slow neutron dose meter by track counting

International Nuclear Information System (INIS)

Nozaki, Tetsuya; Takeuchi, Hiroshi; Hasnel, S.; Honda, Teruyuki; Harasawa, Susumu.

1993-01-01

Prototypes of a plate type and two types of semispherical dosemeter were manufactured and their performances were tested. Polyallyl diglycol carbonate (PADC) was employed as neutron detector and covered with natural LiF and 6 Li-enriched LiF ceramics. They were irradiated in the TRIGA II Reactor of the Rikkyo University, and the sensitivity characteristics for incident angles and neutron energies were analyzed. It is concluded that it may be possible to manufacture small sized neutron dosemeter of flat response to wide energy range from thermal neutrons to epithermal neutrons with sufficient sensitivity for personal monitoring, using LiF ceramic as neutron filters and neutron converters. However the following issues are to be solved: the optimization of thickness of the LiF filter, the effects of albedo of the human body, and the applicability to the intermediate neutrons. (A.Y.)

The background cross section method for calculating the epithermal neutron spectra

International Nuclear Information System (INIS)

Martinez, A.S.

1983-01-01

We have developed a new methodology to the multigroup constants calculations, for thermal and fast reactors. The method to obtain the constants is extremely fast and simple, and it avoid repeated computations of the detailed neutron spectrum for different cell configurations (composition, geometry and temperature). (author) [pt

Physical engineering and medical physics on boron neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori

2011-01-01

The contents of physical engineering and medical physics that support boron neutron capture therapy (BNCT) can be roughly classified to the four items, (1) neutron irradiation system, (2) development and improvement of dose assessment techniques, (3) development and improvement of dose planning system, and (4) quality assurance and quality control. This paper introduces the BNCT at Kyoto University Research Reactor Institute, with a focus on the basic physics of BNCT, thermal neutron irradiation and epithermal neutron irradiation, heavy water neutron irradiation facilities of KUR, and medical irradiation system of KUR. It also introduces the world's first BNCT clinical cyclotron irradiation system (C-BENS) of Kyoto University Research Reactor Institute, BNCT dose assessment techniques, dose planning system, and quality assurance and quality control. (A.O.)

An accelerator neutron source for BNCT. Technical progress report, 1 June 1993--31 May 1994

International Nuclear Information System (INIS)

Blue, T.E.; Vafai, K.

1994-02-01

This is the progress report for the project entitled, ''An Accelerator Neutron Source for BNCT.'' The progress report is for the period from July 1, 1993 to date. The overall objective of our research project is to develop an Accelerator Epithermal Neutron Irradiation Facility (AENIF) for Boron Neutron Capture Therapy (BNCT). The AENIF consists of a 2.5 MeV high current proton accelerator, a lithium target to produce source neutrons, and a moderator/reflector assembly to obtain from the energetic source neutrons an epithermal neutron field suitable for BNCT treatments. Our project goals are to develop the non-accelerator components of the AENIF, and to specifically include in our development: (1) design, numerical simulation, and experimental verification of a target assembly which is capable of removing 75 kW of beam power; (2) re-optimization of the moderator assembly design based on in-phantom dose assessments using neutron spectra calculated in phantom and an energy-dependent neutron Relative Biological Effectiveness (RBE); (3) construction of a prototype moderator assembly and confirmation of its design by measurements; (4) design of the shielding of the accelerator and treatment rooms for an AENIF; and (5) design of a high energy beam transport system which is compatible with the shielding design and the thermal-hydraulic design

Progress in study of a medical reactor for boron neutron capture therapy

International Nuclear Information System (INIS)

Sasaki, Makoto; Hirota, Jitsuya; Tamao, Shigeo; Kanda, Keiji; Mishima, Yutaka.

1993-01-01

A design study of a medical reactor for Boron Neutron Capture Therapy has made progress. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of hexagonal lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam holes are to be provided for simultaneous treatments by thermal and epithermal neutrons and for further biomedical research. The design objectives for the beam holes are to deliver the therapeutic doses in a modest time (30 to 60 min) with minimal fast neutron and gamma contaminants. The n-γ coupling Sn transport calculations have been carried out using n-21 and γ-9 group cross sections on 2-dim. practical models. The calculated results indicate that the design objectives will be achievable even if the thermal power of the reactor is reduced to 1 MW. (author)

An Assessment of Elemental Compositions and Concentrations in a Marble Sample from Hatnub Area using Internal Mono Standard Neutron Activation Analysis

International Nuclear Information System (INIS)

Soliman, N.F.; EL-Shershaby, A.; EL-Bahi, S.; Walley El-Dine, N.; Abd El-Halim, E.S.; Afifi, S.

2012-01-01

A marble sample was taken from Hatnub area near Asyot, where a high pollution is expected. The sample was specially treated and prepared for elemental analysis using the k 0 -based internal mono standard neutron activation analysis . The irradiation facility of the first Egyptian Training Research Reactor (ETRR-1) was used to irradiate the sample together with gold and zirconium monitor samples. The pneumatic irradiation rabbit system (PIRS) built in the vertical thermal column of the second Egyptian Training Research Reactor (ETRR-2) was used to irradiate another weight of the marble sample and a sample from a certificated reference material JB-1 for short time irradiation. A new method was developed to measure the thermal to epithermal neutron flux ratio and the deviation in the real epithermal neutron spectrum ( 1/E 1±a lpha ) from the ideal (1/E ). The elemental analysis was carried out by high-resolution gamma-ray spectrometry. The accuracy of the internal mono standard method has been evaluated by analyzing the JB-1 certified reference material

Experiences of reconstruction of the epithermal neutron beam at THOR

International Nuclear Information System (INIS)

Liu Hongming; Hsu Pinchieh; Liu Chaochin; Jiang Shianghuei; Liu Yenwan Hsueh; Kai Jijung

2006-01-01

Tsing Hua Open-pool Reactor (THOR) had completed the renovation for an epithermal neutron beam in August 2004. The major tasks for this renovation were moderator/filter design and assembling, and concrete cutting for a better beam quality and larger irradiation room. Besides moderator/filter design, the associated works involved radiation monitoring, structure analysis, and shielding design. The radiation monitoring was performed to predict the probable accumulated dose for the workers involved in this reconstruction project. Special shielding design and construction processes were adopted to lower the radiation level and the probable accumulated dose for the workers. Before concrete cutting, structure analysis based on SAP-2000 code was performed to assure the structure is safe from the earthquake in Taiwan. A wall saw was then used for concrete cutting to enlarge the space of the irradiation room. Moderator/filter components were assembled on a trolley outside the beam exit prior to installation, which can effectively reduce the duration of a worker staying inside the reconstruction area and thereby reduce the accumulated dose. The shielding for the irradiation room was designed based on MCNP simulation using a pre-calculated source plane at the beam exit. The thickness of the concrete (density=3 g/cm 3 ) of the walls and ceiling of the irradiation room were designed to be 100cm. On-going tasks include beam parameters measurement and in vitro/ in vivo study and calibration of treatment planning system, with the hope that the team can be ready for clinical trials in 2-3 years. (author)

Assessment of Laser-Driven Pulsed Neutron Sources for Poolside Neutron-based Advanced NDE – A Pathway to LANSCE-like Characterization at INL

Energy Technology Data Exchange (ETDEWEB)

Roth, Markus [Technische Univ. Darmstadt (Germany); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bourke, Mark Andrew M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fernandez, Juan Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mocko, Michael Jeffrey [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Glenzer, Siegfried [Stanford Univ., CA (United States); Leemans, Wim [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Siders, Craig [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Haefner, Constantin [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2017-04-19

A variety of opportunities for characterization of fresh nuclear fuels using thermal (~25meV) and epithermal (~10eV) neutrons have been documented at Los Alamos National Laboratory. They include spatially resolved non-destructive characterization of features, isotopic enrichment, chemical heterogeneity and stoichiometry. The LANSCE spallation neutron source is well suited in neutron fluence and temporal characteristics for studies of fuels. However, recent advances in high power short pulse lasers suggest that compact neutron sources might, over the next decade, become viable at a price point that would permit their consideration for poolside characterization on site at irradiation facilities. In a laser-driven neutron source the laser is used to accelerate deuterium ions into a beryllium target where neutrons are produced. At this time, the technology is new and their total neutron production is approximately four orders of magnitude less than a facility like LANSCE. However, recent measurements on a sub-optimized system demonstrated >10¹⁰ neutrons in sub-nanosecond pulses in predominantly forward direction. The compactness of the target system compared to a spallation target may allow exchanging the target during a measurement to e.g. characterize a highly radioactive sample with thermal, epithermal, and fast neutrons as well as hard X-rays, thus avoiding sample handling. At this time several groups are working on laser-driven neutron production and are advancing concepts for lasers, laser targets, and optimized neutron target/moderator systems. Advances in performance sufficient to enable poolside fuels characterization with LANSCE-like fluence on sample within a decade may be possible. This report describes the underlying physics and state-of-the-art of the laser-driven neutron production process from the perspective of the DOE/NE mission. It also discusses the development and understanding that will be necessary to provide customized capability for

Use of neutron capture gamma radiation for determining grade of iron ore in blast holes and exploration holes

International Nuclear Information System (INIS)

Eisler, P.L.; Huppert, P.; Mathew, P.J.; Wylie, A.W.; Youl, S.F.

1977-01-01

Neutron radiative capture and neutron-neutron logging have been applied to determining the grade of ore in dry blast holes and a dry exploration hole drilled into a layered iron deposit. Both thermal and epithermal neutron responses were measured as well as the gamma-ray responses due to neutron capture by iron and by hydrogen present in hydrated minerals. The results were fitted by a stepwise multiple linear regression technique to give expressions for mean grade of ore in the drill hole and 95% confidence intervals for estimation of this mean. For an overall range of ore grades of 20-68% Fe and a mean grade of 63% Fe, the confidence interval for prediction of mean grade for the neutron-gamma technique was 0.3% Fe for pooled data from all five blast holes and 0.8% Fe for a single hole. It was also shown that for this type of layered deposit a simpler neutron-neutron log incorporating simultaneous measurement of both thermal and epithermal neutron responses gave almost as good a grade prediction result for pooled results from five drill holes, namely 63+-0.4% Fe, as that obtained by the neutron-gamma technique. The results of both types of log are compared with those obtained by the spectral gamma-ray backscattering [Psub(z)] technique, or by logging of natural gamma radiations from the shale component of the ore. From this comparison conclusions are drawn regarding the most suitable technique to employ for determining grade of iron ore in various practical logging situations. (author)

Neutron activation analysis in Romania

International Nuclear Information System (INIS)

Apostolescu, St.

1985-01-01

The following basic nuclear facilities are used for neutron activation analysis: a 2000 KW VVR-S Nuclear Reactor, a U-200 Cyclotron, a 30 MeV Betatron, several 14 MeV neutron generators and a king size High Voltage tandem Van de'Graaff accelerator. The main domains of application of the thermal neutron activation analysis are: geology and mining, processing of materials, environment and biology, achaeology. Epithermal neutron activation analysis has been used for determination of uranium and thorium in ores with high Th/U ratios or high rare earth contents. One low energy accelerator, used as 14.1 Mev neutron source, is provided with special equipmen for oxigen and low mass elements determination. An useful alternating way to support fast neutron activation analysis is an accurate theoretical description of the fast neutron induced reactions based on the statistical model (Hauser-Feubach STAPRE code) and the preequilibrium decay geometry dependent model. A gravitational sample changer has been installed at the end of a beam line of the Cyclotron, which enables to perform charged particles activation analysis for protein determination in grains

High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Guillen, Donna; Greenwood, Lawrence R.; Parry, James

2014-06-22

A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

Neutron spectra and H*(10) around and 18 MV Linac by Ann's

Energy Technology Data Exchange (ETDEWEB)

Banuelos F, A.; Valero L, C.; Borja H, C. G.; Hernandez D, V. M.; Vega C, H. R., E-mail: alanb535@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2011-10-15

Neutron spectra and ambient dose equivalent H*(10) were calculated for a radiotherapy room in 16 point-like detectors, 15 located inside the vault room and 1 located outside the bunker. The calculation was carried out using Monte Carlo Methods with the MCNP5 code for a generic radiotherapy room model operating with a 18 MV Linac, obtaining 16 neutron spectra with 47 energy bins, the H*(10) values were calculated from the neutron spectra by the use of the fluence-dose conversion factors. An artificial neural network were designed and trained to determine the neutron H*(10) in 15 different locations inside the vault room from the H*(10) dose calculated for the detector located outside the room, using the calculated dose values as training set, using the scaled conjugated gradient training algorithm. The mean squared error set for the network training was 1E(-14), adjusting the data in 99.992 %. In the treatment hall, as the distance respect to the isocenter is increased, the amount of neutrons and the H*(10) are reduced, neutrons in the high-energy region are shifted to lower region peaking around 0.1 MeV, however the epithermal and thermal neutrons remain constant due to the room-return effect. In the maze the spectra are dominated by epithermal and thermal neutrons that contributes to produce activation and the production of prompt gamma-rays. The results shows the using this artificial intelligence technic as a useful tool for the neutron spectrometry and dosimetry by the simplification on the neutronic fields characterization inside radiotherapy rooms avoiding the use of traditional spectrometric systems. And once the H*(10) doses have been calculated, to take the appropriated actions to reduce or prevent the patient and working staff exposure to this undesirable neutron radiation. (Author)

Modelisation and distribution of neutron flux in radium-beryllium source (226Ra-Be)

Science.gov (United States)

Didi, Abdessamad; Dadouch, Ahmed; Jai, Otman

2017-09-01

Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium-beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium-beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium-beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.

Accelerator Based Neutron Beams for Neutron Capture Therapy

International Nuclear Information System (INIS)

Yanch, Jacquelyn C.

2003-01-01

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

Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution

International Nuclear Information System (INIS)

Aredes, Vitor Ottoni Garcia

2014-01-01

In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

Coupled fast-thermal system at the RB, masters thesis; Spregnuti brzo-termicki sistem na reaktoru RB, magistarski rad

Energy Technology Data Exchange (ETDEWEB)

Pesic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1984-05-15

Coupled fast-thermal system at the RB reactor was formed owing to availability of highly enriched fuel. This paper deals with reactor parameters calculations and measurements of coupled core taking into account safety constraints. Validity of applied calculation methods was confirmed. The following parameters were analyzed: critical height of the core; reactivity dependent on heavy water level in the core; fast neutron spectrum in the fast region channel; spatial distribution of thermal. epithermal and fat neutrons in the fast region channel; reactivity of safety rods; neutron and gamma absorption doses in the center of the coupled core.

Characterization of bauxite residue (red mud) for 235U, 238U, 232Th and 40K using neutron activation analysis and the radiation dose levels as modeled by MCNP.

Science.gov (United States)

Landsberger, S; Sharp, A; Wang, S; Pontikes, Y; Tkaczyk, A H

2017-07-01

This study employs thermal and epithermal neutron activation analysis (NAA) to quantitatively and specifically determine absorption dose rates to various body parts from uranium, thorium and potassium. Specifically, a case study of bauxite residue (red mud) from an industrial facility was used to demonstrate the feasibility of the NAA approach for radiological safety assessment, using small sample sizes to ascertain the activities of 235 U, 238 U, 232 Th and 40 K. This proof-of-concept was shown to produce reliable results and a similar approach could be used for quantitative assessment of other samples with possible radiological significance. 238 U and 232 Th were determined by epithermal and thermal neutron activation analysis, respectively. 235 U was determined based on the known isotopic ratio of 238 U/ 235 U. 40 K was also determined using epithermal neutron activation analysis to measure total potassium content and then subtracting its isotopic contribution. Furthermore, the work demonstrates the application of Monte Carlo Neutral-Particle (MCNP) simulations to estimate the radiation dose from large quantities of red mud, to assure the safety of humans and the surrounding environment. Phantoms were employed to observe the dose distribution throughout the human body demonstrating radiation effects on each individual organ. Copyright © 2016 Elsevier Ltd. All rights reserved.

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

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

Semiconductor Thermal Neutron Detector

Directory of Open Access Journals (Sweden)

Toru Aoki

2014-02-01

Full Text Available The  CdTe  and  GaN  detector  with  a  Gd  converter  have  been developed  and  investigated  as  a  neutron  detector  for neutron  imaging.  The  fabricated  Gd/CdTe  detector  with  the  25  mm  thick  Gd  was  designed  on  the  basis  of  simulation results  of  thermal  neutron  detection  efficiency  and  spatial  resolution.  The  Gd/CdTe  detector  shows  the  detection  of neutron  capture  gamma  ray  emission  in  the  155Gd(n,  g156Gd,  157Gd(n,  g158Gd  and  113Cd(n,  g114Cd  reactions  and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application. Moreover, a BGaN detector has also investigated to separate neutron signal from gamma-ray clearly. 

The Thermal Neutron Beam Option for NECTAR at MLZ

Science.gov (United States)

Mühlbauer, M. J.; Bücherl, T.; Genreith, C.; Knapp, M.; Schulz, M.; Söllradl, S.; Wagner, F. M.; Ehrenberg, H.

The beam port SR10 at the neutron source FRM II of Heinz Maier-Leibnitz Zentrum (MLZ) is equipped with a moveable assembly of two uranium plates, which can be placed in front of the entrance window of the beam tube via remote control. With these plates placed in their operating position the thermal neutron spectrum produced by the neutron source FRM II is converted to fission neutrons with 1.9 MeV of mean energy. This fission neutron spectrum is routinely used for medical applications at the irradiation facility MEDAPP, for neutron radiography and tomography experiments at the facility NECTAR and for materials testing. If, however, the uranium plates are in their stand-by position far off the tip of the beam tube and the so-called permanent filter for thermal neutrons is removed, thermal neutrons originating from the moderator tank enter the beam tube and a thermal spectrum becomes available for irradiation or activation of samples. By installing a temporary flight tube the beam may be used for thermal neutron radiography and tomography experiments at NECTAR. The thermal neutron beam option not only adds a pure thermal neutron spectrum to the energy ranges available for neutron imaging at MLZ instruments but it also is an unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages. The thermal neutron beam option for NECTAR is funded by BMBF in frame of research project 05K16VK3.

Medical set-up of boron neutron capture therapy (BNCT) for malignant glioma at the Japan research reactor (JRR)-4

International Nuclear Information System (INIS)

Yamamoto, T.; Matsumura, A.; Nose, T.; Shibata, Y.; Nakai, K.; Sakurai, F.; Kishi, T.; Kumada, H.; Yamamoto, K.; Torii, Y.

2001-01-01

The University of Tsukuba project for boron neutron capture therapy (BNCT) was initiated at the Japan Atomic Energy Research Institute (JAERI) in 1992. The clinical study for BNCT began at the Japan Research Reactor (JRR)-2 of the JAERI in November 1995. By the end of 1998, a new medical irradiation facility had been installed in JRR-4 of that included a new medical treatment room and patient-monitoring area adjacent to the irradiation room. The medical treatment room was built to reflect a hospital-type operation room that includes an operating table with a carbon head frame, anesthesia apparatus with several cardiopulmonary monitors, etc. Following craniotomy in the treatment room, a patient under anesthesia is transported into the irradiation room for BNCT. The boron concentration in tissue is measured with prompt gamma ray analysis (PGA) and simultaneously by inductively coupled plasma atomic emission spectroscopy (ICP-AES) methods. For the immediate pre- and post-BNCT care, a collaborating neurosurgical department of the University of Tsukuba was prepared in the vicinity of the JAERI. The long term follow-up is done at the University of Tsukuba Hospital. Epithermal neutron beam also became available at the new JRR-4. By changing the thickness and/or the configuration of heavy water, a cadmium plate, and a graphite reflector, the JRR-4 provides a variety of neutron beams, including three typical beams (Epithermal mode and Thermal modes I and II). Intraoperative BNCT using the thermal beam is planned to study at the beginning of the clinical trial. The ongoing development of the JAERI Computational Dosimetry System (JCDS) and radiobiological studies have focused in the application of the epithermal beam for BNCT. After obtaining these basic data, we are planning to use the epithermal beam for intraoperative BNCT. (author)

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.)

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

International Nuclear Information System (INIS)

1968-01-01

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

Distribution of 35 Elements in Peat Cores from Ombrotrophic Bogs Studied by Epithermal Neutron Activation Analysis

CERN Document Server

Frontasyeva, M V

2004-01-01

In ombrotrophic bogs the surface peat layer is supplied with chemical substances only from the atmosphere. Peat cores from these bogs therefore can be used to study temporal trends in atmospheric deposition of pollutants. In this work epithermal neutron activation analysis was applied for the first time to study the distribution of 35 elements in peat profiles from ombrotrophic bogs. The selected examples were from Finnmark county in northern Norway: one pristine site far from any local pollution source, and another strongly affected by long-term operation of Russian copper-nickel smelters located close to the border. The elements are classified with respect to their behavior in the uppermost 40 cm of the peat, and similarities and differences between the two profiles are discussed. As compared with other more commonly used analytical techniques based on acid decomposition of the sample ENAA has the advantage of providing the total concentrations of the elements.

Instrumentation at pulsed neutron sources

International Nuclear Information System (INIS)

Carpenter, J.M.; Lander, G.H.; Windsor, C.G.

1984-01-01

Scientific investigations involving the use of neutron beams have been centered at reactor sources for the last 35 years. Recently, there has been considerable interest in using the neutrons produced by accelerator driven (pulsed) sources. Such installations are in operation in England, Japan, and the United States. In this article a brief survey is given of how the neutron beams are produced and how they can be optimized for neutron scattering experiments. A detailed description is then given of the various types of instruments that have been, or are planned, at pulsed sources. Numerous examples of the scientific results that are emerging are given. An attempt is made throughout the article to compare the scientific opportunities at pulsed sources with the proven performance of reactor installations, and some familiarity with the latter and the general field of neutron scattering is assumed. New areas are being opened up by pulsed sources, particularly with the intense epithermal neutron beams, which promise to be several orders of magnitude more intense than can be obtained from a thermal reactor

Neutron dosimeter utilizing CR-39

International Nuclear Information System (INIS)

Souza, H.V.C. de.

1991-05-01

A personnel neutron dosimeter has been developed with discretization in a wide range of energies of real interest, utilizing the CR-39 polymer, to detect recoil protons in the fast range, and alpha particles in the thermal and epithermal ranges, with possibility to be disposed in the IRD/CNEN's conventional film badge suport. They are presented, abstractly, the difficulties and importance of the neutron dosimetry, beyond the general objectives that motivated this work execution. The details of the materials utilized in the dosimeter confection, and the experimental methodology employed to obtain the performance curves are presented. The results about linearity response of the dosimeter with respect to equivalent dose, in a wide range of doses, and about the verified angular dependence are analysed. (author)

High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen; Larry R. Greenwood; James R. Parry

2014-06-22

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

Pulsed neutron uranium borehole logging with prompt fission neutrons

International Nuclear Information System (INIS)

Bivens, H.M.; Smith, G.W.; Jensen, D.H.

1976-01-01

The gross count natural gamma log normally used for uranium borehole logging is seriously affected by disequilibrium. Methods for the direct measurement of uranium, such as neutron logging, which are not affected by disequilibrium have been the object of considerable effort in recent years. This paper describes a logging system for uranium which uses a small accelerator to generate pulses of 14 MeV neutrons to detect and assay uranium by the measurement of prompt fission neutrons in the epithermal energy range. After an initial feasibility study, a prototype logging probe was built for field evaluation which began in January 1976. Physical and operational characteristics of the prototype probe, the neutron tube-transformer assembly, and the neutron tube are described. In logging operations, only the epithermal prompt fission neutrons detected between 250 microseconds to 2500 microseconds following the excitation neutron pulse are counted. Comparison of corrected neutron logs with the conventional gross count natural gamma logs and the chemical assays of cores from boreholes are shown. The results obtained with this neutron probe clearly demonstrate its advantages over the gross count natural gamma log, although at this time the accuracy of the neutron log assay is not satisfactory under some conditions. The necessary correction factors for various borehole and formation parameters are being determined and, when applied, should improve the assay accuracy

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Epithermal neutron activation analysis of blue-green algae Spirulina Platensis as a matrix for selenium-containing pharmaceuticals

International Nuclear Information System (INIS)

Mosulishvili, L.M.; Kirkesali, E.I.; Belokobyl'skij, A.I.; Khizanishvili, A.I.; Frontas'eva, M.V.; Gundorina, S.F.; Oprea, C.D.

2000-01-01

To evaluate the potentiality of the blue-green algae Spirulina Platensis as a matrix for the production of Se-containing pharmaceuticals, the background levels of 31 major, minor and trace elements (Na, Mg, Al, Cl, K, Ca, Sc, V, Cr, Mn, Fe, Co, Ni (using (n,p)-reaction), As, Br, Zn, Rb, Mo, Ag, Sb, I, Ba, Sm, Tb, Tm, Hf, Ta, W, Au, Hg, Th) in Spirulina Platensis biomass were determined by means of epithermal neutron activation analysis. The possibility of the purpose-oriented incorporation of Se into Spirulina Platensis biomass was demonstrated. The polynomial dependence of the Se accumulation on nutritional medium loading was revealed. The employed analytical technique allows one to reliably control the amount of toxic elements in algae Spirulina Platensis. Based on this study, a conclusion of the possibility to use Spirulina Platensis as a matrix for the production of Se-containing pharmaceuticals was drawn

Proton linacs for boron neutron capture therapy

International Nuclear Information System (INIS)

Lennox, A.J.

1993-08-01

Recent advances in the ability to deliver boron-containing drugs to brain tumors have generated interest in ∼4 MeV linacs as sources of epithermal neutrons for radiation therapy. In addition, fast neutron therapy facilities have been studying methods to moderate their beams to take advantage of the high cross section for epithermal neutrons on boron-10. This paper describes the technical issues involved in each approach and presents the motivation for undertaking such studies using the Fermilab linac. the problems which must be solved before therapy can begin are outlined. Status of preparatory work and results of preliminary measurements are presented

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.)

A PC-program for the calculation of neutron flux and element contents using the ki-method of neutron activation analysis

International Nuclear Information System (INIS)

Boulyga, E.G.; Boulyga, S.F.

2000-01-01

A computer program is described, which calculates the induced activities of isotopes after irradiation in a known neutron field, thermal and epithermal neutron fluxes from the measured induced activities and from nuclear data of 2-4 monitor nuclides as well as the element concentrations in samples irradiated together with the monitors. The program was developed for operation in Windows 3.1 (or higher). The application of the program for neutron activation analysis allows to simplify the experimental procedure and to reduce the time. The program was tested by measuring different types of standard reference materials at the FRJ-2 (Research Centre, Juelich, Germany) and Triga (University Mainz, Germany) reactors. Comparison of neutron flux parameters calculated by this program with those calculated by a VAX program developed at the Research Centre, Juelich was done. The results of testing seem to be satisfactory. (author)

The final power calibration of the IPEN/MB-01 nuclear reactor for various configurations obtained from the measurements of the absolute average neutron flux

Energy Technology Data Exchange (ETDEWEB)

Silva, Alexandre Fonseca Povoa da, E-mail: alexandre.povoa@mar.mil.br [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto Credidio; Lima, Ana Cecilia de Souza; Betti, Flavio; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The use of neutron activation foils is a widely spread technique applied to obtain nuclear parameters then comparing the results with those calculated using specific methodologies and available nuclear data. By irradiation of activation foils and subsequent measurement of its induced activity, it is possible to determine the neutron flux at the position of irradiation. The power level during operation of the reactor is a parameter which is directly proportional to the average neutron flux throughout the core. The objective of this work is to gather data from irradiation of gold foils symmetrically placed along a cylindrically configured core which presents only a small excess reactivity in order to derive the power generated throughout the spatial thermal and epithermal neutron flux distribution over the core of the IPEN/MB-01 Nuclear Reactor, eventually lending to a proper calibration of its nuclear channels. The foils are fixed in a Lucite plate then irradiated with and without cadmium sheaths so as to obtain the absolute thermal and epithermal neutron flux. The correlation between the average power neutron flux resulting from the gold foils irradiation, and the average power digitally indicated by the nuclear channel number 6, allows for the calibration of the nuclear channels of the reactor. The reactor power level obtained by thermal neutron flux mapping was (74.65 ± 2.45) watts to a mean counting per seconds of 37881 cps to nuclear channel number 10 a pulse detector, and 0.719.10{sup -5} ampere to nuclear linear channel number 6 (a non-compensated ionization chamber). (author)

Review of the neutron capture process in fission reactors

International Nuclear Information System (INIS)

Poenitz, W.P.

1981-07-01

The importance of the neutron capture process and the status of the more important cross section data are reviewed. The capture in fertile and fissile nuclei is considered. For thermal reactors the thermal to epithermal capture ratio for 238 U and 232 Th remains a problem though some improvements were made with more recent measurements. The capture cross section of 238 U in the fast energy range remains quite uncertain and a long standing discrepancy for the calculated versus experimental central reaction rate ratio C28/F49 persists. Capture in structural materials, fission product nuclei and the higher actinides is also considered

Notes on neutron flux measurement; Notas sobre medida de flujos neutronicos

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1984-07-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs.

Irradiation system for neutron capture therapy using the small accelerator

International Nuclear Information System (INIS)

Kobayashi, Tooru; Hoshi, Masaharu

2002-01-01

Neutron capture therapy (NCT) is to kill tumor cells that previously incorporated the stable isotope which generates heavy charged particles with a short range and a high linear energy transfer (LET) on neutron irradiation. Boron-10 is ordinarily used as such an isotope. The tumor tissue is neutron-irradiated at craniotomy after preceding craniotomy for tumor extraction: therefore two surgeries are required for the present NCT in Japan. The reactions 10 B(n, αγ) 7 Li and 7 Li (p, n) 7 Be are thought preferential for patients and doctors if a convenient small accelerator, not the reactor used at present, is available in the hospital because only one craniotomy is sufficient. Authors' examinations of the system for NCT using the small accelerator involve irradiation conditions, desirable energy spectrum of neutron, characterization of thermal and epi-thermal neutrons, social, practical and technical comparison of the reactor and accelerator, and usefulness of the reaction 7 Li (p, n) 7 Be. The system devoted to the NCT is awaited in future. (K.H.)

Attenuation of thermal neutron through graphite

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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.

Neutron absorbed dose in a pacemaker CMOS

International Nuclear Information System (INIS)

Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R.; Paredes G, L.

2012-01-01

The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10 -17 Gy per neutron emitted by the source. (Author)

Neutron absorbed dose in a pacemaker CMOS

Energy Technology Data Exchange (ETDEWEB)

Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: fermineutron@yahoo.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2012-06-15

The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10{sup -17} Gy per neutron emitted by the source. (Author)

Beam neutron energy optimization for boron neutron capture therapy using monte Carlo method

International Nuclear Information System (INIS)

Pazirandeh, A.; Shekarian, E.

2006-01-01

In last two decades the optimal neutron energy for the treatment of deep seated tumors in boron neutron capture therapy in view of neutron physics and chemical compounds of boron carrier has been under thorough study. Although neutron absorption cross section of boron is high (3836b), the treatment of deep seated tumors such as glioblastoma multiform requires beam of neutrons of higher energy that can penetrate deeply into the brain and thermalized in the proximity of the tumor. Dosage from recoil proton associated with fast neutrons however poses some constraints on maximum neutron energy that can be used in the treatment. For this reason neutrons in the epithermal energy range of 10eV-10keV are generally to be the most appropriate. The simulation carried out by Monte Carlo methods using MCBNCT and MCNP4C codes along with the cross section library in 290 groups extracted from ENDF/B6 main library. The ptimal neutron energy for deep seated tumors depends on the sue and depth of tumor. Our estimated optimized energy for the tumor of 5cm wide and 1-2cm thick stands at 5cm depth is in the range of 3-5keV

Thermal neutron diffusion parameters in homogeneous mixtures

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

Thermal neutron diffusion parameters in homogeneous mixtures

International Nuclear Information System (INIS)

Drozdowicz, K.; Krynicka, E.

1995-01-01

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

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

Science.gov (United States)

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

2002-01-01

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

Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution; Caracterizacao do nucleo cilindrico de menor excesso de reatividade do reator IPEN/MB-01, pela medida da distribuicao espacial e energetica do fluxo de neutrons

Energy Technology Data Exchange (ETDEWEB)

Aredes, Vitor Ottoni Garcia

2014-07-01

In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes; Mesures des sections efficaces de capture et potentiels d'incineration des actinides mineurs dans les hauts flux de neutrons: Impact sur la transmutation des dechets

Energy Technology Data Exchange (ETDEWEB)

Bringer, O

2007-10-15

This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of {sup 241}Am and {sup 237}Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the {sup 241}Am isomeric branching ratio were precisely measured at thermal energy point. (author)

Determination of silicon in biological and botanical reference materials by epithermal INAA and Compton suppression

International Nuclear Information System (INIS)

Landsberger, S.; Peshev, S.; Becker, D.A.

1994-01-01

Silicon determination in sixteen botanical and biological standard reference materials is described using the 29 Si(n, p) 29 Al reaction through instrumental epithermal neutron activation analysis and Compton suppression gamma-ray spectroscopy. By simultaneous utilization of both cadmium and boron epithermal filters along with anticoincidence gamma-counting, detection limits as low as 12 ppm were obtained for certain matrices, much lower than previously reported values for this type of analysis. The method is applicable to many botanical and biological matrices and is attractive with its interference free, purely instrumental nature, compared with methods using the 28 Si(n, p) 28 Al reaction or chemical separation techniques. ((orig.))

Effects of secondary interactions on the dose calculation in treatments with Boron Neutron Capture Therapy (BNCT)

International Nuclear Information System (INIS)

Monteiro, E.

2004-01-01

The aimed of this work consists of evaluating the influence of the secondary contributions of dose (thermal neutrons dose, epithermal neutrons dose, fast neutrons dose and photon dose) in treatment planning with BNCT. MCNP4B Code was used to calculate RBE-Gy doses through the irradiation of the modified Snyder head head phantom.A reduction of the therapeutical gain of monoenergetic neutron beans was observed in non invasive treatments, provoked for the predominance of the fast neutron dose component in the skin, showing that the secondary contributions of dose can contribute more in the direction to raise the dose in the fabric healthy that in the tumor, thus reducing the treatment efficiency. (author)

Evaluation of neutron irradiation fields for BNCT by using absorbed dose in a phantom

International Nuclear Information System (INIS)

Aizawa, O.

1993-01-01

In a previous paper, the author defined the open-quotes irradiation timeclose quotes as the time of irradiation in which the maximum open-quotes total background doseclose quotes becomes 2,500 RBE-cGy. In this paper, he has modified the definition a little as the time of irradiation in which the maximum open-quotes lμg/g B-10 doseclose quotes becomes 3,000 RBE-cGy, because he assumed that normal tissue contained 1μg/g B-10. Moreover, he has modified the dose criteria for BNCT as follows: The open-quotes eye doseclose quotes, open-quotes total body doseclose quotes and open-quotes except-head doseclose quotes should be less that 200, 100 and 50 RBE-cGy, respectively. He has added one more criterion for BNCT that the thermal neutron fluence at the tumor position should be over 2.5x10 12 n/cm 2 at the open-quotes irradiation timeclose quotes. The distance from the core side to the irradiation port in the open-quotes old configurationclose quotes of the Musashi reactor (TRIGA-II, 100kW) was 160 cm. He is now planning to design an eccentric core and to move the reactor core nearer to the irradiation port, distance between the core side and the irradiation port to be 140, 130 and 120cm. The other assumptions used in this paper are as follows: (1) The B-10 concentrations in tumor are 30 and/or 10μg/g. (2) The depth of the tumor is 5.0 cm to 5.5 cm from the surface. (3) The RBE values used are 1.0 for all gamma rays and 2.3 for B 10 (n,α) reaction products. (4) The RBE values for neutrons are the following three cases: the first case is using 1.6 for all neutrons; the second one is using 3.2 for non-thermal neutrons and 1.6 for thermal neutrons; the third case is using 4.8 for fast neutrons, 3.2 for faster epithermal and epithermal neutrons, and 1.6 for thermal neutrons

The k0-NAA Standardization Method Using an Am-Be Neutron Source

International Nuclear Information System (INIS)

Soliman, N.F.; Mohamed, G.Y.; Hassan, M.F.; Ali, M.A.

2012-01-01

Instrumental neutron activation analysis is a well established technique for the analysis of trace elements in different samples. Precise elemental concentrations of Al, Mn, Mg and Na in two unknown geological samples were determined by using the k 0 -standardization method. For such measurements two sets of standard monitors of Gold (Au), Indium (In), Tungsten (W) and Titanium (Ta) were used. One set is bare and the other is cadmium covered. These monitors were used for measuring the irradiation position factors f and α and using the cadmium ratios of the 115 In(n,γ) 116 In and 182 Ta(n,γ) 183 Ta interactions. Neutrons were obtained from CNIF 2 facility that uses an Am-Be radio-isotopic neutron source with a modification to have thermal and epi-thermal neutrons. Measurements were carried out using a gamma-ray spectrometer consisting of a hyper pure germanium detector and necessary associated electronics. The k 0 -standardization method can be used for quality control tests.

Hydrogen content in Zircaloy-IV by neutron emissions

International Nuclear Information System (INIS)

Granada, J.R.

1991-01-01

The feasibility of using thermal neutron transmission techniques to determine the initial hydrogen content in Zry-IV is discussed in this work. The usual conditions of very low H concentration (∼ 10-20 ppm in weight) in the alloy, require a high degree of accuracy in the relevant cross sections in order to obtain reliable values of that concentration from the measured transmission spectra. Total cross sections corresponding to the metallic alloy and to hydrogen in that matrix were evaluated using interaction models developed in this Laboratory, for neutrons covering the range from subthermal up to epithermal energies. The sensitivity of this method as a function of neutron energy is discussed, and an optimized and self consistent procedure for data analysis after a transmission experiment is proposed. (Author) [es

Halogen determination in Arctic aerosols by neutron activation analysis with Compton suppression methods

International Nuclear Information System (INIS)

Landsberger, S.; Basunia, M.S.; Iskander, F.

2001-01-01

The study of halogens particularly bromine and chlorine in Arctic aerosols has received a great deal of attention in the past decade in ozone depletion during polar sunrise studies. Iodine has also been studied as part of geochemical cycling. It was shown that all three of the above elements can be determined simultaneously with very low detection limits using epithermal NAA in conjunction with Compton suppression methods. Besides lowering the background considerably, Compton suppression can eliminate or minimize the overlapping peak of the 620 keV photopeak arising form the 1642 keV double escape peak of 38 Cl interfering with the 616.9 keV photopeak of 79 Br(n,γ) 80 Br reaction. Iodine is ideally determined by epithermal NAA because of its very good resonance integral cross-section. Although chlorine is usually determined using thermal neutrons via the 37 Cl(n,γ) 38 Cl reactions, epithermal NAA is still feasible for the Arctic aerosol, since it has a major sea-salt component. (author)

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

Desain Beam Shaping Assembly (BSA berbasis D-D Neutron Generator 2,45 MeV untuk Uji Fasilitas BNCT

Directory of Open Access Journals (Sweden)

Desman P. Gulo

2015-12-01

Full Text Available Boron Neutron Capture Therapy (BNCT is one of the cancer treatments that are being developed in nowadays. In order to support BNCT treatment for cancer that exists in underneath skin like breast cancer, the facility needs a generator that is able to produce epithermal neutron. One of the generator that is able to produce neutron is D-D neutron generator with 2.45 MeV energy. Based on the calculation of this paper, we found that the total production of neutron per second (neutron yield from Neutron Generator (NG by PSTA-BATAN Yogyakarta is 2.55×1011 n/s. The energy and flux that we found is in the range of quick neutron. Thus, it needs to be moderated to the level of epithermal neutron which is located in the interval energy of 1 eV to 10 KeV with 109 n/cm2s flux. This number is the recommendation standard from IAEA. Beam Shaping Assembly (BSA is needed in order to moderate the quick neutron to the level of epithermal neutron. One part of BSA that has the responsibility in moderating the quick neutron to epithermal neutron is the moderator. The substance of moderator used in this paper is MgF2 and A1F3. The thickness of moderator has been set in in such a way by using MCNPX software in order to fulfill the standard of IAEA. As the result of optimizing BSA moderator, the data obtain epithermal flux with the total number of 4.64×108 n/cm2/s for both of moderators with the thickness of moderator up to 15 cm. At the end of this research, the number of epithermal flux does not follow the standard of IAEA. This is because the flux neutron that is being produced by NG is relatively small. In conclusion, the NG from PSTA-BATAN Yogyakarta is not ready to be used for the BNCT treatment facility for the underneath skin cancer like breast cancer.

Alanine and TLD coupled detectors for fast neutron dose measurements in neutron capture therapy (NCT)

Energy Technology Data Exchange (ETDEWEB)

Cecilia, A.; Baccaro, S.; Cemmi, A. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Colli, V.; Gambarini, G. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy); Rosi, G. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Scolari, L. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy)

2004-07-01

A method was investigated to measure gamma and fast neutron doses in phantoms exposed to an epithermal neutron beam designed for neutron capture therapy (NCT). The gamma dose component was measured by TLD-300 [CaF{sub 2}:Tm] and the fast neutron dose, mainly due to elastic scattering with hydrogen nuclei, was measured by alanine dosemeters [CH{sub 3}CH(NH{sub 2})COOH]. The gamma and fast neutron doses deposited in alanine dosemeters are very near to those released in tissue, because of the alanine tissue equivalence. Couples of TLD-300 and alanine dosemeters were irradiated in phantoms positioned in the epithermal column of the Tapiro reactor (ENEA-Casaccia RC). The dosemeter response depends on the linear energy transfer (LET) of radiation, hence the precision and reliability of the fast neutron dose values obtained with the proposed method have been investigated. Results showed that the combination of alanine and TLD detectors is a promising method to separate gamma dose and fast neutron dose in NCT. (authors)

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

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Toru

1999-01-01

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

A target-moderator-reflector concept of the JAERI 5 MW pulsed spallation neutron source

International Nuclear Information System (INIS)

Watanabe, Noboru; Teshigawara, Makoto; Aizawa, Kazuya; Suzuki, Jyunichi; Oyama, Yukio

1998-03-01

In Japan Atomic Energy Research Institute the construction of a 5 MW (short) pulsed spallation neutron source is under planning using a projected high power superconducting proton (or H - ) linac of 8 MW in total beam power. In the present paper we report our consideration on target-moderator-reflector concept, based on the layout of the tentative neutron instruments for the assumed neutron scattering experiments in future. The choice of cold neutron moderators for high resolution and high intensity experiments, thermal and epithermal neutron moderators for high resolution uses was discussed and a reference layout of target-moderator-reflector system was proposed for detailed neutronic calculation and optimization. The proposed system was designed like that it can provide, at least, 30 beam lines for more than 40 instruments. (author)

Conceptual design of a clinical BNCT beam in an adjacent dry cell of the Jozef Stefan Institute TRIGA reactor

NARCIS (Netherlands)

Maucec, M

The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation