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Sample records for thermal neutron measurements

  1. Measurement of thermal neutron capture cross section

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  2. Method and apparatus for measuring thermal neutron characteristics

    International Nuclear Information System (INIS)

    Jacobson, L.A.; Johnstone, C.M.

    1980-01-01

    A system of measuring the thermal neutron decay characteristics of an earth formation is described. The method consists of detecting indications of the thermal neutron concentration in the formation during a selected set of two measurement intervals following irradiation of the formation with a burst of fast neutrons. (U.K.)

  3. The measurements of thermal neutron flux distribution in a paraffin

    Indian Academy of Sciences (India)

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

  4. The measurements of thermal neutron flux distribution in a paraffin ...

    Indian Academy of Sciences (India)

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

  5. The measurements of thermal neutron flux distribution in a paraffin ...

    Indian Academy of Sciences (India)

    at 293.4 K. In order to measure the thermal neutron flux density, the foil activation method was used. Thermal neutron flux determination in paraffin phantom by counting the emitted rays of indium foils with two different detectors (Geiger–. Muller counter and NaI(Tl)) was the aim of this project. The relative differences of the ...

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

    International Nuclear Information System (INIS)

    Egbert, Stephen D.

    2005-01-01

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

  7. 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

  8. Measurement of the diffusion length of thermal neutrons inside graphite

    International Nuclear Information System (INIS)

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

    1948-11-01

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

  9. Measurements of angular and energy distributions of prompt neutrons from thermal neutron-induced fission

    Science.gov (United States)

    Vorobyev, A. S.; Shcherbakov, O. A.; Pleva, Yu. S.; Gagarski, A. M.; Val'ski, G. V.; Petrov, G. A.; Petrova, V. I.; Zavarukhina, T. A.

    2009-01-01

    The experimental setup and methodology used to measure prompt neutron angular and energy distributions from thermal neutron-induced fission are described. The neutrons are detected using two scintillation detectors, while the fission fragments are detected by multi-wire proportional detectors in conjunction with the TOF technique. To separate events corresponding to neutrons and γ-quanta, a double discrimination by the pulse shape and the time-of-flight is applied. Some preliminary results of an experiment performed with the 235U target are presented and briefly discussed. The yield of "scission" neutrons has been estimated in the framework of a simple evaporation model and was found not to exceed 5% of the total neutron yield. Including an assumed of anisotropy of the fission neutron angular distribution in the center-of-mass system of fission fragments into the model calculation leads to an increase in the "scission" neutron yields inferred from the data.

  10. The measurements of thermal neutron flux distribution in a paraffin ...

    Indian Academy of Sciences (India)

    neutron flux density, the foil activation method was used. Thermal neutron flux determination in paraffin phantom by counting the emitted rays of indium foils with two different detectors (Geiger–. Muller counter and NaI(Tl)) was the aim of this project. The relative differences of the outcome of the experiments were between ...

  11. Thermal neutron cross section measurements for technetium-99

    International Nuclear Information System (INIS)

    Yates, M.A.; Schroeder, N.C.; Fowler, M.M.

    1993-01-01

    Technetium, because of its long half-like (213,000 years) and ability to migrate in the environment, is a primary contributor to the long-term radioactivity related risk associated with geologic nuclear waste disposal. One proposal for converting technetium to an environmentally benign element investigating transmutation with an accelerator-based system, (i.e., Accelerator Transmutation of Waste, ATW). Planning for efficient processing of technetium through the transmuter will require knowledge of the thermal neutron cross section for the 99 Tc (n,γ) 100 Tc reaction. The authors have recently remeasured this cross section. Weighed aliquots (19-205 μg) of a NIST traceable 99 Tc standard were irradiated for 30-150 sec using the pneumatic open-quotes rabbitclose quotes system of LANL's Omega West Reactor. The two gamma rays from the 15.7-sec half-life product were measured immediately after irradiation on a high-resolution Ge detector. Thermal fluxes were measured using gold foils and Cd wrapped gold foils. The observation cross section is 19 ± 1 b. This agrees well with the 1977 value but has half the uncertainty

  12. Thermal and epithermal neutron flux distributions measurement in thermal column of TRR using an experimental-simulation method.

    Science.gov (United States)

    Adeli, Ruhollah; Kasesaz, Yaser; Shirmardi, Seyed Pezhman; Ezaty, Arsalan

    2018-03-01

    For designing an appropriate neutron beam, the determination of neutron flux at any irradiation facility is an important key factor. Due to the importance of determining the thermal and epithermal neutron fluxes in a typical thermal column of a reactor, a simple and accurate technique is introduced in this study. Absolute thermal and epithermal fluxes were measured experimentally at a certain point using the foil activation method by neutron bombardment of bare and cadmium covered Au foils. The relative neutron fluxes were also derived simply by means of Monte Carlo simulation by accurate modelling of the reactor components. Finally, by normalization of the relative distribution flux with regard to information about the absolute neutron flux, the accurate thermal and epithermal neutron distributions were derived, separately. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Thermal neutron absorption borehole logging

    International Nuclear Information System (INIS)

    Flaum, C.

    1982-01-01

    A method is described of quantitatively determining the macroscopic thermal neutron cross-section of a geological formation traversed by a borehole by measuring the flux of both thermal and epithermal neutrons following the irradiation of the formation with neutrons from a continuous source in a neutron sonde. (U.K.)

  14. Measurements of thermal and fast neutron fluxes at the TRIGA reactor

    International Nuclear Information System (INIS)

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

    1966-01-01

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

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

    Science.gov (United States)

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

    2018-01-01

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

  16. The concept of measurement of thermal neutron absorption cross section in small samples

    International Nuclear Information System (INIS)

    Czubek, J.A.

    1980-01-01

    Theoretical principles of the method of measurement of the absorption cross section for thermal neutrons are presented in the one velocity approach. In consecutive measurements the sample investigated is enveloped in shells of a known moderator of varying thickness and irradiated with the pulsed beam of fast neutrons. The die-away rate of thermal neutrons escaping from such a system is measured. The absorption cross section of the unknown sample is found as the intersection of the experimental curve (die-away rate viz. thickness of the moderator) with the theoretical one calculated for the case of the zero value of the material buckling of the sample. (author)

  17. Apparatus and process for continuous measurement of moisture in moving coal by neutron thermalization

    International Nuclear Information System (INIS)

    Stewart, R.F.

    1967-01-01

    The invention relates to an apparatus and process for the measurement of moisture contents in solid materials. More particularly, the invention makes available a continuous moisture analysis of a moving mass of material, such as coal, by penetrating such material with neutrons emitted from a source of fast neutrons and detecting, counting, and recording slowed or thermalized neutrons reflected from the internal structure of the material. (U.S.)

  18. Neutron measurement at the thermal column of the Malaysian Triga Mark II reactor using gold foil activation method and TLD

    Science.gov (United States)

    Shalbi, Safwan; Salleh, Wan Norhayati Wan; Mohamad Idris, Faridah; Aliff Ashraff Rosdi, Muhammad; Syahir Sarkawi, Muhammad; Liyana Jamsari, Nur; Nasir, Nur Aishah Mohd

    2018-01-01

    In order to design facilities for boron neutron capture therapy (BNCT), the neutron measurement must be considered to obtain the optimal design of BNCT facility such as collimator and shielding. The previous feasibility study showed that the thermal column could generate higher thermal neutrons yield for BNCT application at the TRIGA MARK II reactor. Currently, the facility for BNCT are planned to be developed at thermal column. Thus, the main objective was focused on the thermal neutron and epithermal neutron flux measurement at the thermal column. In this measurement, pure gold and cadmium were used as a filter to obtain the thermal and epithermal neutron fluxes from inside and outside of the thermal column door of the 200kW reactor power using a gold foil activation method. The results were compared with neutron fluxes using TLD 600 and TLD 700. The outcome of this work will become the benchmark for the design of BNCT collimator and the shielding

  19. Measurement of the diffusion length of thermal neutrons in the beryllium oxide

    International Nuclear Information System (INIS)

    Koechlin, J.C.; Martelly, J.; Duggal, V.P.

    1955-01-01

    The diffusion length of thermal neutrons in the beryllium oxide has been obtained while studying the spatial distribution of the neutrons in a massive parallelepiped of this matter placed before the thermal column of the reactor core of Saclay. The mean density of the beryllium oxide (BeO) is 2,95 gr/cm 3 , the mean density of the massif is 2,92 gr/cm 3 . The value of the diffusion length, deducted of the done measures, is: L = 32,7 ± 0,5 cm (likely gap). Some remarks are formulated about the influence of the spectral distribution of the neutrons flux used. (authors) [fr

  20. Status of measured neutron cross sections of transactinium isotopes for thermal reactors

    International Nuclear Information System (INIS)

    Benjamin, R.W.

    1976-01-01

    Experimentally determined neutron cross sections, resonance parameters, and the average number of neutrons per fission for neutron-induced fission of actinide nuclides in the production chains associated with thermal and near-thermal reactors are summarized and compared with user requests for experimental data. The primary fertile and fissile isotopes 232 Th, 233 U, 235 U, 238 U, and 239 Pu are excluded from this survey. Integral data, i.e., spectrum-averaged thermal cross sections and resonance integrals, are included, but the emphasis is placed on energy-dependent differential cross sections because of their general utility with any specified neutron energy spectrum. Included with the data summaries are an extensive survey of the literature through August 1975, brief descriptions of measurements known to be in progress or firmly planned for the immediate future, and recommendations for needed measurements. (author)

  1. Status of measured neutron cross sections of transactinium isotopes for thermal reactors

    International Nuclear Information System (INIS)

    Benjamin, R.W.

    1975-01-01

    Experimentally determined neutron cross sections, resonance parameters, and the average number of neutrons per fission for neutron-induced fission of actinide nuclides in the production chains associated with thermal and near-thermal reactors are summarized and compared with user requests for experimental data. The primary fertile and fissile isotopes 232 Th, 233 U, 235 U, 238 U, and 239 Pu are excluded from this survey. Integral data, i.e., spectrum-averaged thermal cross sections and resonance integrals, are included, but the emphasis is placed on energy-dependent differential cross sections because of their general utility with any specified neutron energy spectrum. Included with the data summaries are an extensive survey of the literature through August 1975, brief descriptions of measurements known to be in progress or firmly planned for the immediate future, and recommendations for needed measurements. (3 figures, 5 tables)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  3. Measured Thermal and Fast Neutron Fluence Rates for ATF-1 Holders During ATR Cycle 157D

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Larry Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, David Torbet [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 157D which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains measurements of the fluence rates corresponding to the particular elevations relative to the 80-ft. core elevation. The data in this report consist of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, and (3) plots of both the thermal and fast neutron fluence rates. The fluence rates reported are for the average power levels given in the table of power history and distribution.

  4. Methodology of measurement of thermal neutron time decay constant in Canberra 35+ MCA system

    International Nuclear Information System (INIS)

    Drozdowicz, K.; Gabanska, B.; Igielski, A.; Krynicka, E.; Woznicka, U.

    1993-01-01

    A method of the thermal neutron time decay constant measurement in small bounded media is presented. A 14 MeV pulsed neutron generator is the neutron source. The system of recording of a die-away curve of thermal neutrons consists of a 3 He detector and of a multichannel time analyzer based on analyzer Canberra 35+ with multi scaler module MCS 7880 (microsecond range). Optimum parameters for the measuring system are considered. Experimental verification of a dead time of the instrumentation system is made and a count-loss correction is incorporated into the data treatment. An attention is paid to evaluate with a high accuracy the fundamental mode decay constant of the registered decaying curve. A new procedure of the determination of the decay constant by a multiple recording of the die-away curve is presented and results of test measurements are shown. (author). 11 refs, 12 figs, 4 tabs

  5. Simultaneous measurement of fission fragments and prompt neutrons for thermal neutron-induced fission of U-235

    Energy Technology Data Exchange (ETDEWEB)

    Nishio, Katsuhisa; Yamamoto, Hideki; Kimura, Itsuro; Nakagome, Yoshihiro [Kyoto Univ. (Japan)

    1997-03-01

    Simultaneous measurement of fission fragments and prompt neutrons following the thermal neutron induced fission of U-235 has been performed in order to obtain the neutron multiplicity (v) and its emission energy ({eta}) against the specified mass (m{sup *}) and the total kinetic energy (TKE). The obtained value of -dv/dTKE(m{sup *}) showed a saw-tooth distribution. The average neutron energy <{eta}>(m{sup *}) had a distribution with a reflection symmetry around the half mass division. The measurement also gave the level density parameters of the specified fragment, a(m{sup *}), and this parameters showed a saw-tooth trend too. The analysis by a phenomenological description of this parameters including the shell and collective effects suggested the existence of a collective motion of the fission fragments. (author)

  6. The dynamic method for time-of-flight measurement of thermal neutron spectra from pulsed sources

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Chuklyaev, S.V.; Tulaev, A.B.; Bobrakov, V.F.

    1995-01-01

    A time-of-flight method for measurement of thermal neutron spectra in pulsed neutron sources with an efficiency more than 10 5 times higher than the standard method is described. The main problems associated with the electric current technique for time-of-flight spectra measurement are examined. The methodical errors, problems of special neutron detector design and other questions are discussed. Some experimental results for spectra from the surfaces of water and solid methane moderators obtained at the IBR-2 pulsed reactor (Dubna, Russia) are presented. (orig.)

  7. Thermal neutron spectra measurements in IEAR-1 Reactor, by using a crystal spectrometer

    International Nuclear Information System (INIS)

    Fulfaro, R.; Figueiredo Neto, A.M.; Stasiulevicius, E.; Vinhas, L.A.

    1975-01-01

    The thermal neutron spectrum of the IEN Argonauta reactor has been measured in the wavelength from 0.7 to 1.9A, using a neutron crystal spectrometer. An aluminium single crystal, in transmission, was used as monochromator. The aluminium crystal reflectivity employed in the analysis of the data was calculated for the first five permitted orders. An effective absorption coefficient of the crystal was used to perform the calculations instead of the macroscopic cross section of the element

  8. Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 160A

    Energy Technology Data Exchange (ETDEWEB)

    Walker, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, D. T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-06-06

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 160A which were measured by the Radiation Measurements Laboratory (RML).

  9. On line local measurement of thermal neutron flux on BNCT patient using SPND

    International Nuclear Information System (INIS)

    Miller, M.E.; Sztejnberg Goncalves-Carralves, M.L.; Gonzalez, S.J.

    2006-01-01

    The first on-line neutron flux measurement on a patient using a self-powered neutron detector (SPND) was assessed during the fourth clinical trial of the Boron Neutron Capture Therapy (BNCT) Project carried out at the National Atomic Energy Commission of Argentina (CNEA) and the medical center Angel H. Roffo. The SPND was specially developed and assembled for BNCT by CNEA. Its small size, 1 cm sensible length and 1.9 mm diameter, allowed performing a localized measurement. Since the treated tumors were cutaneous melanomas of nodular type, the SPND was located on the patient's skin. The patient was exposed to three different and consecutive fields and in each of them the SPND was used to measure local thermal neutron fluxes at selected dosimetric reference points. The values of the measured fluxes agreed with the ones estimated by calculation. This trial also demonstrated the usefulness of the SPND for assessing flux on-line. (author)

  10. Methods and apparatus for measuring thermal neutron decay characteristics of earth formations

    International Nuclear Information System (INIS)

    Johnstone, C. W.

    1981-01-01

    In the exemplary embodiments of the invention disclosed, thermal neutron decay characteristics of an earth formation, E.G. the thermal neutron decay time constant, tau, and the macroscopic capture cross section sigma, are measured by detecting indications of the thermal neutron concentration in the formation during a selected set of two different measurement intervals between neutron bursts and forming a ratio of such measurements, as accumulated for an appropriate length of time, from which the measured value of the decay characteristic is derived. The particular set of measurement intervals used is selected from among a number of possible sets as a function of a previously measured value of the decay characteristic. Each measurement interval set is used over only a specific range of decay characteristic values for which it has been determined, in accordance with a previously established relationship between the decay characteristic value and the ratio of the thermal neutron concentration measurements for the set, to afford enhanced statistical accuracy in the measured value of the decay characteristic. The measured value of the decay characteristic is then determined from the actual measured value of the ratio for the selected measurement interval set and the previously established relationship for that set. Each measurement interval preferably consists of a plurality of contiguous, discrete time gates whose durations and times of occurrence are adjustable in common by a finite number, E.G. Four, of discrete scale factor values F. Each F value corresponds to a specific range of decay characteristic values, with the F value used for any given measurement being selected on the basis of a previously measured value of the decay characteristic. Regions of overlap are provided between the decay characteristic value ranges for adjacent F values so that F need not be changed merely as a result of statistical variations in the decay characteristic

  11. 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.)

  12. Measuring thermal neutron spectra of RIEN-1 reactor with a chopper

    International Nuclear Information System (INIS)

    Jesus Vilar, G. de.

    1977-03-01

    The setting up of a time-of-flight spectrometer (Fermi Chopper) and its use in measurements of thermal neutron spectra in the irradiation channels of the Argonaut Reactor(Instituto de Engenharia Nuclear, Brazil), is described. These distributions are obtained using a multichannel analyser with the necessary corrections being made for counting losses in the analyser, dectector efficiency experimental resolution and chopper transmission function. The results obtained show that the thermal neutron flux emerging from the canal J-9 can be approximately described by a Maxwellian distribution with and associated characteristic temperature fo 430+-30 0 K [pt

  13. Measurements of thermal neutron capture cross sections for some FP nuclides

    International Nuclear Information System (INIS)

    Nakamura, Shoji; Furutaka, Kazuyoshi; Harada, Hideo; Katoh, Toshio

    1999-01-01

    The thermal neutron capture cross sections (σ 0 ) and the resonance integrals (I 0 ) of some FP elements, such as 80 Se, 94 Zr, 124 Sn, 127 I and 133 Cs, were measured by the activation and γ-ray spectroscopic method. (author)

  14. Measurement of spectrum for thermal neutrons produced from H2O moderator coupled with mercury target

    International Nuclear Information System (INIS)

    Meigo, S.; Maekawa, F.; Kasugai, Y.; Nakashima, H.; Ikeda, Y.; Watanabe, N.

    2001-01-01

    In order to obtain fundamental data for the design of pulsed spallation neutron source, the slowing-down and thermalized neutrons from an H 2 O moderator coupled with the mercury target were measured using GeV proton beams at AGS (Alternative Gradient Synchrotron) in BNL (Brookhaven National Laboratory) under the ASTE (AGS-Spallation Target Experiment) collaboration. The mercury target (φ 20 cm x L 130 cm) was surrounded by a lead reflector (1 x 1 x 1 m 3 ) was irradiated by 1.94-, 12- and 24-GeV protons. The spectral intensities of thermal neutrons from the moderator are measured by the current-mode time-of-flight technique using enriched 6 Li and 7 Li glass scintillators. By this technique, only several incident pulses were needed to obtain sufficient statistics for incident energy. The results have shown that the neutron spectral intensity per proton integrated over the Maxwellian region was almost proportional to the proton energy. By moving the target along the beam direction within 15 cm, the dependence of the relative moderator position to the target on the neutron flux was also measured. With this position change, the difference with flux was found within 10%. (author)

  15. Moisture disturbance when measuring boron content in wet glass fibre materials with thermal neutron transmission method

    International Nuclear Information System (INIS)

    Zhang Zhiping; Liu Shengkang; Zhang Yongjie

    2001-01-01

    The theoretical calculation and experimental study on the moisture disturbance in the boron content measurement of wet glass fibre materials using the thermal neutron transmission method were reported. The relevant formula of the moisture disturbance was derived. For samples with a mass of 16 g, it was found that a moisture variation of 1% (mass percent) would result in a deviation of 0.28% (mass percent) in the measurement of boron contents

  16. Bibliography for thermal neutron scattering

    International Nuclear Information System (INIS)

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

    1976-12-01

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

  17. Measurements of the thermal neutron flux for an accelerator-based photoneutron source.

    Science.gov (United States)

    Taheri, Ali; Pazirandeh, Ali

    2016-12-01

    To have access to an appropriate neutron source is one of the most demanding requirements for neutron studies. This is important specially in laboratory and clinical applications, which need more compact and accessible sources. The most known neutron sources are fission reactors and natural isotopes, but there is an increasing interest for using accelerator based neutron sources because of their advantages. In this paper, we shall present a photo-neutron source prototype which is designed and fabricated to be used for different neutron researches including in-laboratory neutron activation analysis and neutron imaging, and also preliminary studies in boron neutron capture therapy (BNCT). Series of experimental tests were conducted to examine the intensity and quality of the neutron field produced by this source. Monte-Carlo simulations were also utilized to provide more detailed evaluation of the neutron spectrum, and determine the accuracy of the experiments. The experiments demonstrated a thermal neutron flux in the order of 10 7 (n/cm 2 .s), while simulations affirmed this flux and showed a neutron spectrum with a sharp peak at thermal energy region. According to the results, about 60 % of produced neutrons are in the range of thermal to epithermal neutrons.

  18. Measurement of very small hydrogen content in zirconium alloys by measuring thermal neutron incoherent scattering

    CERN Document Server

    Choi, Y N; Lee, C H; Oh, H S; Park, S D; Somenkov, V A

    2002-01-01

    In neutron-scattering experiments, the incoherent scattering contributes to the background signal, which is an unwelcome property of matter. Among natural nuclei, the hydrogen nucleus (proton) has a remarkably large value of incoherent neutron scattering cross section. Therefore, a very small amount of hydrogen in a material could be analyzed by measuring the neutron incoherent scattering of the material. The hydrogen content of a metal or semiconductor is a matter of concern because it can affect significantly the physical, mechanical or chemical properties of materials although the amount of hydrogen is very small. In this study, the neutron incoherent scattering was measured using a 1-D position-sensitive detector at 1.835 A. Estimated detection limits are about 5 and 2 mu g/g for 10-min and 1-h measurements, respectively. Using the calibration data obtained by measurement of artificial samples (zircaloy+polypropylene films), the relative amounts of hydrogen in three commercial zircaloy samples were estima...

  19. Measurement system of gamma, X, thermal and/or fast neutron flux

    International Nuclear Information System (INIS)

    Siffert, P.; Regal, R.; Koebel, J.M.; Teissier, C.

    1987-01-01

    The system includes detection means of gamma or/and X radiation, detection means of gamma and/or X radiation from thermal neutrons, detection means of gamma and/or X radiation from thermal and/or fast neutrons. It includes also processing devices of the signals given by the detection means able to get a linear combination of the detected signals. These processing devices give a signal selectively representative of photon, thermal or fast neutron rate [fr

  20. Measurement of thermal neutrons density by the cadmium filter method

    International Nuclear Information System (INIS)

    Manent, G.

    1967-01-01

    The cadmium correction factor is defined and measured by the Tittle method for the gold. The author shows that this factor is constant in function of the detector thickness and that the resonant part of the cross section is preponderant for its determination. This property is needful to the cadmium correction factor calculation if, as usual, the effective cut energy variation of the filter is not taken into account with the detector thickness and the cadmium thickness. (A.L.B.) [fr

  1. Flux depression and the absolute measurement of the thermal neutron flux density

    International Nuclear Information System (INIS)

    Bensch, Friedrich.

    1977-01-01

    The thermal neutron flux depression in a diffusing medium by an absorbing foil has been treated in numerous papers. The results are re-examined in an attempt to find a uniform and physically meaningful representation of the 'activation correction'. This quantity can be split up into a combination of probabilities. Thus, it is possible to determine the activation correction for any moderator and foil material. Measurements confirm the utility of the concepts introduced

  2. Measurement of {sup 238}Np fission cross-section by neutrons near thermal point (preliminary results)

    Energy Technology Data Exchange (ETDEWEB)

    Abramo; vich, S.N.; Andreev, M.F.; Bol`shakov, Y.M. [Institute of Experimental Physics, Arzamas (Russian Federation)] [and others

    1995-10-01

    Measurements have been carried out of {sup 238}Np fission cross-section by thermal neutrons. The isotope {sup 238}Np was built up through the reaction {sup 238}U(p,n) on an electrostatic accelerator. Extraction and cleaning of the sample were done by ion-exchange chromatography. Fast neutrons were generated on the electrostatic accelerator through the reaction {sup 9}Be(d,n); a polyethylene block was used to slow down neutrons. Registration of fission fragments was performed with dielectric track detectors. Suggesting that the behavior of {sup 238}Np and {sup 238}U. Westscott`s factors are indentical the fission cross-section of {sup 238}Np was obtained: {sigma}{sub fo}=2110 {plus_minus} 75 barn.

  3. In situ changes in the moisture content of heated, welded tuff based on thermal neutron measurements

    International Nuclear Information System (INIS)

    Ramirez, A.L.; Carlson, R.C.; Buscheck, T.A.

    1991-07-01

    Thermal neutron logs were collected to monitor changes in moisture content within a welded tuff rock mass heated from a borehole containing an electrical heater which remained energized for 195 days. Thermal neutron measurements were made in sampling boreholes before, during and after heating. The results generally corroborated our conceptual understanding of hydrothermal flow as well as most of the numerical modeling conducting for this study. Conceptual models have been developed in conjunction with the numerical model calculations to explain differences in the drying and re-wetting behavior above and below the heater. Numerical modeling indicated that the re-wetting of the dried-out zone was dominated by the binary diffusion of water vapor through fractures. Saturation gradients in the rock matrix resulted in relative humidity gradients which drove water vapor (primarily along fractures) back to the dried-out zone where it condensed along the fracture walls and was imbibed by the matrix. 4 refs., 28 figs

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

    Science.gov (United States)

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

    2018-01-01

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

  5. Measurements of thermal- and slow-neutron dose distributions in ordinary concrete shield using a reactor neutron beam of different energy ranges

    Energy Technology Data Exchange (ETDEWEB)

    Megahid, R.M.; Makarious, A.S.; El-Kolaly, M.A.; Afifi, Y.A.

    1980-01-01

    Experimental studies on the distribution and attenuation of thermal and slow neutron doses in ordinary concrete shield have been carried-out. A collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor was used. Measurements were performed using, a direct beam, cadmium filtered beam and boron carbide filtered beam. The neutron doses were measured using thermolumin-escent Li/sub 2/B/sub 4/O/sub 7/ detectors. The measured data have been analyzed and a group of attenuation curves were given for beams of reactor neutrons of different energy. These curves show that cadmium and boron carbide filters tend to decrease the neutron doses specially at the beginning of penetration. The data were transformed to that which would be obtained using neutron sources of different geometries.

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

    International Nuclear Information System (INIS)

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

    2001-03-01

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

  7. Neutron flux measurement in the thermal column of the Malaysian TRIGA mark II reactor with MCNP verification

    International Nuclear Information System (INIS)

    Abdel Munem, E.; Shukri, A.; Tajuddin, A.A.

    2006-01-01

    A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)

  8. Thermal neutron flux measurements in the rotary specimen rack of the IPR-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G. do Prado; Rodrigues, Rogério R.; Souza, Luiz Claudio A., E-mail: souzarm@cdtn.br, E-mail: rrr@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The thermal neutron flux in the rotary specimen rack of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil, has been measured by the neutron activation method, using bare and cadmium covered gold foils. Those foils were irradiated in the rotary specimen rack with the reactor at 100 kW. The reactor core configuration has 63 fuel elements, composed of 59 original aluminum-clad elements and 4 stainless steel-clad fuel elements. The gamma activities of the foils were measured using Ge spectrometer. The perturbations of the thermal neutron flux caused by the introduction of an absorbing foil into the medium were considered in order to obtain accurate determination of the flux. The thermal neutron flux obtained was 7.4 x 10{sup 11} n.cm{sup -2}.s{sup -1}. (author)

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

    International Nuclear Information System (INIS)

    Marques, Andre Luis Ferreira

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  11. 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

  12. Preliminary fission product energy release measurements for thermal neutron fission of 235U

    International Nuclear Information System (INIS)

    Dickens, J.K.; Love, T.A.; McConnell, J.W.; Emery, J.F.; Peelle, R.W.

    1976-03-01

    An experimental system to measure time-dependent spectra of β and γ rays from fission-product production by thermal neutron fission of 235 U is described, and for each component (β and γ) the system has been tested with a pilot data-accumulation run. Data reduction techniques are described and test results given. Gamma-ray spectra are compared with calculations using ENDF/B-IV data files. Both β- and γ-ray spectra were integrated to give total yields and total energy-release results for times after fission between 3 and 14400 sec. These preliminary integral data are compared with previous measurements and with integral calculations using ENDF/B-IV data files

  13. Measurement of ν(m*) for thermal neutron-induced fission of Pu-239

    International Nuclear Information System (INIS)

    Nakagome, Yoshihiro; Nishio, Katsuhisa; Jinno, Ikuo; Kimura, Itsuo

    1995-01-01

    On the thermal fission of Pu-239, the distribution of prompt neutron number ν(m*) as a function of fragment was measured by the dual velocity dual energy (2V2E) measurement method. Each kinetic energy and velocity (flight time) of both fragments were determined independently. The coincidence counting phenomena were only recorded of every phenomenum by the multi dimension data input system at the test mode. In TOF spectrum of Pu(n th , f)-239 fragments, (flight range 28.9 cm) at light nuclear group. Two peaks were observed in the kinetic spectrum at 103.5 MeV (light nuclear group) and 75.4 MeV (heavy one), more wide than that of light one. The mass distribution of fragments agreed very closely with the result of Neiler et al. The kinetic energy distribution of fragments was infair agreement with the results of Wagemans et al., too. ν(m+) was calculated on the basis of Brosa's model and compared with the experiment. Both results were seemed to be the same qualitatively, but different quantitatively. On each m*, two dispersion values were derived from the probability distribution of kinetic energy (Pm*(E)) and prompt neutron emission number (Pm*(ν)). We found that both values gave a good correlation. (S.Y.)

  14. Bibliography for thermal neutron scattering

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  15. Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 158B/159A

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Larry Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, David Torbet [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walker, Billy Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-11-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 158B/159A which were measured by the Radiation Measurements Laboratory (RML).

  16. Measurements of thermal neutron fluence in the bunker of a cyclotron for PET isotope production; Medidas de fluencia de neutrones termicos en el bunker de un ciclotron de produccion de isotopos para PET

    Energy Technology Data Exchange (ETDEWEB)

    Mendez Villafane, R.; Sansoloni florit, F.; Lagares gonzalez, J. L.; Llop Roig, J.; Guerrero Araque, J. E.; Muniz Gutierrez, J. L.; Perez Morales, J. M.

    2011-07-01

    To measure the neutron spectrum has been used spectrometry system based on Bonner spheres with Au flakes as thermal neutron detector at its center while the results are still pending and will be analyzing another job.

  17. Measurement of the thermal and fast neutron flux in a research reactor with a Li and Th loaded optical fibre detector

    CERN Document Server

    Yamane, Y; Misawa, T; Karlsson, J K H; Pázsit, I

    1999-01-01

    The spatial dependence of thermal and fast neutron flux was measured axially in the core of a 1 MW research reactor. The measurements were made by a thin optical fibre detector with a neutron sensitive ZnS(Ag) scintillation tip. For thermal neutrons sup 6 Li was used, whereas for fast neutrons sup 2 sup 3 sup 2 Th was used as neutron converter. The spatial dependence was measured by moving the fibre axially with a uniform speed. The measurement takes a few minutes, compared to up to 10 h with the conventional wire activation method. Comparison with traditional measurements shows a good agreement. (author)

  18. Angular Correlation Measurements on Gamma-Ray Cascades Following Thermal Neutron Capture

    International Nuclear Information System (INIS)

    Schmidt, H.; Heck, D.

    1969-01-01

    An experimental arrangement for angular correlation measurements on gamma-ray cascades following thermal neutron capture which is operated on-line to a computer is described. All of the interesting information can be accumulated simultaneously so that the determination of a large number of angular correlations is possible. Systematic errors arising from the coincident background under the peaks and from the complex structure of the capture spectrum can be eliminated either by a two-dimensional spectrum stripping procedure or, with sufficient accuracy, by application of the double-window technique. Practicable handling of data and comprehensive controls are thus possible. The extensive set of data obtained allows the results and their mutual consistency to be checked. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  20. 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. 

  1. Electronic instrumentation system for pulsed neutron measurements

    International Nuclear Information System (INIS)

    Burda, J.; Igielski, A.; Kowalik, W.

    1982-01-01

    An essential point of pulsed neutron measurement of thermal neutron parameters for different materials is the registration of the thermal neutron die-away curve after a fast neutron bursts have been injected into the system. An electronic instrumentation system which is successfully applied for pulsed neutron measurements is presented. An important part of the system is the control unit which has been designed and built in the Laboratory of Neutron Parameters of Materials. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

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

    CERN Document Server

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  5. Neutron thermalization in light water

    International Nuclear Information System (INIS)

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

    1975-05-01

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

  6. Extended use of alanine irradiated in experimental reactor for combined gamma- and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of (14)C by LSC.

    Science.gov (United States)

    Bartoníček, B; Kučera, J; Světlík, I; Viererbl, L; Lahodová, Z; Tomášková, L; Cabalka, M

    2014-11-01

    Gamma- and neutron doses in an experimental reactor were measured using alanine/electron spin resonance (ESR) spectrometry. The absorbed dose in alanine was decomposed into contributions caused by gamma and neutron radiation using neutron kerma factors. To overcome a low sensitivity of the alanine/ESR response to thermal neutrons, a novel method has been proposed for the assessment of a thermal neutron flux using the (14)N(n,p) (14)C reaction on nitrogen present in alanine and subsequent measurement of (14)C by liquid scintillation counting (LSC). Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Measurements of thermal neutron cross section and resonance integral for the 237Np(n,γ)238Np reaction

    International Nuclear Information System (INIS)

    Kobayashi, Katsuhei; Yamanaka, Akihiro; Kimura, Itsuro

    1994-01-01

    Making use of a standard neutron spectrum field with a pure Maxwellian distribution at the heavy water thermal neutron facility of the Kyoto University Reactor (KUR), the thermal neutron cross section for the 237 Np(n,γ) 238 Np reaction was measured by the activation method, using a high purity Ge detector. The result is 158 ± 3 b, which is obtained relative to the reference value of 98.65 ± 0.09 b for the 197 Au(n,γ) 198 Au reaction. The present value is lower by about 13% than that of the ENDF/B-VI data. The data given by Mughabghab and obtained from JENDL-3 are larger by 11 to 14% than the present measurement. The resonance integral for the 237 Np(n,γ) 238 Np reaction was measured with a 1/E standard neutron spectrum in the cavity of the central graphite reflector region between the two-divided cores of the Kinki University Reactor (UTR-KINK), relative to the reference value of 1,550 ± 28 b for the 197 Au(n,γ) 198 Au reaction. By defining the Cd cut-off energy as 0.5 eV, the present resonance integral is 652 ± 24 b, which is in good agreement with the JENDL-3, ENDF/B-VI and Mughabghab data. However, most of the old experimental data are, in general, larger by 24 to 38% than the present measurement

  8. Electronic Reliability and the Environmental Thermal Neutron Flux

    National Research Council Canada - National Science Library

    Clark, John

    2002-01-01

    .... The goal of this project is to characterize the environmental thermal neutron flux with respect to electronic reliability by performing measurements of the thermal neutron flux in various locations...

  9. Measurement of fission yields far from the center of isotopic distributions in the thermal neutron fission of 235U

    International Nuclear Information System (INIS)

    Shmid, M.

    1979-08-01

    The main purpose of this work was to measure independent yields, in the thermal neutron fission of 235 U, of fission products which lie far from the centers of the isotopic and isobaric yield distributions. These measurements were used to test the predictions of semi-empirical systematics of fission yields and theoretical fission models. Delay times were measured as a function of temperature in the range 1200-2000degC. The very low delay times achieved in the present work permitted expanding the measurable region to the isotopes 147 , 148 Cs and 99 Rb which are of special interest in the present work. The delay times of Sr and Ba isotopes achieved were more than two orders of magnitude lower than values reported in the literature and thus short-lived isotopes of these elements could be separated for the first time by mass spectrometry. The half-lives of 147 Ba, 148 Ba, 149 La and 149 Ce were measured for the first time. The isotopic distributions of fission yields were measured for the elements Rb, Sr, Cs and Ba in the thermal neutron fission of 235 U, those of 99 Rb, 147 Cs and 148 Cs having been measured for the first time. A comparison of the experimental yields with the predictions of the currently accepted semi-empirical systematics of fission yields, which is the odd-even effect systematics, shows that the systematics succeeds in accounting for the strong odd-even proton effect and the weaker odd-even neutron effect and also in predicting the shape of the distributions in the central region. It is shown that prompt neutron emission broadens the distribution only slightly in the wing of heavy isotopes and more significantly in the wing of light isotopes. But the effect of prompt neutron emission cannot explain the large discrepancies existing between the predictions of fission models and the experimentally measured fission yield in the wings of the isotopic distributions. (B.G.)

  10. Activation measurements for thermal neutrons. Part A. Cobalt (60Co) activation

    International Nuclear Information System (INIS)

    Kerr, George D.; Shizuma, Kiyoshi; Endo, Satoru; Maruyama, Takashi; Cullings, Harry M.; Komura, Kazuhisa; Okumura, Yutaka; Egbert, Stephen D.

    2005-01-01

    The 60 Co measurements at Hiroshima and Nagasaki were reviewed and compared with the results of the new DS02 calculations for the two cities. Some corrections were made to previously published data from the 1965 measurements of JNIRS, ground ranges of all of the 60 Co measurements were reviewed and new ground ranges were determined when possible by transforming sample coordinates to the new city maps for Hiroshima and Nagasaki, and transmission factors were investigated for a number of samples used in the 60 Co measurements. Transmission factors were not calculated for all samples used in the 60 Co measurements but some general rules were given for estimating transmission factors for these samples. Experimental measurements of the background 60 Co activity from environmental neutrons were also reviewed, and it was found that the 60 Co background from environmental neutrons was not an important factor in the 60 Co measurements of bomb-induced activity at either city. (J.P.N.)

  11. A base to standardize data processing of cadmium ratio RCd and thermal neutron flux measurements on reactor

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1993-08-01

    The cadmium ratio R Cd and thermal neutron flux are usually measured in a reactor. But its data process is rather complex. The results from same measured data differ by different existing process methods. The purpose of this work is to standardize data processing in R Cd and thermal neutron flux measurements. A natural choice for this purpose is to derive a R Cd formula based on standard average thermal activation cross section and resonance integral and to define related parameters or factors that provide an unique base for comparison between different measurements in laboratories. The parameters or factors include E c , F m , F m ' and G th ' in thermal energy region due to upper truncated Maxwellian distribution and E Cd , F Cd , G r and S r in intermediate energy region. They are the function of multiple variables. The Au foil is used as an example to demonstrate their behaviors by chosen figures and tables which provide for practical data process by hand. The work also discusses limitation of R Cd measurement in terms of so called available and optimum region and notes that Co and Mn foils have a much wider available region among Au, In, Mn, W and Co, the commonly used detector foils

  12. First Measurements of the Inclined Boron Layer Thermal-Neutron Detector for Reflectometry

    Energy Technology Data Exchange (ETDEWEB)

    Clonts, Lloyd G [ORNL; Crow, Lowell [ORNL; Van Vuure, Thorwald L [ORNL; Robertson, Lee [ORNL; Riedel, Richard A [ORNL; Richards, John D [ORNL; Cooper, Ronald G [ORNL; Remec, Igor [ORNL; Ankner, John Francis [ORNL; Browning, Jim [ORNL

    2010-01-01

    A prototype detector based on the inclined boron layer principle is introduced. For typical measurement conditions at the Liquids Reflectometer at the Spallation Neutron Source, its count rate capability is shown to be superior to that of the current detector by nearly two orders of magnitude.

  13. Design and construction of a 10B coated ion chamber for the measurement of a thermal neutron flux

    International Nuclear Information System (INIS)

    Vite T, M.

    1979-01-01

    A model of an ionization chamber cover with 10 B, for the measurement of thermal neutron flux is presented, the developed chamber is made of a cylindrical vessel in which interior exist 7 aluminium electrodes which by electrodeposition were covered with a coat of 10 B with a thickness of 0.68 mg/cm 2 . Once the chamber is filled with N 2 to a pressure of 1.33 atm., it was exposed to a flux of thermal neutrons at the order 10 4 n/cm 2 -seg obtaining an ionization current of 10 -11 amp. which can be easily distinguished from the leak current which order is of 10 -12 amp. The conventional electronics associated to the chamber allows to process the ionization current in pulse form. The relation of noise to signal is approximately of 8, for which the pulses can be differentiated from the electronic noise without any problem. (author)

  14. Simultaneous thermal neutron decay time and porosity logging system

    International Nuclear Information System (INIS)

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

    1981-01-01

    A pulsed high-energy neutron source irradiates the earth formations surrounding a borehole with bursts of fast neutrons. A pair of detectors, one sensitive to epithermal neutrons and the other sensitive to fast neutrons and thermal neutron capture gamma rays, measure the fast and epithermal neutron populations at their effective distances from the source. The fast neutron measurements can be separated from thermal neutron interactions by time gating techniques and by pulse shape discrimination. The measurments of the fast and epithermal neutron populations at the two detectors may then be interpreted in terms of the earth formation porosity in accordance with predetermined relationships. Between neutron bursts capture gamma rays are detected in two or more time intervals, and these measurements are used to derive the thermal neutron capture cross section of the formation

  15. 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.)

  16. Thermal neutron absorption cross-section measured on rock samples and brines in the Institute of Nuclear Physics

    International Nuclear Information System (INIS)

    Czubek, J.A.; Drozdowicz, K.; Krynicka-Drozdowicz, E.; Igielski, A.; Woznicka, U.

    1983-01-01

    In consecutive measurements the rock sample (having a fixed and well known shape -in our case it is a sphere or a cylinder and the sample is powdered or liquid) is enveloped in shells of a plexiglass moderator (the neutron parameters of which are known) of variable thickness and irradiated with the pulsed beam of fast neutrons. The die-away rate of thermal neutrons escaping from the whole system is measured. The absorption cross-section of the sample is found as the intersection of the experimental curve (i.e. die -away rate vs thickness of the moderator) with the theoretical one. The theoretical curve is calculated for a given moderator under the assumption of a constant value of the neutron flux inside the sample. This method is independent of the value of the transport cross-section of the sample. It has been checked on artificial materials with a well known elemental composition (liquid or solid) and on the natural brines and rock samples (basalts and dolomite). A special method of calculation of the variance of the measurement has been established. It is based on the multiple computer simulations of all experimental data used in the computation. The one standard deviation of our methods is of the order of 1 up to 3 capture units (1 c.u. = 10 -3 cm -1 ). The volume of the sample needed is of the order of 500ccm. (author)

  17. Alternative method for thermal neutron flux measurements based on common boric acid as converter and Lr-15 detectors

    Energy Technology Data Exchange (ETDEWEB)

    Palacios, D.; Greaves, E. D.; Sajo B, L.; Barros, H. [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas (Venezuela, Bolivarian Republic of); Ingles, R. [Universidad Nacional de San Antonio Abad del Cusco, Av. de la Cultura No. 733, Cusco (Peru)

    2010-02-15

    A method to determine the flux and angular distribution of thermal neutrons with the use of Lr-115 detectors was developed. The use of the Lr-115 detector involves the exposure of a pressed boric acid sample (tablet) as a target, in tight contact with the track detector, to a flux of thermalized neutrons. The self-absorption effects in thin films or foil type thermal neutron detectors can be neglected by using the Lr-115 detector and boric acid tablet setup to operate via backside irradiation. The energy window and the critical angle-residual energy curve were determined by comparisons between the experimental and simulated track parameters. A computer program was developed to calculate the detector registration efficiency, so that the thermal neutron flux can be calculated from the track densities induced in the Lr-115 detector using the derived empirical formula. The proposed setup can serves as directional detector of thermal neutrons. (Author)

  18. 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

  19. Activation measurements for thermal neutrons. Part D. 36Cl measurements in the United States

    International Nuclear Information System (INIS)

    Straume, Tore; Men, Ping; Marchetti, Alfredo A.; Egbert, Stephen D.; Roberts, James A.; Fujita, Shoichiro; Shizuma, Kiyoshi; Hoshi, Masaharu

    2005-01-01

    A large number of measurements were performed in the United States of 36 Cl in both granite and concrete samples obtained from various locations and distances in Hiroshima and Nagasaki. These measurements employed accelerator mass spectrometry (AMS) to quantify the number of atoms of 36 Cl per atom of total Cl in the sample. Results of these measurements are presented in this section and discussed in the context of the DS02 dosimetry reevaluation effort for Hiroshima and Nagasaki atomic-bomb survivors. (J.P.N.)

  20. Measurement of the thermal neutron capture cross section and the resonance integral of the {sup 109}Ag(n,{gamma}){sup 110m}Ag reaction

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, S.; Wada, H.; Furutaka, K.; Harada, H.; Katoh, T. [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan)

    2001-03-01

    The thermal neutron capture cross section ({sigma}{sub 0}) and the resonance integral (I{sub 0}) of the {sup 109}Ag(n,{gamma}) reaction were measured by the activation and {gamma}-ray spectroscopic methods to develop a neutron flux monitor for the long irradiation. (author)

  1. Measurement of the thermal neutron capture cross section and the resonance integral of the 109Ag(n,γ)110mAg reaction

    International Nuclear Information System (INIS)

    Nakamura, S.; Wada, H.; Furutaka, K.; Harada, H.; Katoh, T.

    2001-01-01

    The thermal neutron capture cross section (σ 0 ) and the resonance integral (I 0 ) of the 109 Ag(n,γ) reaction were measured by the activation and γ-ray spectroscopic methods to develop a neutron flux monitor for the long irradiation. (author)

  2. RPC for thermal neutron detection

    Energy Technology Data Exchange (ETDEWEB)

    Arnaldi, R [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Chiavassa, E [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Colla, A [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Cortese, P [INFN Alessandria and Universita del Piemonte Orientale ' Amedeo Avogadro' , Alessandria (Italy); Dellacasa, G [INFN Alessandria and Universita del Piemonte Orientale ' Amedeo Avogadro' , Alessandria (Italy); Marco, N De [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Ferretti, A [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Gagliardi, M [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Gallio, M [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Gemme, R [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Musso, A [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Oppedisano, C [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Piccotti, A [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Poggio, F [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Scomparin, E [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy); Vercellin, E [INFN Torino and Dip. di Fisica Sperimentale dell' Universita di Torino (Italy)

    2006-05-15

    The possibility to detect thermal neutrons with single gap Resistive Plate Chambers has been investigated. To detect neutrons a {sup 10}B{sub 4}C thin coating on the inner surface of one RPC electrode is used as thermal neutron converter. The RPC detects the charged particles generated by neutrons via the (n, {alpha}) reaction on Boron. Tests on converter samples have been performed with a thermalized {sup 252}Cf source in order to evaluate the conversion efficiency: a good agreement between experimental results and simulation has been achieved. A detector prototype has been developed and tested on a low energy neutron beam at the European laboratories JRC in Belgium. A detailed description of the detector and the experimental test results are presented.

  3. Measurement of the para-hydrogen concentration in the ISIS moderators using neutron transmission and thermal conductivity

    Science.gov (United States)

    Romanelli, Giovanni; Rudić, Svemir; Zanetti, Matteo; Andreani, Carla; Fernandez-Alonso, Felix; Gorini, Giuseppe; Krzystyniak, Maciej; Škoro, Goran

    2018-04-01

    We present an experimental study to determine the para-hydrogen concentration in the hydrogen moderators at the ISIS pulsed neutron and muon source. The experimental characterisation is based on neutron transmission experiments performed on the VESUVIO spectrometer, and thermal conductivity measurements using the TOSCA para-hydrogen rig. A reliable estimation of the level of para-hydrogen concentration in the hydrogen moderators is of crucial importance in the framework of a current project to completely refurbish the first target station at ISIS. Moreover, we report a new measurement of the total neutron cross section for normal hydrogen at 15 K on the broad energy range 3 meV -10 eV suggesting a revision of the most recent nuclear libraries for incident neutron energies lower than 10 meV. Finally, we characterise systematic errors affecting the para-hydrogen level estimation due to conversion from para to ortho hydrogen, as a function of the time a batch of gas spends in every component of our gas panel and apparatus.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  5. Photoneutron cross sections measurements in 13C with thermal neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Semmler, Renato; Carbonari, Artur W.; Terremoto, Luis A.A.; Goncalez, Odair L.

    2007-01-01

    Photoneutrons cross sections measurements of 13 C have been obtained in energy interval between 5,3 and 10,8 MeV, using neutron capture gamma-rays with high resolution in energy (3 - 21 eV), produced by 21 target materials, placed inside a tangential beam port, near the core of the IPEN/CNEN-SP IEA-R1 (2MW) research reactor. The sample have been irradiated inside a 4p geometry neutron detector system 'Long Counter', 520,5 cm away from the capture target. The capture gamma-ray flux was determined by means of the analysis of the gamma spectrum obtained by using a Ge(Li) solid-state detector (EG and G ORTEC, 25 cm 3 , 5%), previously calibrated with capture gamma-rays from a standard target of Nitrogen (Melamine). The neutron photoproduction cross section has been measured for each target capture gamma-ray spectrum (compound cross section). A methodology for unfolding the set of experimental compound cross sections, have been used in order to obtain the cross sections at specific excitation energy values (principal gamma lines energies of the capture targets). The cross sections were compared with experimental data, reported by other authors, using different gamma-ray sources. A good agreement was observed between in this work and reported in the literature. (author)

  6. Photoneutron cross sections measurements in {sup 13}C with thermal neutron capture gamma-rays

    Energy Technology Data Exchange (ETDEWEB)

    Semmler, Renato; Carbonari, Artur W.; Terremoto, Luis A.A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: rsemmler@ipen.br; carbonar@ipen.br; laaterre@ipen.br; Goncalez, Odair L. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mail: odairl@ieav.cta.br

    2007-07-01

    Photoneutrons cross sections measurements of {sup 13}C have been obtained in energy interval between 5,3 and 10,8 MeV, using neutron capture gamma-rays with high resolution in energy (3 - 21 eV), produced by 21 target materials, placed inside a tangential beam port, near the core of the IPEN/CNEN-SP IEA-R1 (2MW) research reactor. The sample have been irradiated inside a 4p geometry neutron detector system 'Long Counter', 520,5 cm away from the capture target. The capture gamma-ray flux was determined by means of the analysis of the gamma spectrum obtained by using a Ge(Li) solid-state detector (EG and G ORTEC, 25 cm{sup 3}, 5%), previously calibrated with capture gamma-rays from a standard target of Nitrogen (Melamine). The neutron photoproduction cross section has been measured for each target capture gamma-ray spectrum (compound cross section). A methodology for unfolding the set of experimental compound cross sections, have been used in order to obtain the cross sections at specific excitation energy values (principal gamma lines energies of the capture targets). The cross sections were compared with experimental data, reported by other authors, using different gamma-ray sources. A good agreement was observed between in this work and reported in the literature. (author)

  7. Measurement of the effective thermal cross section of {sup 134}Cs by triple neutron capture reaction

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Shoji; Harada, Hideo; Katoh, Toshio [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works; Hatsukawa, Yuichi; Shinohara, Nobuo; Hata, Kentaro; Kobayashi, Katsutoshi; Motoishi, Shoji; Tanase, Masakazu

    1998-03-01

    The effective thermal cross section ({sigma}{sub eff}) of the {sup 134}Cs(n,{gamma}){sup 135}Cs reaction was measured by the activation method and the {gamma}-ray spectroscopic method in order to obtain fundamental data for research on the transmutation of nuclear wastes. The effective thermal cross section of the reaction {sup 134}Cs(n,{gamma}){sup 135}Cs was found to be 140.6{+-}8.5 barns. (author)

  8. Determination of the thermal neutron absorption cross section for rock samples by a single measurement of the time decay constant

    International Nuclear Information System (INIS)

    Krynicka, E.

    1993-01-01

    A calibration method for the determination of the thermal neutron macroscopic mass absorption cross section for rock samples is presented. The standard deviation of the final results is discussed in detail. A big advantage of the presented method is that the calibration curves have been found using the results obtained for a variety of natural rock samples of different stratigraphies and lithologies measured by Czubek's methods. An important part of the paper is a through analysis of the standard deviation of the final result. (author). 13 refs, 11 figs, 5 tabs

  9. Cosmic-ray thermal neutron detection for environmental purposes

    Science.gov (United States)

    Looms, M. C.; Rosolem, R.; Klinkby, E. B.; Andreasen, M.

    2017-12-01

    Cosmic-ray neutron detection has been successfully used to produce time-series of hectometer-scale soil moisture estimates at various soil types and land covers. The method relies on measurements of epithermal neutron intensities with energies in the range of approximately 10-1000 eV (electron Volt). As the cosmic-ray neutron technology matures, additional sensing possibilities emerge, such as biomass, snow and litter layer thickness detection. The physical processes controlling neutron transport depend on the neutron energy. Because of this, many of these new applications benefit from measurements of cosmic-ray neutrons at multiple energy levels. For instance, several published studies suggest a correlation between the thermal-to-epithermal ratio and amount of biomass, where thermal neutrons refer to neutron energies below 0.5 eV. However, the vast majority of the theoretical investigations to date have focused on epithermal neutrons for environmental applications, since epithermal neutrons are more sensitive to the presence of hydrogen than at other energies. As a result, not much is known about thermal neutron transport in environmental systems. In this study, we investigate the thermal neutron behavior in environmental settings using the neutron transport model Monte Carlo N-Particle radiation transport code (MCNP6). First, we model the two common detector types: 1) The bare detector, measuring mainly thermal neutrons, and 2) the moderated detector, measuring mainly epithermal neutrons. The percentage of epithermal neutrons captured using the bare detector and the percentage of the thermal neutrons captured using the moderated detector is quantified for two separate detector systems and compared to measured values. Second, we determine whether it is relevant to correct thermal measurements for changes in vapor pressure, using a similar procedure previously proposed for epithermal neutron measurements. Finally, we investigate the area of influence of the two

  10. Three frequency modulated combination thermal neutron lifetime log and porosity

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. Thermal neutron diffusion cooling coefficient for plexiglass

    International Nuclear Information System (INIS)

    Drozdowicz, K.

    1992-08-01

    The thermal neutron diffusion cooling coefficient is a macroscopic material parameter. It is needed for description of the decay of the thermal neutron pulse in a medium and gives information of the diffusion cooling of the thermal neutron spectrum in a bounded volume. Experimental results from various measurements for plexiglass are overviewed in the paper. A method for theoretical, exact calculation of the parameter is presented. The formula utilizes some other thermal neutron parameters and a cooling function, i.e. the function which describes the deviation of the neutron spectrum in a bounded system from the distribution in an infinite one. The energy dependence of the function is obtained numerically from relations which results from the eigenvalue problem of the scattering operator when both the decay constant and the spectrum of the thermal neutron flux are developed on powers of the geometrical buckling. The case of a 1/ν absorption cross section is considered. The calculation utilizes a synthetic scattering function elaborated for hydrogenous media by GRANADA (1985). The influence of some quantities used in the calculation on the final result is investigated. The obtained value of the diffusion cooling coefficient for plexiglass is C = 6514 cm 4 s -1 at the temperature of 20 degrees C. The uncertainty is estimated to be ± 100 cm 4 s -1 within the physical model of the scattering kernel used. (au)

  12. Dynamic response of thermal neutron measurements in electrochemically produced cold fusion subject to pulsed current

    International Nuclear Information System (INIS)

    Granada, Jose; Converti, Jose; Mayer, Roberto; Guido, German; Florido, Pablo; Patino, Nestor; Sobehart, Leonardo; Gomez, Silvia; Larreteguy, Axel

    1988-01-01

    The present work shows the results of measurements performed on electrolytic cells using a high efficiency (22%) neutron detection system in combination with a procedure involving a non-stationary current through the cell's circuit. Cold fusion was produced in electrolytic cells containing LiH dissolved in heavy water with a palladium cathode. The dynamic response to low frequency current pulses was measured. Characteristic patterns showing one or two bumps were obtained in a repeatable fashion. These patterns are strongly dependent on the previous charging history of the cathode. The technique employed seems to be very convenient as a research tool for a systematic study of the different variables governing the phenomenon. (Author)

  13. 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

  14. Measurement of the fission cross-section of 235U and 239Pu for thermal neutrons

    International Nuclear Information System (INIS)

    Fraysse, G.; Prosdocimi, A.; Netter, F.; Samour, C.

    1965-01-01

    Improved techniques of fast detection have been applied for determining the fission cross-sections of 235 U and 239 Pu with reference to the absorption cross-section of Boron. Monochromatic neutron beams of 0.0322 eV, 0.0626 eV and 0.275 eV have been employed. Use has been made of a Xe-filled gaseous scintillator and of a low-geometry solid state ion chamber. Both measured alpha and fission rates. The results at the reference energy of 0.0253 eV are: (σ F ) 0 235 U = 588 ± 10 barns (σ F ) 0 239 Pu = 738 ± 7 barns. (authors) [fr

  15. Microdosimetric measurements in the thermal neutron irradiation facility of LENA reactor

    International Nuclear Information System (INIS)

    Colautti, P.; Moro, D.; Chiriotti, S.; Conte, V.; Evangelista, L.; Altieri, S.; Bortolussi, S.; Protti, N.; Postuma, I.

    2014-01-01

    A twin TEPC with electric-field guard tubes has been constructed to be used to characterize the BNCT field of the irradiation facility of LENA reactor. One of the two mini TEPC was doped with 50 ppm of 10 B in order to simulate the BNC events occurring in BNCT. By properly processing the two microdosimetric spectra, the gamma, neutron and BNC spectral components can be derived with good precision (∼6%). However, direct measurements of 10 B in some doped plastic samples, which were used for constructing the cathode walls, point out the scarce accuracy of the nominal 10 B concentration value. The influence of the Boral ® door, which closes the irradiation channel, has been measured. The gamma dose increases significantly (+51%) when the Boral ® door is closed. The crypt-cell-regeneration weighting function has been used to measure the quality, namely the RBE µ value, of the radiation field in different conditions. The measured RBE µ values are only partially consistent with the RBE values of other BNCT facilities. - Highlights: • A counter with two mini TEPCs, both equipped with electrical-field guard tubes, has been constructed. • The microdosimetric spectrum of the LENA-reactor irradiation vane has been studied. • The radiation-field quality (RBE) assessment confirms that the D n /D tot ratio is not an accurate parameter to characterize the BNCT radiation field

  16. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Science.gov (United States)

    Carcreff, H.; Salmon, L.; Lepeltier, V.; Guyot, J. M.; Bouard, E.

    2018-01-01

    Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the "zero method". Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the "zero method" measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between "zero" and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions, obtained with the Self-Power Neutron

  17. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Directory of Open Access Journals (Sweden)

    Carcreff H.

    2018-01-01

    Full Text Available Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the “zero method”. Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the “zero method” measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between “zero” and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions

  18. 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

  19. 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

  20. 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

  1. Precision measurement of thermal neutron beam densities using a 3He proportional counter

    DEFF Research Database (Denmark)

    Als-Nielsen, Jens Aage; Bahnsen, A.; Brown, W.K.

    1967-01-01

    of ±0.4%. Fundamental advantages of the method include the 1ν dependence of the 3He(n, p)T cross section up to 1 keV, and the assurance of homogeneity even for very small macroscopic cross sections, because of the gaseous detector material. Although the method requires a relatively clean neutron beam...

  2. Direct thermal neutron capture

    International Nuclear Information System (INIS)

    Raman, S.; Kahane, S.; Lynn, J.E.

    1987-01-01

    We discuss the direct-capture theory pertaining to primary electric dipole (E1) transitions following slow-neutron capture. For light nuclides that we have studied (including 9 Be, 12 C, 13 C, 24 Mg, 25 Mg, 26 Mg, 32 S, 33 S, 34 S, 40 Ca, and 44 Ca), estimates of direct-capture cross sections using optical-model potentials with physically realistic parameters, are in reasonable agreement with the data. Minor disagreements that exist are consistent with extrapolations to light nuclides of generally accepted formulations of compound-nucleus capture. We also discuss the channel-capture approximation which is, in general, a good representation of these cross sections in heavier nuclei particularly if the scattering lengths are not different from the corresponding potential radii. We also draw attention to cases where the use of this formula leads to inaccurate predictions. 9 refs., 1 fig., 2 tab

  3. Development of a fast, thermal neutron, and gamma ray component discriminator

    International Nuclear Information System (INIS)

    Yoneda, K.; Kobayashi, H.

    1996-01-01

    A fast neutron, thermal neutron, and gamma ray component (FTG) discriminator is proposed in this study. The FTG discriminator was tested using neutron beams generated from the YAYOI fast neutron reactor and the Kyoto University Research Reactor. Various beams were successfully discriminated into fast, thermal, cold neutrons and gamma rays. The measurement technique and the reliability of the FTG discriminator are discussed. (orig.)

  4. Applications of thermal neutron scattering in biology, biochemistry and biophysics

    International Nuclear Information System (INIS)

    Worcester, D.L.

    1977-01-01

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

  5. 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)

  6. 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.))

  7. Summary of neutron measurements for the Viking Program

    International Nuclear Information System (INIS)

    Anderson, M.E.

    1975-01-01

    The results of neutron measurements for 238 Pu-fueled, 683-W (thermal) capsules fabricated for the Viking Program (Mars Lander) are presented. These results include, for each capsule, the total neutron emission rate and neutron multiplication and, for one capsule, the neutron energy spectrum. A precision long counter was used for the neutron emission rate measurements and a single stilbene crystal for the neutron spectrum measurement. (U.S.)

  8. Measurement of neutron diffraction with compact neutron source RANS

    Science.gov (United States)

    Ikeda, Y.; Takamura, M.; Taketani, A.; Sunaga, H.; Otake, Y.; Suzuki, H.; Kumagai, M.; Oba, Y.; Hama, T.

    2016-11-01

    Diffraction is used as a measurement technique for crystal structure. X-rays or electron beam with wavelength that is close to the lattice constant of the crystal is often used for the measurement. They have sensitivity in surface (0.01mm) of heavy metals due to the mean free path for heavy ions. Neutron diffraction has the probe of the internal structure of the heavy metals because it has a longer mean free path than that of the X-rays or the electrons. However, the neutron diffraction measurement is not widely used because large facilities are required in the many neutron sources. RANS (Riken Accelerator-driven Compact Neutron Source) is developed as a neutron source which is usable easily in laboratories and factories. In RANS, fast neutrons are generated by 7MeV protons colliding on a Be target. Some fast neutrons are moderated with polyethylene to thermal neutrons. The thermal neutrons of 10meV which have wavelength of 10nm can be used for the diffraction measurement. In this study, the texture evolution in steels was measured with RANS and the validity of the compact neutron source was proved. The texture of IF steel sheets with the thickness of 1.0mm was measured with 10minutes run. The resolution is 2% and is enough to analyze a evolution in texture due to compression/tensile deformation or a volume fraction of two phases in the steel sample. These results have proven the possibility to use compact neutron source for the analysis of mesoscopic structure of metallic materials.

  9. Neutron fluence measurement in nuclear facilities

    International Nuclear Information System (INIS)

    Camacho L, M.E.

    1997-01-01

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

  10. Activation Measurements for Thermal Neutrons, U.S. Measurements of 36Cl in Mineral Samples from Hiroshima and Nagasaki; and Measurement of 63 Ni in Copper Samples From Hiroshima by Accelerator Mass Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-01-14

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

  11. 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.

  12. Absolute measurements of the thermal neutron flux by the foil activation method using the 4πβ-γ coincidence technique

    International Nuclear Information System (INIS)

    Aranda Amado Furieri, R.C. de.

    1977-12-01

    An analysis of the correction factors required for the β-γ coincidence method is presented together with a listing of the various formulae involved in the determination of radioactive sources. The detection system including the activation detectores are described and the results are shown for the absolute measurements of thermal neutron flux carried out in the core of the Argonaut Reactor, at Instituto de Engenharia Nuclear (IEN), Brazil. (Author) [pt

  13. Storage phosphors for thermal neutron detection

    CERN Document Server

    Sidorenko, A V; Dorenbos, P; Le Masson, N J M; Rodnyi, P A; Eijk, C W E; Berezovskaya, I V; Dotsenko, V P

    2002-01-01

    The commercial BaFBr:Eu sup 2 sup +centre dot Gd sub 2 O sub 3 image plate (IP) is used nowadays for thermal neutron detection. However, it is rather sensitive to gamma-ray background, which can deteriorate the image quality. We focused our research on the development of new types of storage phosphors with the general formula M sub 2 B sub 5 O sub 9 Br:Ce sup 3 sup + (M=Sr, Ca). Neutron detection is based on the sup 1 sup 0 B(n,alpha) reaction. The advantages of this system are the low Z sub e sub f sub f , and the 40 times higher energy deposition resulting from the neutron capture reaction in comparison with that in the commercial IP. Here we present storage and spectroscopic properties of a series of newly synthesized haloborates. Comparative measurements with commercial IPs were done under neutron and beta irradiation. A satisfying light output of optically stimulated luminescence was achieved upon neutron irradiation.

  14. Development of high flux thermal neutron generator for neutron activation analysis

    International Nuclear Information System (INIS)

    Vainionpaa, Jaakko H.; Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K.; Jones, Glenn; Pantell, Richard H.

    2015-01-01

    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 7 n/cm 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 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

  15. 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.

  16. 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.

  17. 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.)

  18. Activation of endogeneous retroviruses in mouse cells by thermal neutrons

    International Nuclear Information System (INIS)

    Niwa, Ohtsura

    1987-01-01

    The effect of thermal neutrons on the induction of murine endogenous viruses from a mouse fibroblast cell line was investigated. Thermal neutrons were more effective than X-rays in induction of endogenous virus as well as in killing of the cells. However, when measured as a function of cell killing, both radiations had similar efficiency of induction. The RBEs of thermal neutrons alone were calculated on the assumption that the contribution of contaminating γ-rays was additive. It was 4.2 for the killing effect and 4-5 for virus induction. (author)

  19. Selected topics in thermal and resonance neutron capture

    International Nuclear Information System (INIS)

    Raman, S.

    1981-01-01

    Several topics of current interest are discussed including energy and intensity standards, direct thermal neutron capture, primary E2 transitions in (n,γ) reactions, nonstatistical effects in resonance neutron capture, transmission measurements of Sc employed in 2-keV facilities, and tests of Axel-Brink predictions of γ-ray strength functions via average resonance capture

  20. Neutron beam measurement dosimetry

    International Nuclear Information System (INIS)

    Amaro, C.R.

    1995-01-01

    This report describes animal dosimetry studies and phantom measurements. During 1994, 12 dogs were irradiated at BMRR as part of a 4 fraction dose tolerance study. The animals were first infused with BSH and irradiated daily for 4 consecutive days. BNL irradiated 2 beagles as part of their dose tolerance study using BPA fructose. In addition, a dog at WSU was irradiated at BMRR after an infusion of BPA fructose. During 1994, the INEL BNCT dosimetry team measured neutron flux and gamma dose profiles in two phantoms exposed to the epithermal neutron beam at the BMRR. These measurements were performed as a preparatory step to the commencement of human clinical trials in progress at the BMRR

  1. Measuring neutron spectra in radiotherapy using the nested neutron spectrometer.

    Science.gov (United States)

    Maglieri, Robert; Licea, Angel; Evans, Michael; Seuntjens, Jan; Kildea, John

    2015-11-01

    Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation-maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors' measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. The NNS may be used to reliably measure the neutron

  2. Measuring neutron spectra in radiotherapy using the nested neutron spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Maglieri, Robert, E-mail: robert.maglieri@mail.mcgill.ca; Evans, Michael; Seuntjens, Jan; Kildea, John [Medical Physics Unit, McGill University, Montreal, Quebec H4A 3J1 (Canada); Licea, Angel [Canadian Nuclear Safety Commission, Ottawa, Ontario K1P 5S9 (Canada)

    2015-11-15

    Purpose: Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. Methods: The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation–maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. Results: The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors’ measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. Conclusions: The NNS may

  3. Energy corrections in pulsed neutron measurements for cylindrical geometry

    International Nuclear Information System (INIS)

    Drozdowicz, K.; Woznicka, U.

    1982-01-01

    A solution of the thermal neutron diffusion equation for a two-region concentric cylindrical system, with a constant neutron flux in the inner medium assumed, is given. The velocity-averaged dynamic parameters for thermal neutrons are used in the method. The corrections due to the diffusion cooling are introduced into the dynamic material buckling and into the velocity distribution of the thermal neutron flux. Detailed relations obtained for a hydrogenous moderator are given. Results of the measurements of the thermal neutron macroscopic absorption cross-sections for the samples in the two-region cylindrical systems are presented. (author)

  4. Thermal neutron detectors based on complex oxide crystals

    CERN Document Server

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  6. The determination of thermal neutron cross section of 81Br

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Tzika, F.; Stamatelatos, I.E.

    2004-01-01

    Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

  8. Welding residual stress estimation by inherent strain analysis and thermal elastic plastic analysis and its verification using neutron diffraction measurement

    International Nuclear Information System (INIS)

    Mochizuki, Masahito; Hayashi, Makoto; Hattori, Toshio

    1997-01-01

    The internal residual stress distribution that resulted from welding a 4-inch-diameter carbon-steel pipe butt-joint was evaluated using several methods and the results were compared. The analytical evaluation methods used were inherent strain analysis and thermal elastic plastic analysis, and the experimental methods were X-ray diffraction and strain gauge for the surface residual stress, and neutron diffraction for the internal stress. The residual stress distributions determined using the various methods agreed well with each other, both for surface stress and internal stress. The characteristics of the evaluation methods were summarized and it was found the most suitable method for any particular situation can be selected by considering the evaluated location and the operating conditions of the object to be evaluated. (author)

  9. 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

  10. 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

  11. Project ''PHANTOM'' - measurement of the absorbed dose, the ''averaged LET'' and the thermal neutron fluence in a tissue equivalent Phantom onboard space station MIR

    International Nuclear Information System (INIS)

    Berger, T.; Hajek, M.; Vana, N.; Schoener, W.; Akatov, Y.; Shurshakov, V.; Arkhangelsky, V.

    2001-01-01

    A water filled phantom with a diameter of 35 cm was developed at the Institute for Biomedical Problems. This tissue equivalent phantom is equipped with 4 channels to deposit dosemeters in different depths. In the framework of the project 'PHANTOM' (phase 1-3) thermoluminescent dosemeters of the commercially available Types TLD - 600 and TLD - 700 were exposed from May 1997 to February 1999 for an overall of 572 days in the different channels (perpendicular and normal to the hull of the spacecraft) of the phantom. The phantom was positioned in the commander cabin, the board engineer cabin and in the module KWANT 2. Besides the measurement of the depth dose distribution, the 'averaged LET' was determined using the HTR - method. The HTR - method utilizes the different LET - efficiencies of the main - and the high temperature glow peaks of LiF dosemeters for the evaluation of the 'averaged LET' in mixed radiation fields. Therefore it is possible to calculate the depth distribution of the biologically relevant dose equivalent. The results show, that despite of the depth decrease of the absorbed dose, the depth dose equivalent is almost constant. This can be explained by the production of secondary particles inside the phantom. The flux of the thermal neutrons was determined using the pair method. Calibration was performed at the research reactor of the Austrian Universities. First results show, that the contribution of thermal neutrons are roughly 10% of the total neutron dose equivalent. (orig.) [de

  12. Neutronic measurements of radioactive waste

    International Nuclear Information System (INIS)

    Perot, B.

    1997-01-01

    This document presents the general matters involved in the radioactive waste management and the different non destructive assays of radioactivity. The neutronic measurements used in the characterization of waste drums containing emitters are described with more details, especially the active neutronic interrogation assays with prompt or delayed neutron detection: physical principle, signal processing and evaluation of the detection limit. (author)

  13. Measurement of the thermal cross section of the reaction 64Zn(n,γ)65Zn in a mixed neutron flux

    International Nuclear Information System (INIS)

    Dorval, E. L; Arribere, M. A; Ribeiro Guevara, S

    2006-01-01

    Zinc is an element that is present in a great variety of biological and geological samples.For its determination by Instrumental Neutron Activation Analysis, the reaction 64 Z n(n,γ) 6 5Zn is used, due to the long half life of the reaction product, the target's bigger isotopic abundance, and an easily measurable 1115 keV main gamma line.In a recent evaluation of thermal cross sections and resonance integrals, a thermal cross section value that is 44.7% bigger than the previous evaluation has been published by the same reference.This difference is not within reported uncertainties.Besides, the relative uncertainty of the new evaluation is much bigger than the one corresponding to the previous evaluation.The adoption of the thermal cross section corresponding to different evaluations may imply, in the case of an irradiation in the peripheral I 6 position at the R A-6 reactor, discrepancies of about 43% in the calculated concentrations.These inconsistencies were evident during the irradiation of certified standard materials.This motivated the measurement of the thermal cross section for the reaction 64 Z n(n,γ) 6 5Zn in the I 6 position at the R A-6 reactor.For the analysis of results, a code was written in order to calculate correction factors from an accurate characterization of the neutron spectrum.The thermal cross section value of the reaction 64 Z n(n,γ) 6 5Zn measured is (0.76± 0.03) b [es

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  15. Study and development of new dosemeters for thermal neutrons

    International Nuclear Information System (INIS)

    Urena N, F.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. First In-Core Simultaneous Measurements of Nuclear Heating and Thermal Neutron Flux Obtained With the Innovative Mobile Calorimeter CALMOS Inside the OSIRIS Reactor

    Science.gov (United States)

    Carcreff, Hubert; Salmon, Laurent; Bubendorff, Jacques; Lepeltier, Valérie

    2016-10-01

    Nuclear heating inside a MTR reactor has to be known in order to design and run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Calorimeter working modes, measurement procedures, main modeling and experimental results and expected advantages of this new technique have been already presented in previous papers. However, these first in-core measurements were not performed beyond 6 W · g-1, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 1014 n · cm-2 · s-1 and nuclear heating up to 12 W · g-1. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a comparison is made between results obtained by the probe calibration coefficient and the zero methods. Thermal neutron flux evaluation from SPND signal processing required a specific TRIPOLI-4 Monte Carlo calculation which has been performed with the precise CALMOS cell geometry. In addition, the Finite Element model for temperatures map prediction inside the calorimetric cell has been upgraded with recent experimental data obtained up to 12 W · g-1. Finally, the experience feedback led us to improvement perspectives. A second device is

  18. Compound Refractive Lenses for Thermal Neutron Applications

    Energy Technology Data Exchange (ETDEWEB)

    Gary, Charles K.

    2013-11-12

    This project designed and built compound refractive lenses (CRLs) that are able to focus, collimate and image using thermal neutrons. Neutrons are difficult to manipulate compared to visible light or even x rays; however, CRLs can provide a powerful tool for focusing, collimating and imaging neutrons. Previous neutron CRLs were limited to long focal lengths, small fields of view and poor resolution due to the materials available and manufacturing techniques. By demonstrating a fabrication method that can produce accurate, small features, we have already dramatically improved the focal length of thermal neutron CRLs, and the manufacture of Fresnel lens CRLs that greatly increases the collection area, and thus efficiency, of neutron CRLs. Unlike a single lens, a compound lens is a row of N lenslets that combine to produce an N-fold increase in the refraction of neutrons. While CRLs can be made from a variety of materials, we have chosen to mold Teflon lenses. Teflon has excellent neutron refraction, yet can be molded into nearly arbitrary shapes. We designed, fabricated and tested Teflon CRLs for neutrons. We demonstrated imaging at wavelengths as short as 1.26 ? with large fields of view and achieved resolution finer than 250 μm which is better than has been previously shown. We have also determined designs for Fresnel CRLs that will greatly improve performance.

  19. Determination of thermal neutrons diffusion length in graphite

    International Nuclear Information System (INIS)

    Garcia Fite, J.

    1959-01-01

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

  20. Accuracy of the thermal neutron absorption cross section measurements (based on examples of selected pulsed beam methods); Dokladnosc pomiarow przekroju czynnego absorpcji neutronow termicznych (na przykladzie wybranych metod impulsowych)

    Energy Technology Data Exchange (ETDEWEB)

    Krynicka, E. [The H. Niewodniczanski Inst. of Nuclear Physics, Cracow (Poland)

    1997-12-31

    The problem of accuracy of the thermal neutron macroscopic absorption cross section determination is discussed on examples of selected measurement methods which use non-stationary neutron fields. The computer simulation method elaborated by the author is presented as a procedure for estimating the standard deviation of the measured absorption cross section. The computer simulation method presented can be easily utilized to estimate the accuracy of measurement of various physical magnitudes. (author) 46 refs, 3 figs, 1 tab

  1. 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

  2. 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

  3. Dosimetric characteristics of the thermal neutron beam facility for neutron capture therapy at Hanaro reactor

    International Nuclear Information System (INIS)

    Lee, Dong Han; Suh, Soheigh; Ji, Young Hoon

    2006-01-01

    The thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the Hanaro, 30 MW multi-purpose research reactor. In order to determine the different dose components in phantoms irradiated with a mixed thermal neutron beam and gamma-ray for clinical applications, various techniques were applied including the use of activation foils, TLDs and ionization chambers. The water phantom was utilized in the measurement. The results of the measurement were compared with MCNP4B calculations. The thermal neutron fluxes were 1.02E9 and 6.07E8/cm 2 ·s at 10 and 20 mm depth in water, respectively. The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water. The result of this study can be used as basic data for subsequent BNCT clinical application. (author)

  4. Reactor neutron flux measuring device

    International Nuclear Information System (INIS)

    Okutani, Yasushi; Hayakawa, Toshifumi.

    1994-01-01

    The present invention concerns a device for displaying an approximate neutron flux distribution to recognize the neutron flux distribution of the whole reactor in a short period of time. The device of the present invention displays, the results of measurement for neutron fluxes collected by a data collecting section on every results of the measurements at measuring points situating at horizontally identical positions of the reactor core. In addition, every results of the measurements at the measuring points situating at the identical height in the reactor core are accumulated, and the results of the integration are graphically displayed. With such procedures, the neutron flux distribution in the entire reactor is approximately displayed. Existent devices could not recognize the neutron flux distribution of the entire reactor at a glance and it took much time for the recognition. The device of the present invention can recognize the neutron flux distribution of the entire reactor in a short period of time. (I.S.)

  5. The review of the application of neutron radiography to thermal hydraulic research

    CERN Document Server

    Mishima, K; Saitô, Y; Nakamura, H; Matsubayashi, M

    1999-01-01

    This paper is concerned with the establishment of thermal neutron radiography as a high accuracy measurement method. This paper reviews the present status on the development of high-frame-rate neutron radiography with a steady thermal neutron beam and its application to multiphase flow research performed at the Research Reactor Institute of Kyoto University in collaboration with the Japan Atomic Energy Research Institute.

  6. More accurate thermal neutron coincidence counting technique

    International Nuclear Information System (INIS)

    Baron, N.

    1978-01-01

    Using passive thermal neutron coincidence counting techniques, the accuracy of nondestructive assays of fertile material can be improved significantly using a two-ring detector. It was shown how the use of a function of the coincidence count rate ring-ratio can provide a detector response rate that is independent of variations in neutron detection efficiency caused by varying sample moderation. Furthermore, the correction for multiplication caused by SF- and (α,n)-neutrons is shown to be separable into the product of a function of the effective mass of 240 Pu (plutonium correction) and a function of the (α,n) reaction probability (matrix correction). The matrix correction is described by a function of the singles count rate ring-ratio. This correction factor is empirically observed to be identical for any combination of PuO 2 powder and matrix materials SiO 2 and MgO because of the similar relation of the (α,n)-Q value and (α,n)-reaction cross section among these matrix nuclei. However the matrix correction expression is expected to be different for matrix materials such as Na, Al, and/or Li. Nevertheless, it should be recognized that for comparison measurements among samples of similar matrix content, it is expected that some function of the singles count rate ring-ratio can be defined to account for variations in the matrix correction due to differences in the intimacy of mixture among the samples. Furthermore the magnitude of this singles count rate ring-ratio serves to identify the contaminant generating the (α,n)-neutrons. Such information is useful in process control

  7. Neutron Thermalization and Reactor Spectra. Vol. I. 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. (author)

  8. 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.

  9. Neutron physics

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Wirtz, K.

    1974-01-01

    This textbook consists of four sections which deal with the following subjects: 1. Production of neutrons and their interactions with the nuclei; neutron sources; neutron detectors; cross-section measurements. 2. Theory of neutron interactions with macroscopic media; neutron slowing down; space distribution of moderated neutrons; neutron thermalization; neutron scattering. 3. Radioactive probe measurements of thermal neutron fluxes; activation by means of epithermal neutrons; threshold detectors of fast neutrons; neutron calibration. 4. Neutron energy; slowing down kernels; neutron age; diffusion length and absorption of neutrons

  10. Thermal neutrons streaming in straight duct

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  11. Effect of thermal neutron irradiation on personal photon dosemeters

    International Nuclear Information System (INIS)

    Alberts, W.G.; Kluge, H.

    1984-01-01

    Since radiation protection dosemeters are often in use in mixed neutron-gamma fields it appeared desirable to the Commission of the European Communities to include in the intercomparison study a small measuring programme for investigating the influence of slow neutrons on personal photon dosemeters. Film dosemeters and TLD badges were sent to the PTB to be irradiated at the Thermal Neutron Reference Beam of the Research and Measuring Reactor Braunschweig, and were then sent back for evaluation. Institutes participating in this particular programme were: CEA, Fontenay-aux-Roses, CEGB, Berkeley, CNEN, Bologna, GSF, Neuherberg, TNO, Arnhem. The results for film dosemeters obtained in this intercomparison show that a careful calibration of every type of dosemeter with thermal neutrons, preferentially with low photon contamination, is desirable to achieve the appropriate corrections for the assessment of photon equivalent in a mixed field

  12. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1992-09-01

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

  13. Development of An Epi-thermal Neutron Field for Fundamental Researches for BNCT with A DT Neutron Source

    Directory of Open Access Journals (Sweden)

    Osawa Yuta

    2017-01-01

    Full Text Available Boron Neutron Capture Therapy (BNCT is known to be a new promising cancer therapy suppressing influence against normal cells. In Japan, Accelerator Based Neutron Sources (ABNS are being developed for BNCT. For the spread of ABNS based BNCT, we should characterize the neutron field beforehand. For this purpose, we have been developing a low-energy neutron spectrometer based on 3He position sensitive proportional counter. In this study, a new intense epi-thermal neutron field was developed with a DT neutron source for verification of validity of the spectrometer. After the development, the neutron field characteristics were experimentally evaluated by using activation foils. As a result, we confirmed that an epi-thermal neutron field was successfully developed suppressing fast neutrons substantially. Thereafter, the neutron spectrometer was verified experimentally. In the verification, although a measured detection depth distribution agreed well with the calculated distribution by MCNP, the unfolded spectrum was significantly different from the calculated neutron spectrum due to contribution of the side neutron incidence. Therefore, we designed a new neutron collimator consisting of a polyethylene pre-collimator and boron carbide neutron absorber and confirmed numerically that it could suppress the side incident neutrons and shape the neutron flux to be like a pencil beam.

  14. Gamma discrimination in pillar structured thermal neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Q; Radev, R P; Conway, A M; Voss, L F; Wang, T F; Nikolic, R J; Deo, N; Cheung, C L

    2012-03-26

    Solid-state thermal neutron detectors are desired to replace {sup 3}He tube based technology for the detection of special nuclear materials. {sup 3}He tubes have some issues with stability, sensitivity to microphonics and very recently, a shortage of {sup 3}He. There are numerous solid-state approaches being investigated that utilize various architectures and material combinations. By using the combination of high-aspect-ratio silicon PIN pillars, which are 2 {micro}m wide with a 2 {micro}m separation, arranged in a square matrix, and surrounded by {sup 10}B, the neutron converter material, a high efficiency thermal neutron detector is possible. Besides intrinsic neutron detection efficiency, neutron to gamma discrimination is an important figure of merit for unambiguous signal identification. In this work, theoretical calculations and experimental measurements are conducted to determine the effect of structure design of pillar structured thermal neutron detectors including: intrinsic layer thickness, pillar height, substrate doping and incident gamma energy on neutron to gamma discrimination.

  15. Neutronics methods for thermal radiative transfer

    International Nuclear Information System (INIS)

    Larsen, E.W.

    1988-01-01

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

  16. Impact of neutron irradiation on thermal helium desorption from iron

    Science.gov (United States)

    Hu, Xunxiang; Field, Kevin G.; Taller, Stephen; Katoh, Yutai; Wirth, Brian D.

    2017-06-01

    The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 1018 He/m2. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

  17. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1990-03-01

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

  18. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.N.

    1990-01-01

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

  19. 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.)

  20. Neutron-emission measurements at a white neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Haight, Robert C [Los Alamos National Laboratory

    2010-01-01

    Data on the spectrum of neutrons emittcd from neutron-induced reactions are important in basic nuclear physics and in applications. Our program studies neutron emission from inelastic scattering as well as fission neutron spectra. A ''white'' neutron source (continuous in energy) allows measurements over a wide range of neutron energies all in one experiment. We use the tast neutron source at the Los Alamos Neutron Science Center for incident neutron energies from 0.5 MeV to 200 MeV These experiments are based on double time-of-flight techniques to determine the energies of the incident and emitted neutrons. For the fission neutron measurements, parallel-plate ionization or avalanche detectors identify fission in actinide samples and give the required fast timing pulse. For inelastic scattering, gamma-ray detectors provide the timing and energy spectroscopy. A large neutron-detector array detects the emitted neutrons. Time-of-flight techniques are used to measure the energies of both the incident and emitted neutrons. Design considerations for the array include neutron-gamma discrimination, neutron energy resolution, angular coverage, segmentation, detector efficiency calibration and data acquisition. We have made preliminary measurements of the fission neutron spectra from {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu. Neutron emission spectra from inelastic scattering on iron and nickel have also been investigated. The results obtained will be compared with evaluated data.

  1. Spectral distribution measurements of neutrons in paraffin borax mixtures

    International Nuclear Information System (INIS)

    El-Khatib, A.M.; Gaber, M.; Abou El-Khier, M.A.

    1987-01-01

    Neutron fluxes from a compact D-T neutron source has been measured in paraffin-borax mixtures by using activation foil detectors with successive threshold energies. The absorbed doses, backscattering coefficients and build-up factors were determined as well. The contribution of thermal and intermediate neutron dose is much lower, compared to that of fast neutrons. Among the used mediums, paraffin loaded with 4% borax concentration was found to be the best absorbing medium against neutrons at near depths within the blocks, while at a depth around 12 cm the neutron absorption (or scattering) is independent on the type of the used medium. (author)

  2. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  3. Neutron ion temperature measurement

    International Nuclear Information System (INIS)

    Strachan, J.D.; Hendel, H.W.; Lovberg, J.; Nieschmidt, E.B.

    1986-11-01

    One important use of fusion product diagnostics is in the determination of the deuterium ion temperature from the magnitude of the 2.5 MeV d(d,n) 3 He neutron emission. The detectors, calibration methods, and limitations of this technique are reviewed here with emphasis on procedures used at PPPL. In most tokamaks, the ion temperature deduced from neutrons is in reasonable agreement with the ion temperature deduced by other techniques

  4. Thermal neutron calibration channel at LNMRI/IRD

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. Measurement of 237Np fission rate ratio relative to 235U fission rate in cores with various thermal neutron spectrum at the Kyoto University Critical Assembly

    International Nuclear Information System (INIS)

    Unesaki, Hironobu; Shiroya, Seiji; Iwasaki, Tomohiko; Fujiwara, Daisuke; Kitada, Takanori; Kuroda, Mitsuo; Kohashi, Akio; Kato, Takeshi; Ikeuchi, Yoshitaka

    2000-01-01

    Integral measurements of 237 Np fission rate ratio relative to 235 U fission rate have been performed at Kyoto University Citrical Assembly. The fission rates have been measured using the back-to back type double fission chamber at five thermal cores with different H/ 235 U ratio so that the neutron spectra of the cores were systematically varied. The measured fission rate ratio per atom was 0.00439 to 0.0298, with a typical uncertainty of 2 to 3%. The measured data were compared with the calculated results using SRAC/TWOTRAN and MVP based on JENDL-3.2, which gave the averaged C/E values of 0.93 and 0.95, respectively. Obtained results of C/E using 237 Np cross sections from JENDL-3/2, ENDF/B-VI.5 and JEF2.2 show that the latter two gave smaller results than JENDL-3.2 by about 4%, which clearly reflects the discrepancy in the evaluated cross section among the libraries. This difference arises from both fast fission and resonance region. Although further improvement is recommended, 237 Np fission cross section in JENDL-3.2 is considered to be superior to those in the other libraries and can be adopted for use in design calculations for minor actinide transmutation system using thermal reactors with prediction precision of 237 Np fission rate with in 10%. (author)

  6. The computer card Accuspec FMS-Canberra in the relaxation time of thermal neutron flux pulses measuring system; Karta komputerowa Accuspec FMS-Canberra w systemie pomiaru stalej zaniku impulsowego strumienia neutronow termicznych

    Energy Technology Data Exchange (ETDEWEB)

    Igielski, A.; Burda, J.; Gabanska, B.; Krynicka, E. [The H. Niewodniczanski Inst. of Nuclear Physics, Cracow (Poland)

    1997-12-31

    In this paper the theoretical basis and a measuring method of the relaxation time of pulsed thermal neutron flux is described. The developed measuring system based on the Accuspec FMS computer card of Canberra is presented. The methods of testing and test results are also described 8 refs, 5 figs, 3 tabs

  7. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Directory of Open Access Journals (Sweden)

    Hu J.-P.

    2016-01-01

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

  8. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Science.gov (United States)

    Hu, J.-P.; Holden, N. E.; Reciniello, R. N.

    2016-02-01

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

  9. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-23

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

  10. Atomic structure holography using thermal neutrons

    International Nuclear Information System (INIS)

    Sur, B.; Rogge, R.B.; Hammond, R.P.; Anghel, V.N.P.; Katsaras, J.

    2001-01-01

    The idea of atomic-resolution holography has its roots in the X-ray work of Bragg and in Gabor's electron interference microscope. Gabor's lensless microscope was not realized in his time, but over the past twelve years there has been a steady increase in the number of reports on atomic-resolution holography. All of this work involves the use of electrons or hard X-rays to produce the hologram. Neutrons are often unique among scattering probes in their interaction with materials: for example, the relative visibility of hydrogen and its isotopes is a great advantage in the study of polymers and biologically relevant materials. Recent work proposed that atomic-resolution holography could be achieved with thermal neutrons. Here we use monochromatic thermal neutrons, adopting the inside-source concept of Szoke, to image planes of oxygen atoms located above and below a single hydrogen atom in the oxide mineral Simpsonite. (author)

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

    International Nuclear Information System (INIS)

    Brugger, R.M.

    1975-12-01

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

  12. Thermal annealing in neutron-irradiated tribromobenzenes

    DEFF Research Database (Denmark)

    Siekierska, K.E.; Halpern, A.; Maddock, A. G.

    1968-01-01

    The distribution of 82Br among various products in neutron-irradiated isomers of tribromobenzene has been investigated, and the effect of thermal annealing examined. Reversed-phase partition chromatography was employed for the determination of radioactive organic products, and atomic bromine...

  13. Thermal Neutron Capture onto the Stable Tungsten Isotopes

    Directory of Open Access Journals (Sweden)

    Nichols A.

    2012-02-01

    Full Text Available Thermal neutron-capture measurements of the stable tungsten isotopes have been carried out using the guided thermal-neutron beam at the Budapest Reactor. Prompt singles spectra were collected and analyzed using the HYPERMET γ-ray analysis software package for the compound tungsten systems 183W, 184W, and 187W, prepared from isotopically-enriched samples of 182W, 183W, and 186W, respectively. These new data provide both confirmation and new insights into the decay schemes and structure of the tungsten isotopes reported in the Evaluated Gamma-ray Activation File based upon previous elemental analysis. The experimental data have also been compared to Monte Carlo simulations of γ-ray emission following the thermal neutron-capture process using the statistical-decay code DICEBOX. Together, the experimental cross sections and modeledfeeding contribution from the quasi continuum, have been used to determine the total radiative thermal neutron-capture cross sections for the tungsten isotopes and provide improved decay-scheme information for the structural- and neutron-data libraries.

  14. Measurement of the thermal neutron capture cross section and the resonance integral of the sup 1 sup 0 sup 9 Ag(n, gamma) sup 1 sup 1 sup 0 sup m Ag reaction

    CERN Document Server

    Nakamura, S; Shcherbakov, O A; Furutaka, K; Harada, H; Katoh, T

    2003-01-01

    In order to develop a neutron flux monitor for long-term neutron irradiation, the thermal neutron (2,200 m/s neutron) capture cross section (sigma sub 0) and the resonance integral (I sub 0) of the sup 1 sup 0 sup 9 Ag(n, gamma) sup 1 sup 1 sup 0 sup m Ag reaction were measured by the activation and gamma-ray spectroscopic methods. Silver foils were irradiated with and without a Cd shield capsule at the Rikkyo Research Reactor. The Co/Al and Au/Al alloy wires were irradiated together with silver foils in order to monitor the thermal neutron flux and the fraction of the epi-thermal neutron part (Westcott's index). A high purity Ge detector was used for the gamma-ray measurements of the irradiated samples. The sigma sub 0 and the I sub 0 of the sup 1 sup 0 sup 9 Ag(n, gamma) sup 1 sup 1 sup 0 sup m Ag reaction are 4.12+-0.10 b and 67.9+-3.1 b, respectively. The sigma sub 0 is 12% smaller than the tabulated one (4.7+-0.2 b). On the other hand, the I sub 0 is in agreement with the tabulated one (72.3+-4.0 b) with...

  15. a Study of Prompt Neutron Emission in Thermal Neutron-Induced Fission of URANIUM-235.

    Science.gov (United States)

    Franklyn, Christopher Barry

    An original experiment was performed to measure the angular correlation of fission neutrons from thermal -neutron-induced fission of ('235)U, with respect to the light fission fragment direction, as a function of fragment mass division and neutron energy. A Monte Carlo model, with a realistic description of the fission fragment de -excitation process, was developed to simulate the observed neutron-fragment angular correlation data. The model was capable of investigating various possible forms of neutron emission which were classified into emission before, during and after full fragment acceleration, and correspondingly named scission acceleration and prompt neutron emission. Simulated neutron-fragment angular correlations displaying similar distributions with respect to the light fragment direction for different forms of neutron emission are shown to exhibit differing distributions when examined as a function of fragment mass division or neutron energy, thus illustrating the sensitivity of the experiment to the forms of neutron emission occurring in fission. A primary conclusion of the investigation was that neutron emission solely from fully accelerated fragments, whether isotropically or anisotropically emitted in the fragment centre of mass system, was unable to adequately describe the observed neutron-fragment angular correlations. Simulation of the fission process with some neutron emission before or during fragment acceleration exhibited a closer correspondence with observed phenomena. Within the scope of this work the form of neutron emission that produced the closest overall correspondence with experimental data was a simulation in which 20% of the emitted neutrons were isotropically emitted scission neutrons with a Maxwellian energy distribution of temperature 1.0 MeV. The remaining neutrons were emitted from fully accelerated fragments, being isotropic in the fragment centre of mass frame, except for the n-th(n > 1) neutrons from the light fragment, which

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

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre Luis Ferreira

    1995-12-31

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

  17. THERMAL NEUTRON INTENSITIES IN SOILS IRRADIATED BY FAST NEUTRONS FROM POINT SOURCES. (R825549C054)

    Science.gov (United States)

    Thermal-neutron fluences in soil are reported for selected fast-neutron sources, selected soil types, and selected irradiation geometries. Sources include 14 MeV neutrons from accelerators, neutrons from spontaneously fissioning 252Cf, and neutrons produced from alp...

  18. Absolute calibration of a cold and thermal neutron detector using monochromatic neutron beam

    Science.gov (United States)

    Choi, Jin Ha; Cude-Woods, Christopher; Ito, Takeyasu; Young, Albert

    2017-09-01

    Time of flight spectra for cold neutrons exiting the moderator volume of the LANSCE UCN source has been obtained using a commercial neutron scintillator, EJ-426, coupled to a Hamamatsu R1355. The absolute efficiency for this detector system was determined using a 37.4 meV (monochromatic) neutron beam from the Neutron Powder Diffraction Facility (NPDF) at North Carolina State University's PULSTAR reactor. We measured the absolute neutron flux at the NPDF through thin foil activation and explored threshold effects through analysis of the measured pulse height distribution for effectively pure neutron signals from the NPDF beam. Non-uniformity of the flux profile across the detector and the detection efficiency as a function of the point of incidence of neutrons on the scintillator was explored using a X-Y translation system to perform scans using either fixed or movable apertures. The results are generally consistent with our expectations for this system, and provide a quantitative assessment of the sensitivity of this system to cold and thermal neutrons. This project was funded by the National Science Foundation and the Department of Energy.

  19. Distribution of thermal neutron flux around a PET cyclotron.

    Science.gov (United States)

    Ogata, Yoshimune; Ishigure, Nobuhito; Mochizuki, Shingo; Ito, Kengo; Hatano, Kentaro; Abe, Junichiro; Miyahara, Hiroshi; Masumoto, Kazuyoshi; Nakamura, Hajime

    2011-05-01

    The number of positron emission tomography (PET) examinations has greatly increased world-wide. Since positron emission nuclides for the PET examinations have short half-lives, they are mainly produced using on-site cyclotrons. During the production of the nuclides, significant quantities of neutrons are generated from the cyclotrons. Neutrons have potential to activate the materials around the cyclotrons and cause exposure to the staff. To investigate quantities and distribution of the thermal neutrons, thermal neutron fluxes were measured around a PET cyclotron in a laboratory associating with a hospital. The cyclotron accelerates protons up to 18 MeV, and the mean particle current is 20 μA. The neutron fluxes were measured during both 18F production and C production. Gold foils and thermoluminescent dosimeter (TLD) badges were used to measure the neutron fluxes. The neutron fluxes in the target box averaged 9.3 × 10(6) cm(-2) s(-1) and 1.7 × 10(6) cm(-2) s(-1) during 18F and 11C production, respectively. Those in the cyclotron room averaged 4.1 × 10(5) cm(-2) s(-1) and 1.2 × 10(5) cm(-2) s(-1), respectively. Those outside the concrete wall shielding were estimated as being equal to or less than ∼3 cm s, which corresponded to 0.1 μSv h(-1) in effective dose. The neutron fluxes outside the concrete shielding were confirmed to be quite low compared to the legal limit.

  20. Measuring the Neutron Lifetime with Magnetically Trapped Ultracold Neutrons

    Science.gov (United States)

    Mumm, H. P.; Huber, M. G.; Yue, A. T.; Thompson, A. K.; Dewey, M. S.; Huffer, C. R.; Huffman, P. R.; Schelhammer, K. W.; O'Shaughnessy, C.; Coakley, K. J.

    2014-03-01

    We describe an experiment to measure the neutron lifetime using a technique with a set of systematic uncertainties largely different than those of previous measurements. In this approach, ultracold neutrons (UCN) are produced by inelastic scattering of cold (0.89 nm) neutrons in a reservoir of superfluid 4He. These neutrons are then confined using a three-dimensional magnetic trap. As the trapped neutrons beta decay, the energetic electrons produced in the decay generate scintillations in the liquid He; each decay is detectable with nearly 100 % efficiency. The neutron lifetime can be directly determined by measuring the scintillation rate as a function of time.

  1. Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson

    1996-01-01

    A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

  2. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1993-01-01

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

  3. THERMAL: A routine designed to calculate neutron thermal scattering

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1995-01-01

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

  4. Neutron energy measurement for practical applications

    Indian Academy of Sciences (India)

    M V Roshan

    2018-02-07

    . Elastic scattering of monoenergetic α-particles from neutron collision enables neutron energy measurement by calculating the amount of deviation from the position where collision takes place. The neutron numbers with ...

  5. Passive neutron-multiplication measurements

    International Nuclear Information System (INIS)

    Zolnay, A.S.; Barnett, C.S.; Spracklen, H.P.

    1982-01-01

    We have developed an instrument to measure neutron multiplication by statistical analysis of the timing of neutrons emitted from fissionable material. This instrument is capable of repeated analysis of the same recorded data with selected algorithms, graphical displays showing statistical properties of the data, and preservation of raw data on disk for future comparisons. In our measurements we have made a comparison of the covariance to mean and Feynman variance to mean analysis algorithms to show that the covariance avoids a bias term and measures directly the effect due to the presence of neutron chains. A spherical assembly of enriched uranium shells and acrylic resin reflector/moderator components used for the measurements is described. Preliminary experimental results of the Feynman variance to mean measurements show the expected correlation with assembly multiplication

  6. Stereoscopic radiographic images with thermal neutrons

    Science.gov (United States)

    Silvani, M. I.; Almeida, G. L.; Rogers, J. D.; Lopes, R. T.

    2011-10-01

    Spatial structure of an object can be perceived by the stereoscopic vision provided by eyes or by the parallax produced by movement of the object with regard to the observer. For an opaque object, a technique to render it transparent should be used, in order to make visible the spatial distribution of its inner structure, for any of the two approaches used. In this work, a beam of thermal neutrons at the main port of the Argonauta research reactor of the Instituto de Engenharia Nuclear in Rio de Janeiro/Brazil has been used as radiation to render the inspected objects partially transparent. A neutron sensitive Imaging Plate has been employed as a detector and after exposure it has been developed by a reader using a 0.5 μm laser beam, which defines the finest achievable spatial resolution of the acquired digital image. This image, a radiographic attenuation map of the object, does not represent any specific cross-section but a convoluted projection for each specific attitude of the object with regard to the detector. After taking two of these projections at different object attitudes, they are properly processed and the final image is viewed by a red and green eyeglass. For monochromatic images this processing involves transformation of black and white radiographies into red and white and green and white ones, which are afterwards merged to yield a single image. All the processes are carried out with the software ImageJ. Divergence of the neutron beam unfortunately spoils both spatial and contrast resolutions, which become poorer as object-detector distance increases. Therefore, in order to evaluate the range of spatial resolution corresponding to the 3D image being observed, a curve expressing spatial resolution against object-detector gap has been deduced from the Modulation Transfer Functions experimentally. Typical exposure times, under a reactor power of 170 W, were 6 min for both quantitative and qualitative measurements. In spite of its intrinsic constraints

  7. Thermal expansion study of simulated DUPIC fuel using neutron diffraction

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Ryu, H. J.; Bae, J. H.; Kim, H. S.; Song, K. C.; Yang, M. S.; Choi, Y. N.; Han, Y. S.; Oh, H. S.

    2001-07-01

    The lattice parameters of simulated DUPIC fuel and UO2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO2 and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO2. For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO2 and simulated DUPIC fuel are 10.471 ''10-6 and 10.751 ''10-6 K-1, respectively

  8. An improved fast neutron radiography quantitative measurement method

    International Nuclear Information System (INIS)

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

    2004-01-01

    The validity of a fast neutron radiography quantification method, the Σ-scaling method, which was originally proposed for thermal neutron radiography was examined with Monte Carlo calculations and experiments conducted at the YAYOI fast neutron source reactor. Water and copper were selected as comparative samples for a thermal neutron radiography case and a dense object, respectively. Although different characteristics on effective macroscopic cross-sections were implied by the simulation, the Σ-scaled experimental results with the fission neutron spectrum cross-sections were well fitted to the measurements for both the water and copper samples. This indicates that the Σ-scaling method could be successfully adopted for quantitative measurements in fast neutron radiography

  9. Application of thermal neutrons in testing of nuclear facilities materials

    International Nuclear Information System (INIS)

    Milczarek, J.J.

    2008-01-01

    Recent applications of thermal neutrons in testing of materials used in nuclear facilities are presented. The neutron radiography technique, characterization of residual stresses with neutron diffraction and small angle neutron scattering is considered in some detail. The results of testing of fuel elements, steel used for pressurized reactor vessels and changes induced in control rods are discussed. (author)

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

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

    1999-01-01

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

  11. Development of temperature related thermal neutron scattering database for MCNP

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  12. Neutron fluence in antiproton radiotherapy, measurements and simulations

    DEFF Research Database (Denmark)

    Bassler, Niels; Holzscheiter, Michael H.; Petersen, Jørgen B.B.

    2010-01-01

    A significant part of the secondary particle spectrum from antiproton annihilation consists of fast neutrons, which may contribute to a significant dose background found outside the primary beam. Using a polystyrene phantom as a moderator, we have performed absolute measurements of the thermalized...... part of the fast neutron spectrum using Lithium-6 and -7 Fluoride TLD pairs. The experimental results are found to be in good agreement with simulations using the Monte Carlo particle transport code FLUKA. The thermal neutron kerma resulting from the measured thermal neutron fluence is insignificant...... compared to the contribution from fast neutrons. The results are found to be similar to values calculated for pion treatment, however exact modeling under more realistic treatment scenarios is still required to quantitatively compare these treatment modalities....

  13. Measurement of actinide neutron cross sections

    International Nuclear Information System (INIS)

    Firestone, Richard B.; Nitsche, Heino; Leung, Ka-Ngo; Perry, DaleL.; English, Gerald

    2003-01-01

    The maintenance of strong scientific expertise is critical to the U.S. nuclear attribution community. It is particularly important to train students in actinide chemistry and physics. Neutron cross-section data are vital components to strategies for detecting explosives and fissile materials, and these measurements require expertise in chemical separations, actinide target preparation, nuclear spectroscopy, and analytical chemistry. At the University of California, Berkeley and the Lawrence Berkeley National Laboratory we have trained students in actinide chemistry for many years. LBNL is a leader in nuclear data and has published the Table of Isotopes for over 60 years. Recently, LBNL led an international collaboration to measure thermal neutron capture radiative cross sections and prepared the Evaluated Gamma-ray Activation File (EGAF) in collaboration with the IAEA. This file of 35, 000 prompt and delayed gamma ray cross-sections for all elements from Z=1-92 is essential for the neutron interrogation of nuclear materials. LBNL has also developed new, high flux neutron generators and recently opened a 1010 n/s D+D neutron generator experimental facility

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

    CERN Document Server

    Karadag, M; Tan, M; Oezmen, A

    2003-01-01

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

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

    Science.gov (United States)

    Endo, Satoru; Sato, Hitoshi; Shimazaki, Takuto; Nakajima, Erika; Kotani, Kei; Suda, Mitsuru; Hamano, Tsuyoshi; Kajimoto, Tsuyoshi; Tanaka, Kenichi; Hoshi, Masaharu

    2017-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

  17. ATW neutron spectrum measurements at LAMPF

    Energy Technology Data Exchange (ETDEWEB)

    Butler, G.W.; Littleton, P.E.; Morgan, G.L. [Los Alamos National Laboratory, NM (United States)] [and others

    1995-10-01

    Accelerator transmutation of waste (ATW) is a proposal to use a high flux of accelerator-produced thermalized neutrons to transmute both fission product and higher actinide commercial nuclear waste into stable or short-lived radioactive species in order to avoid long-term storage of nuclear waste. At LAMPF the authors recently performed experiments that were designed to measure the spectrum of neutrons produced per incident proton for full-scale proposed ATW targets of lead and lithium. The neutrons produced in such targets have a spectrum of energies that extends up to the energy of the incident proton beam, but the distribution peaks between 1 and 5 MeV. Transmutation reactions and fission of actinides are most efficient when the neutron energy is below a few eV, so the target must be surrounded by a non-absorbing material (blanket) to produce additional neutrons and reduce the energy of high energy neutrons without loss. The experiments with the lead target, 25 cm diameter by 40 cm long, were conducted with 800 MeV protons, while those with the lithium target, 25 cm diameter by 175 cm long, were conducted with 400 MeV protons. The blanket in both sets of experiments was a 60 cm diameter by 200 cm long annulus of lead that surrounded the target. Surrounding the blanket was a steel water tank with dimensions of 250 cm diameter by 300 cm long that simulated the transmutation region. A small sample pipe penetrated the length of the lead blanket and other sample pipes penetrated the length of the water tank at different radii from the beam axis so that the neutron spectra at different locations could be measured by foil activation. After irradiation the activated foil sets were extracted and counted with calibrated high resolution germanium gamma ray detectors at the Los Alamos nuclear chemistry counting facility.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

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

  19. 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)

  20. Thermal Neutron Capture Cross Sections of the PalladiumIsotopes

    Energy Technology Data Exchange (ETDEWEB)

    Firestone, R.B.; Krticka, M.; McNabb, D.P.; Sleaford, B.; Agvaanluvsan, U.; Belgya, T.; Revay, Zs.

    2006-07-17

    Precise gamma-ray thermal neutron capture cross sectionshave been measured at the Budapest Reactor for all elements withZ=1-83,92 except for He and Pm. These measurements and additional datafrom the literature been compiled to generate the Evaluated Gamma-rayActivation File (EGAF), which is disseminated by LBNL and the IAEA. Thesedata are nearly complete for most isotopes with Z<20 so the totalradiative thermal neutron capture cross sections can be determineddirectly from the decay scheme. For light isotopes agreement with therecommended values is generally satisfactory although large discrepanciesexist for 11B, 12,13C, 15N, 28,30Si, 34S, 37Cl, and 40,41K. Neutroncapture decay data for heavier isotopes are typically incomplete due tothe contribution of unresolved continuum transitions so only partialradiative thermal neutron capture cross sections can be determined. Thecontribution of the continuum to theneutron capture decay scheme arisesfrom a large number of unresolved levels and transitions and can becalculated by assuming that the fluctuations in level densities andtransition probabilities are statistical. We have calculated thecontinuum contribution to neutron capture decay for the palladiumisotopes with the Monte Carlo code DICEBOX. These calculations werenormalized to the experimental cross sections deexciting low excitationlevels to determine the total radiative thermal neutron capture crosssection. The resulting palladium cross sections values were determinedwith a precision comparable to the recommended values even when only onegamma-ray cross section was measured. The calculated and experimentallevel feedings could also be compared to determine spin and parityassignments for low-lying levels.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  2. Design of the thermal neutron detection system for CJPL-II

    Science.gov (United States)

    Zeng, Zhao-Ming; Gong, Hui; Li, Jian-Min; Yue, Qian; Zeng, Zhi; Cheng, Jian-Ping

    2017-05-01

    A low background thermal neutron flux detection system has been designed to measure the ambient thermal neutron flux of the second phase of the China Jinping Underground Laboratory (CJPL-II), right after completion of the rock bolting work. A 3He proportional counter tube combined with an identical 4He proportional counter tube was employed as the thermal neutron detector, which has been optimised in energy resolution, wall effect and radioactivity of construction materials for low background performance. The readout electronics were specially designed for long-term stable operation and easy maintenance in an underground laboratory under construction. The system was installed in Lab Hall No. 3 of CJPL-II and accumulated data for about 80 days. The ambient thermal neutron flux was determined under the assumption that the neutron field is fully thermalized, uniform and isotropic at the measurement position. Supported by National Natural Science Foundation of China (11475094)

  3. Dependence of the thermal neutron fluence at the size installations radiotherapy bunker; Dependencia de la fluencia termica de neutrones en el tamano del bunquer en instalaciones de radioterapia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Soto, X.; Amgarou, K.; Langares, J. L.; Exposito, M. R.; Gomez, F.; Domingo, C.; Sanchez-Doblado, F.

    2011-07-01

    The project aims to infer the dose deposited by neutrons in the patient treated by radiation therapy, from a measurement of the thermal neutron fluence at a selected point within the treatment room. These thermal neutrons are created when fast neutrons produced in the linac head are moderate, mainly in the walls of the bunker, and its yield depends on both the volume of the room and its geometry.

  4. Neutron irradiation effect of thermally-sensitized stainless steels

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  5. Neutron background measurements in the underground laboratory of Modane

    International Nuclear Information System (INIS)

    Chazal, V.; Chambon, B.; De Jesus, M.; Drain, D.; Pastor, C.; Vagneron, L.; Brissot, R.; Cavaignac, J.F.; Stutz, A.; Giraud-Heraud, Y.

    1997-07-01

    Measurements of the background neutron environment, at a depth of 1780 m (4800 mWe) in the Underground Laboratory of Modane (L.S.M) are reported. Using a 6 Li liquid scintillator, the energy spectrum of the fast neutron flux has been determined. Monte-Carlo calculations of the (α,n) and spontaneous fission processes in the surrounding rock has been performed and compared to the experimental result. In addition, using two 3 He neutron counters, the thermal neutron flux has been measured. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

  7. Neutron moisture measurement in materials

    International Nuclear Information System (INIS)

    Thony, J.L.

    1985-01-01

    This method is generally used for soil moisture determination but also for moisture in building materials. After a review of neutron interaction with matter (elastic and inelastic scattering, radiative capture and absorption with emission of charged particles) and of the equipment (source, detector and counting), gravimetric and chemical calibration are described and accuracy of measurement is discussed. 5 refs [fr

  8. Thermal neutron detectors based on hexagonal boron nitride epilayers

    Science.gov (United States)

    Doan, T. C.; Marty, A.; Li, J.; Lin, J. Y.; Jiang, H. X.

    2016-09-01

    Solid-state neutron detectors with high performances are urgently sought after for the detection of fissile materials. Until now, direct-conversion neutron detectors based on semiconductors with a measureable efficiency have not been realized. We have successfully synthesized hexagonal boron nitride (h-BN) epilayers with varying thicknesses (0.3 μm - 50 μm) by metal organic chemical vapor deposition (MOCVD) on sapphire substrates. In this paper, we present the detailed characterization of thermal neutron detectors fabricated from h-BN epilayers with a thickness up to 5 m to obtain insights into the h-BN epilayer thickness dependence of the device performance. The results revealed that the charge collection efficiency is almost independent of the h-BN epilayer thickness. By minimizing h-BN material removal by dry etching, it was shown that detectors incorporating an isotopically 10B-enriched h-BN epilayer of 2.7 μm in thickness exhibited an overall detection efficiency for thermal neutrons of 4% and a charge collection efficiency as high as 83%. By doing away altogether with dry etching, we have successfully realized a simple vertical 43 μm thick h-10BN detector which delivers a detection efficiency of 51.4% for thermal neutrons, which is the highest reported efficiency for any semiconductor-based neutron detector The h-BN detectors possess all the advantages of semiconductor devices including low cost, high efficiency and sensitivity, wafer-scale processing, compact size, light weight, and ability to integrate with other functional devices.

  9. Thermal neutron standard field with a Maxwellian distribution using the KUR heavy water facility

    International Nuclear Information System (INIS)

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

    1978-01-01

    A heavy water facility attached to the KUR (Kyoto University Reactor, swimming pool type. 5 MW) yeilds pure thermal neutrons with a Maxwellian distribution. The facility is placed next to the core of KUR and contains about 2t of heavy water. The width of the heavy water 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 a Maxwellian distribution in the whole 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 from the γ-n reaction in heavy water is very small. The maximum intensity of thermal neutrons is 3 X 10 11 n/cm.s. When a bismuth scatterrer is attached, the gamma ray contamination is decreased to a 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, diffusion length measurement, detector calibration, activation analysis and for biomedical purposes. (Auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-10-21

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

  11. Neutron transmission measurements at the IRMM pulsed neutron facility GELINA

    CERN Document Server

    Brusegan, A

    2002-01-01

    Measurements of neutron transmission are performed at the pulsed neutron source GELINA (Geel Linear Accelerator) of the JRC Institute of Reference Materials and Measurements in Belgium. This white neutron possibility is one of the world best as it posses an excellent energy resolution and good intensity, which offer the possibility of obtaining high quality neutron data. In the lecture are given typical characteristics of the time-of-flight (TOF) measurements performed here. The neutron transmission through a sample is connected directly with the total neutron cross section of the isotope. The precision of the experimental data is important for the exact determination of the neutron resonance parameters, which are relevant in applications like astrophysics and nuclear reactor design. The knowledge of the average nuclear properties, like the strength function, is desirable in nuclear reaction theory for comparison with predictions of different optical models. High quality of the transmission data is assured by...

  12. 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)

  13. Neutron energy measurement for practical applications

    Science.gov (United States)

    Roshan, M. V.; Sadeghi, H.; Ghasabian, M.; Mazandarani, A.

    2018-03-01

    Industrial demand for neutrons constrains careful energy measurements. Elastic scattering of monoenergetic α -particles from neutron collision enables neutron energy measurement by calculating the amount of deviation from the position where collision takes place. The neutron numbers with specific energy is obtained by counting the number of α -particles in the corresponding location on the charged particle detector. Monte Carlo simulation and COMSOL Multiphysics5.2 are used to account for one-to-one collision of neutrons with α -particles.

  14. Computed tomography with thermal neutrons and gaseous position sensitive detector

    International Nuclear Information System (INIS)

    Souza, Maria Ines Silvani

    2001-12-01

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

  15. Neutron energy measurement for practical applications

    Indian Academy of Sciences (India)

    M V Roshan

    2018-02-07

    Feb 7, 2018 ... monoenergetic α-particles from neutron collision enables neutron energy measurement by calculating the amount of deviation from the position ... nuclear fusion, and is of interest in the theory of light ion interactions [1]. Spin–orbit ... neutron modular detector to detect neutron energies. In our last paper [6], ...

  16. Nuclear reactor, fuel assembly and neutron measuring system

    International Nuclear Information System (INIS)

    Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

    1998-01-01

    The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

  17. Thermalization of monoenergetic neutrons in a concrete room

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  18. Measurement of internal stress within bulk materials using neutron diffraction

    International Nuclear Information System (INIS)

    Allen, A.; Andreani, C.; Hutchings, M.T.; Windsor, C.G.

    1981-01-01

    Neutron diffraction measures stress through the small changes in atomic lattice parameters caused by strain. The method is similar to X-ray diffraction with the vital difference that a thermal neutron beam penetrates several centimetres in most materials allowing measurements within bulk samples. Ways of achieving the high resolution necessary for quantitative measurements are described. Results are presented for the strains in a mild steel bar subjected to known elastic stresses, and on the internal stresses in a deformed bar. (author)

  19. distributions for the thermal neutron induced fission of 234U

    Directory of Open Access Journals (Sweden)

    Al-Adili A.

    2016-01-01

    In addition, the analysis of thermal neutron induced fission of 234U(n,f will be discussed. Currently analysis of data is ongoing, originally taken at the ILL reactor. The experiment is of particular interest since no measurement exist of the mass and energy distributions for this system at thermal energies. One main problem encountered during analysis was the huge background of 235U(nth,f. Despite the negligible isotopic traces in the sample, the cross section difference is enormous. Solution to this parasitic background will be highlighted.

  20. Measurement of neutron diffusion length in heavy concrete

    International Nuclear Information System (INIS)

    Krejci, D.

    2007-04-01

    Using an aluminium sampler filled with heavy concrete the neutron diffusion length was determined, measuring thermal and fast neutrons over the whole beam hole with various threshold detectors using gold samples. These calculations should describe the neutron distribution in the whole concrete shield of the reactor and contribute to the investigation of the activation of the concrete shield using reactor parameters like operating time, power and neutron flux. Instrumentation, activation and positioning of the samples in the beam hole of the TRIGA Mark II reactor are described. (nevyjel)

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

    International Nuclear Information System (INIS)

    Abdo, E.A.; Amin, E.

    1997-01-01

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

  2. Thermal neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Tuli, J.K.

    1983-01-01

    The energy and intensity of gamma rays as seen in thermal neutron capture are presented. Only those (n,α), E = thermal, reactions for which the residual nucleus mass number is greater than or equal to 45 are included. These correspond to evaluations published in Nuclear Data Sheets. The publication source data are contained in the Evaluated Nuclear Structure Data File (ENSDF). The data presented here do not involve any additional evaluation. Appendix I lists all the residual nuclides for which the data are included here. Appendix II gives a cumulated index to A-chain evaluations including the year of publication. The capture gamma ray data are given in two tables - the Table 1 is the list of all gamma rays seen in (n,#betta#) reaction given in the order of increasing energy; the Table II lists the gamma rays according to the nuclide

  3. 6Li foil thermal neutron detector

    Energy Technology Data Exchange (ETDEWEB)

    Ianakiev, Kiril D [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Favalli, Andrea [Los Alamos National Laboratory; Chung, Kiwhan [Los Alamos National Laboratory; Macarthur, Duncan W [Los Alamos National Laboratory

    2010-01-01

    In this paper we report on the design of a multilayer thermal neutron detector based on {sup 6}Li reactive foil and thin film plastic scintillators. The {sup 6}Li foils have about twice the intrinsic efficiency of {sup 10}B films and about four times higher light output due to a unique combination of high energy of reaction particles, low self absorption, and low ionization density of tritons. The design configuration provides for double sided readout of the lithium foil resulting in a doubling of the efficiency relative to a classical reactive film detector and generating a pulse height distribution with a valley between neutron and gamma signals similar to {sup 3}He tubes. The tens of microns thickness of plastic scintillator limits the energy deposited by gamma rays, which provides the necessary neutron/gamma discrimination. We used MCNPX to model a multilayer Li foil detector design and compared it with the standard HLNCC-II (18 {sup 3}He tubes operated at 4 atm). The preliminary results of the {sup 6}Li configuration show higher efficiency and one third of the die-away time. These properties, combined with the very short dead time of the plastic scintillator, offer the potential of a very high performance detector.

  4. Neutron spectrometry measurements in iron

    International Nuclear Information System (INIS)

    Perlini, G.; Acerbis, S.; Carter, M.

    1988-01-01

    A compact structure was prepared for use in making measurements of neutron penetration in iron which could serve as reference data and as a check for computer codes. About 30 iron plates were put together giving a useful overall length of 130 cm. At various depths along the central axis of the iron block, measurements were made with liquid scintillator spectrometers and proton recoil proportional counters. The energy band explored was between 14 KeV and 10 MeV. Here we report the original spectra of the impulses and the neutron spectra found by the NE213 code based on the differential method and by unfolding with the SPEC4 code for liquid scintillation counters and proton recoil spectrometers, respectively. 12 figs., 60 tabs., 6 refs

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

    Energy Technology Data Exchange (ETDEWEB)

    Urena N, F

    1998-12-31

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

  6. 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)

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

    Science.gov (United States)

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

    2016-01-01

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

  8. Neutron transmission measurements at the IRMM pulsed neutron facility GELINA

    International Nuclear Information System (INIS)

    Brusegan, A.

    2002-01-01

    Measurements of neutron transmission are performed at the pulsed neutron source GELINA (Geel Linear Accelerator) of the JRC Institute of Reference Materials and Measurements in Belgium. This white neutron possibility is one of the world best as it posses an excellent energy resolution and good intensity, which offer the possibility of obtaining high quality neutron data. In the lecture are given typical characteristics of the time-of-flight (TOF) measurements performed here. The neutron transmission through a sample is connected directly with the total neutron cross section of the isotope. The precision of the experimental data is important for the exact determination of the neutron resonance parameters, which are relevant in applications like astrophysics and nuclear reactor design. The knowledge of the average nuclear properties, like the strength function, is desirable in nuclear reaction theory for comparison with predictions of different optical models. High quality of the transmission data is assured by precise measurements and data analysis. The latter is done carefully by using sophisticated computer codes. The resolution function of the neutron source and the resonance Doppler broadening are objects of special attention. In the lecture are presented typical examples of the neutron transmission measurements at CELINA, the main direction of data applications and an overview of the future investigations in the field

  9. Measurement of in-phantom neutron flux and gamma dose in Tehran research reactor boron neutron capture therapy beam line.

    Science.gov (United States)

    Bavarnegin, Elham; Sadremomtaz, Alireza; Khalafi, Hossein; Kasesaz, Yaser

    2016-01-01

    Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. The doses from thermal neutron reactions with 14N and 10B are calculated by kinetic energy released per unit mass approach, after measuring thermal neutron flux using neutron activation technique. Gamma dose is measured using TLD-700 dosimeter. Different dose components have been measured in a head phantom which has been designed and constructed for BNCT purpose in TRR. Different in-phantom beam quality factors have also been determined. This study demonstrates that the TRR BNCT beam line has potential for treatment of superficial tumors.

  10. Thermal Properties Measurement Report

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Braase, Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hurley, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tonks, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gofryk, Krzysztof [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fielding, Randy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Knight, Collin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Meyer, Mitch [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    The Thermal Properties Measurement Report summarizes the research, development, installation, and initial use of significant experimental thermal property characterization capabilities at the INL in FY 2015. These new capabilities were used to characterize a U3Si2 (candidate Accident Tolerant) fuel sample fabricated at the INL. The ability to perform measurements at various length scales is important and provides additional data that is not currently in the literature. However, the real value of the data will be in accomplishing a phenomenological understanding of the thermal conductivity in fuels and the ties to predictive modeling. Thus, the MARMOT advanced modeling and simulation capability was utilized to illustrate how the microstructural data can be modeled and compared with bulk characterization data. A scientific method was established for thermal property measurement capability on irradiated nuclear fuel samples, which will be installed in the Irradiated Material Characterization Laboratory (IMCL).

  11. Actinide neutron-induced fission cross section measurements at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-01

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

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

    Indian Academy of Sciences (India)

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

  13. Real-time thermal neutron radiographic detection systems

    International Nuclear Information System (INIS)

    Berger, H.; Bracher, D.A.

    1976-01-01

    Systems for real-time detection of thermal neutron images are reviewed. Characteristics of one system are presented; the data include contrast, resolution and speed of response over the thermal neutron intensity range 2.5 10 3 n/cm 2 -sec to 10 7 n/cm 2 -sec

  14. Void fraction measurements using neutron radiography

    International Nuclear Information System (INIS)

    Glickstein, S.S.; Vance, W.H.; Joo, H.

    1992-01-01

    Real-time neutron radiography is being evaluated for studying the dynamic behavior of two phase flow and for measuring void fraction in vertical and inclined water ducts. This technique provides a unique means of visualizing the behavior of fluid flow inside thick metal enclosures. To simulate vapor conditions encountered in a fluid flow duct, an air-water flow system was constructed. Air was injected into the bottom of the duct at flow rates up to 0.47 I/s (1 cfm). The water flow rate was varied between 0--3.78 I/m (0--1 gpm). The experiments were performed at the Pennsylvania State University nuclear reactor facility using a real-time neutron radiography camera. With a thermal neutron flux on the order of 10 6 n/cm 2 /s directed through the thin duct dimension, the dynamic behavior of the air bubbles was clearly visible through 5 cm (2 in.) thick aluminum support plates placed on both sides of the duct wall. Image analysis techniques were employed to extract void fractions from the data which was recorded on videotape. This consisted of time averaging 256 video frames and measuring the gray level distribution throughout the region. The distribution of the measured void fraction across the duct was determined for various air/water mixtures. Details of the results of experiments for a variety of air and water flow conditions are presented

  15. Determination of thermal neutrons diffusion length in graphite; Determinacion de la Longitud de Difusion de los Neutrones Termicos en Grafito

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Fite, J.

    1959-07-01

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

  16. Neutron Diffraction Measurements and First Principles Study of Thermal Motion of Atoms in Select M_{n+1}AX_n and Binary MX Transition Metal Carbide Phases

    OpenAIRE

    Lane, Nina J.; Vogel, Sven C.; Hug, Gilles; Togo, Atsushi; Chaput, Laurent; Hultman, Lars; Barsoum, Michel W.

    2012-01-01

    Herein, we compare the thermal vibrations of atoms in select ternary carbides with the formula Mn+1AXn ("MAX phases," M = Ti, Cr; A = Al, Si, Ge; X = C, N) as determined from first principles phonon calculations to those obtained from high-temperature neutron powder diffraction studies. The transition metal carbides TiC, TaC, and WC are also studied to test our methodology on simpler carbides. Good qualitative and quantitative agreement is found between predicted and experimental values for t...

  17. Measurement channel of neutron flow based on software

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  18. Measurement of thermal conductance

    International Nuclear Information System (INIS)

    Kuchnir, M.

    1977-01-01

    The 6-m long, 45-kG, warm-iron superconducting magnets envisioned for the Energy Doubler stage of the Fermilab accelerator require stiff supports with minimized thermal conductances in order to keep the refrigeration power reasonable. The large number of supports involved in the system required a careful study of their heat conduction from the room temperature wall to the intercepting refrigeration at 20 0 K and to the liquid helium. For this purpose the thermal conductance of this support was measured by comparing it with the thermal conductance of a copper strap of known geometry. An association of steady-state thermal analysis and experimental thermal conductivity techniques forms the basis of this method. An important advantage is the automatic simulation of the 20 0 K refrigeration intercept by the copper strap, which simplifies the apparatus considerably. This relative resistance technique, which uses electrical analogy as a guideline, is applicable with no restrictions for materials with temperature-independent thermal conductivity. For other materials the results obtained are functions of the specific temperature interval involved in the measurements. A comprehensive review of the literature on thermal conductivity indicates that this approach has not been used before. A demonstration of its self-consistency is stressed here rather than results obtained for different supports

  19. Cascade γ rays following capture of thermal neutrons on 113Cd

    Science.gov (United States)

    Rusev, G.; Jandel, M.; Krtička, M.; Arnold, C. W.; Bredeweg, T. A.; Couture, A.; Moody, W. A.; Mosby, S. M.; Ullmann, J. L.

    2013-11-01

    Intensity distributions of cascade γ-ray transitions following the capture of thermal neutrons by 113Cd have been measured at the Los Alamos Neutron Science Center for various γ-ray multiplicities. The experiment was carried out at the highly segmented 4π γ-ray calorimeter—Detector for Advanced Neutron Capture Experiments (DANCE). A measured two-dimensional spectrum of counts versus γ-ray energy versus γ-ray multiplicity, from the strongest resonance in the 113Cd(n,γ) reaction at 0.178 eV has been compared to predictions from the statistical model. The best representation of the γ-ray cascades following the capture of thermal neutrons on 113Cd is presented. The intensity distribution of these cascades is of great importance for estimates of response to thermal neutrons of devices that use natural or enriched cadmium.

  20. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    Knoll, G.F.

    1992-01-01

    From its inception, the Nuclear Data Project at the University of Michigan has concentrated on two major objectives: (1) to carry out carefully controlled nuclear measurements of the highest possible reliability in support of the national nuclear data program, and (2) to provide an educational opportunity for students with interests in experimental nuclear science. The project has undergone a successful transition from a primary dependence on our photoneutron laboratory to one in which our current research is entirely based on a unique pulsed 14 MeV fast neutron facility. The new experimental facility is unique in its ability to provide nanosecond bursts of 14 MeV neutrons under conditions that are ''clean'' and as scatter-free as possible, and is the only one of its type currently in operation in the United States. It has been designed and put into operation primarily by graduate students, and has met or exceeded all of its important initial performance goals. We have reached the point of its routine operation, and most of the data are now in hand that will serve as the basis for the first two doctoral dissertations to be written by participating graduate students. Our initial results on double differential neutron cross sections will be presented at the May 1993 Fusion Reactor Technology Workshop. We are pleased to report that, after investing several years in equipment assembly and optimization, the project has now entered its ''data production'' phase

  1. 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.

  2. Neutron stress measurement using neutron image plate. 4

    International Nuclear Information System (INIS)

    Sasaki, Toshihiko; Hirose, Yukio; Minakawa, Noriaki; Morii, Yukio; Niimura, Nobuo

    2001-01-01

    A fundamental study on the neutron stress measurement using an image plate neutron detector was described. First, a method for determining internal mean stress over the area where incident beams pass through was explained. The method is based on the α angle method by which the stress analysis is conducted using information along a diffraction ring. Second, a neutron diffraction experiment was shown. This was conducted using the research reactor, JRR-3M, at Japan Atomic Energy Research Institute (JAERI). The 211 diffraction rings from steel sample were obtained using the image plate neutron detector. It was found that the mean stress obtained by the image plate agreed with stress applied. (author)

  3. An extension of diffusion theory for thermal neutrons near boundaries

    International Nuclear Information System (INIS)

    Alvarez Rivas, J. L.

    1963-01-01

    The distribution of thermal neutron flux has been measured inside and outside copper rods of several diameters, immersed in water. It has been found that these distributions can be calculated by means of elemental diffusion theory if the value of the coefficient of diffusion is changed. this parameter is truly a diffusion coefficient, which now also depends on the diameter of the rod. Through a model an expression of this coefficient is introduced which takes account of the measurements of the author and of those reported in PIGC P/928 (1995), ANL-5872 (1959), DEGR 319 (D) (1961). This model could be extended also to plane geometry. (Author) 19 refs

  4. Calibration of the nuclear power channels of the IPEN/MB-01 reactor obtained from the measurements of the spatial thermal neutron flux distribution in the reactor core through the irradiation of infinitely diluted gold foils

    International Nuclear Information System (INIS)

    Goncalves, Lucas Batista

    2008-01-01

    Several nuclear parameters are obtained through the gamma spectrometry of targets irradiated in a research reactor core and this is the case of the activation foils which make possible, through the measurements of the activity induced, to determine the neutron flux in the place where they had been irradiated. The power level operation of the reactor is a parameter directly proportional to the average neutron flux in the core. This work aims to get the power operation of the reactor through of spatial neutron flux distribution in the core of IPEN/MB-01 reactor by the irradiation of infinitely diluted gold foils and prudently located in its interior. These foils were made in the form of metallic alloy in concentration levels such that the phenomena of flux disturbance, as the self-shielding factors to neutrons become worthless. These activation foils has only 1% of dispersed gold atoms in an aluminium matrix content of 99% of this element. The irradiations of foils have been carried through with and without cadmium plate. The total correlation between the average thermal neutron flux obtained by irradiation of infinitely diluted activation foils and the average digital value of current of the nuclear power channels 5 and 6 (non-compensated ionization chambers - CINC), allow the calibration of the nuclear channels of the IPEN/MB-01 reactor. (author)

  5. Neutron spectrum measurement in a medical Cyclotron

    International Nuclear Information System (INIS)

    Sathian, V.; Phadnis, U.V.; Sharma, D.N.; Deepa, S.; Soni, P.S.

    2008-01-01

    Radio nuclides are being used in diagnostic purpose as well as therapeutic purpose under medical category for early detection of malignancies. The diagnosis using radio nuclides are playing very important role. PET scan is a miles stone in medical history. These scan is done using a positron emitter 18 F. This radio nuclide is produced by the reaction 18 O(p, n) 18 F in Medical cyclotron of Radiation Medicine center (RMC), India. The neutron production is an unwanted by product. These neutrons will make lot of radiation protection problem as neutron itself and also these neutrons will produce induced activity in the different elements of cyclotron shielding material as well as the wall materials in long run. So the accurate neutron dose measurement is very important and this depends on the accurate neutron spectrum. The unfolding of neutron spectrum and measurement of neutron dose in cyclotron of RMC using activation foil technique is presented in this paper. (author)

  6. Physics with Ultracold and Thermal Neutron Beams

    International Nuclear Information System (INIS)

    None

    2004-01-01

    The final report is broken into 5 segments, reflecting research conclusions reached during specific time periods: 1991-1997, 1997-1999, 1999-2000, 2000-2001, and 2001-2002. The first part of the work reported was carried out at the 2 Mw research reactor of the Rhode Island Nuclaer Science Center (RJNSC). Chosen for study was the slow phase separation in mixtures of oil and water in the presence of a surfactant, and the structural features of an oil layer during the slow build-up from the gas phase. The results of these measurements, as well as studies of the capillary wave properties of oil/surfactant/water interfaces are described. The second part of the work was performed at the neutron reflection facilities of the Intennse Pulsed Neutron Source at Argonne and of the NBSR reactor at NIST. At Argonne, the uniaxial magnetic order of an Fe/CR superlattice was investigated, while the experiments at NIST studied the swelling behavior of ordered thin films of diblock copolymers when they were exposed to solvent vapors. The third part of the work was concerned with the storage properties of ultracold neturons in a trap. New experiments on spectral evolution during storage, using the UCN source of the Institut Laue-Langevin were able to be run. Subsequent periods focussed on the ultracold neutrons work, spin valve multilayer systems, and pseudo-partial wetting

  7. Thermal Testing Measurements Report

    Energy Technology Data Exchange (ETDEWEB)

    R. Wagner

    2002-09-26

    The purpose of the Thermal Testing Measurements Report (Scientific Analysis Report) is to document, in one report, the comprehensive set of measurements taken within the Yucca Mountain Project Thermal Testing Program since its inception in 1996. Currently, the testing performed and measurements collected are either scattered in many level 3 and level 4 milestone reports or, in the case of the ongoing Drift Scale Test, mostly documented in eight informal progress reports. Documentation in existing reports is uneven in level of detail and quality. Furthermore, while all the data collected within the Yucca Mountain Site Characterization Project (YMP) Thermal Testing Program have been submitted periodically to the Technical Data Management System (TDMS), the data structure--several incremental submittals, and documentation formats--are such that the data are often not user-friendly except to those who acquired and processed the data. The documentation in this report is intended to make data collected within the YMP Thermal Testing Program readily usable to end users, such as those representing the Performance Assessment Project, Repository Design Project, and Engineered Systems Sub-Project. Since either detailed level 3 and level 4 reports exist or the measurements are straightforward, only brief discussions are provided for each data set. These brief discussions for different data sets are intended to impart a clear sense of applicability of data, so that they will be used properly within the context of measurement uncertainty. This approach also keeps this report to a manageable size, an important consideration because the report encompasses nearly all measurements for three long-term thermal tests. As appropriate, thermal testing data currently residing in the TDMS have been reorganized and reformatted from cumbersome, user-unfriendly Input-Data Tracking Numbers (DTNs) into a new set of Output-DTNs. These Output-DTNs provide a readily usable data structure

  8. UCN Source at an External Beam of Thermal Neutrons

    Directory of Open Access Journals (Sweden)

    E. V. Lychagin

    2015-01-01

    Full Text Available We propose a new method for production of ultracold neutrons (UCNs in superfluid helium. The principal idea consists in installing a helium UCN source into an external beam of thermal or cold neutrons and in surrounding this source with a solid methane moderator/reflector cooled down to ~4 K. The moderator plays the role of an external source of cold neutrons needed to produce UCNs. The flux of accumulated neutrons could exceed the flux of incident neutrons due to their numerous reflections from methane; also the source size could be significantly larger than the incident beam diameter. We provide preliminary calculations of cooling of neutrons. These calculations show that such a source being installed at an intense source of thermal or cold neutrons like the ILL or PIK reactor or the ESS spallation source could provide the UCN density 105 cm−3, the production rate 107 UCN/s−1. Main advantages of such an UCN source include its low radiative and thermal load, relatively low cost, and convenient accessibility for any maintenance. We have carried out an experiment on cooling of thermal neutrons in a methane cavity. The data confirm the results of our calculations of the spectrum and flux of neutrons in the methane cavity.

  9. Neutron capture therapy with thermal neutrons at IRT MIFI

    International Nuclear Information System (INIS)

    Zajtsev, K.N.; Portnov, A.A.; Savkin, V.A.; Kulakov, V.N.; Khokhlov, V.F.; Shejno, I.N.; Vajnson, A.A.; Kozlovskaya, N.G.; Meshcherikova, V.V.; Mitin, V.N.; Yarmonenko, S.P.

    2001-01-01

    Combined preclinical investigations into neutron capture therapy are conducted. Malignant melanoma was adopted as the line of investigation; boron-containing and gadolinium-containing preparations were used during the neutron capture therapy working off. Preparations produce secondary varying radiations when used in tumor. Dogs with spontaneous melanoma were used for the experiments. Procedures for the irradiation of dogs by neutron beam as the stage before use for the treatment of oncology patients were finished off; efficiency of neutron beam influence on normal tissues during the irradiation of dogs with melanoma (and without it) in antitumor and side effect sense was estimated [ru

  10. Diverse Neutron Scattering Measurements in Unconventional Superconductors

    Science.gov (United States)

    Carr, Scott

    Superconductivity is an amazing phenomena. Beyond the eye-popping demos and promises of floating trains, it is a macroscopic manifestation of pure quantum physics. Where most quantum effects are smothered by thermal energy even at the lowest of temperatures, superconductivity brings the strangeness of quantum mechanics to reasonably accessible temperatures. An while it does have applications in terms of efficient charge transfer and the production of powerful magnetic fields, this is not why we study it. The mystery of its origin, oddness of its effects, and beauty of its consequences drives us to delve into the richness of this amazing phenomena. One key player in the physics of these exotic superconductors appears to be magnetism. Magnetism is a constant presence in the phase diagrams of exotic superconductors and appears to couple to the superconducting transition in some interesting ways. The details of the connection remain unknown, but it is clear that magnetism, either through static order, fluctuations, or both, is an important aspect of the crystalline environments which foster superconductivity. Through the tool of neutron scattering I investigate this connection across an array or different compounds using a variety of techniques. In this thesis, I conclude on the role of phase separation in superconducting K yFe1.6+xSe2 using neutron diffraction, perform a survey of magnetic excitations across the superconducting phase diagram in NaFe1-xCoxAs with neutron time-of-flight spectroscopy, and reveal new magnetic properties of the old heavy fermion superconductor UPt3 via more traditional neutron measurements.

  11. Measurement of internal conversion electrons from Gd neutron capture

    Energy Technology Data Exchange (ETDEWEB)

    Kandlakunta, P. [Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, Columbus, OH 43210 (United States); Cao, L.R., E-mail: cao.152@osu.edu [Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, Columbus, OH 43210 (United States); Mulligan, P. [Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, Columbus, OH 43210 (United States)

    2013-03-21

    Gadolinium (Gd) is a suitable material for neutron conversion because of its superior neutron absorption cross-section. However, the principal secondary particles that generate electron-hole pairs in a semiconductor detector after Gd neutron capture are low-energy internal conversion (IC) electrons. We measured the IC electron spectrum due to Gd neutron capture by using a thermal neutron beam and a digitizer-based multidetector spectroscopy. We also discussed the effective use of the IC electrons in the context of a twin-detector design and the associated gamma-ray rejection issues. Extensive simulations of the spectra of IC electrons and gamma rays agreed well with the experimental results; both types of results support the feasibility of the proposed n–γ separation method.

  12. Feasibility of fiber-optic radiation sensor using Cerenkov effect for detecting thermal neutrons.

    Science.gov (United States)

    Jang, Kyoung Won; Yagi, Takahiro; Pyeon, Cheol Ho; Yoo, Wook Jae; Shin, Sang Hun; Misawa, Tsuyoshi; Lee, Bongsoo

    2013-06-17

    In this research, we propose a novel method for detecting thermal neutrons with a fiber-optic radiation sensor using the Cerenkov effect. We fabricate a fiber-optic radiation sensor that detects thermal neutrons with a Gd-foil, a rutile crystal, and a plastic optical fiber. The relationship between the fluxes of electrons inducing Cerenkov radiation in the sensor probe of the fiber-optic radiation sensor and thermal neutron fluxes is determined using the Monte Carlo N-particle transport code simulations. To evaluate the fiber-optic radiation sensor, the Cerenkov radiation generated in the fiber-optic radiation sensor by irradiation of pure thermal neutron beams is measured according to the depths of polyethylene.

  13. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    1991-01-01

    In the 14 MeV Neutron Laboratory, we have continued the development of a facility that is now the only one of its kind in operation in the United States. We have refined the klystron bunching system described in last year's report to the point that 1.2 nanosecond pulses have been directly measured. We have tested the pulse shape discrimination capability of our primary NE 213 neutron detector. We have converted the RF sweeper section of the beamline to a frequency of 1 MHz to replace the function of the high voltage pulser described in last year's report which proved to be difficult to maintain and unreliable in its operation. We have also overcome several other significant experimental difficulties, including a major problem with a vacuum leak in the main accelerator column. We have completed additional testing to prove the remainder of the generation and measurement systems, but overcoming some of these experimental difficulties has delayed the start of actual data taking. We are now in a position to begin our first series of ring geometry elastic scattering measurements, and these will be underway before the end of the current contract year. As part of our longer term planning, we are continuing the conceptual analysis of several schemes to improve the intensity of our current pulsed beam. These include the provision of a duoplasmatron ion source and/or the provision of preacceleration bunching. Additional details are given later in this report. A series of measurements were carried out at the Tandem Dynamatron Facility involving the irradiation of a series of yttrium foils and the determination of activation cross sections using absolute counting techniques. The experimental work has been completed, and final analysis of the cross section data will be completed within several months

  14. IMPROVED COMPUTATIONAL CHARACTERIZATION OF THE THERMAL NEUTRON SOURCE FOR NEUTRON CAPTURE THERAPY RESEARCH AT THE UNIVERSITY OF MISSOURI

    Energy Technology Data Exchange (ETDEWEB)

    Stuart R. Slattery; David W. Nigg; John D. Brockman; M. Frederick Hawthorne

    2010-05-01

    Parameter studies, design calculations and initial neutronic performance measurements have been completed for a new thermal neutron beamline to be used for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evaluation are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. This is essential for detailed dosimetric studies required for the anticipated research program.

  15. A thermal neutron scattering law for yttrium hydride

    Science.gov (United States)

    Zerkle, Michael; Holmes, Jesse

    2017-09-01

    Yttrium hydride (YH2) is of interest as a high temperature moderator material because of its superior ability to retain hydrogen at elevated temperatures. Thermal neutron scattering laws for hydrogen bound in yttrium hydride (H-YH2) and yttrium bound in yttrium hydride (Y-YH2) prepared using the ab initio approach are presented. Density functional theory, incorporating the generalized gradient approximation (GGA) for the exchange-correlation energy, is used to simulate the face-centered cubic structure of YH2 and calculate the interatomic Hellmann-Feynman forces for a 2 × 2 × 2 supercell containing 96 atoms. Lattice dynamics calculations using PHONON are then used to determine the phonon dispersion relations and density of states. The calculated phonon density of states for H and Y in YH2 are used to prepare H-YH2 and Y-YH2 thermal scattering laws using the LEAPR module of NJOY2012. Analysis of the resulting integral and differential scattering cross sections demonstrates adequate resolution of the S(α,β) function. Comparison of experimental lattice constant, heat capacity, inelastic neutron scattering spectra and total scattering cross section measurements to calculated values are used to validate the thermal scattering laws.

  16. A thermal neutron scattering law for yttrium hydride

    Directory of Open Access Journals (Sweden)

    Zerkle Michael

    2017-01-01

    Full Text Available Yttrium hydride (YH2 is of interest as a high temperature moderator material because of its superior ability to retain hydrogen at elevated temperatures. Thermal neutron scattering laws for hydrogen bound in yttrium hydride (H-YH2 and yttrium bound in yttrium hydride (Y-YH2 prepared using the ab initio approach are presented. Density functional theory, incorporating the generalized gradient approximation (GGA for the exchange-correlation energy, is used to simulate the face-centered cubic structure of YH2 and calculate the interatomic Hellmann-Feynman forces for a 2 × 2 × 2 supercell containing 96 atoms. Lattice dynamics calculations using PHONON are then used to determine the phonon dispersion relations and density of states. The calculated phonon density of states for H and Y in YH2 are used to prepare H-YH2 and Y-YH2 thermal scattering laws using the LEAPR module of NJOY2012. Analysis of the resulting integral and differential scattering cross sections demonstrates adequate resolution of the S(α,β function. Comparison of experimental lattice constant, heat capacity, inelastic neutron scattering spectra and total scattering cross section measurements to calculated values are used to validate the thermal scattering laws.

  17. Shift-register coincidence electronics system for thermal neutron counters

    International Nuclear Information System (INIS)

    Swansen, J.E.; Collinsworth, P.R.; Krick, M.S.

    1980-04-01

    An improved shift-register, coincidence-counting logic circuit, developed for use with thermal neutron well counters, is described in detail. A distinguishing feature of the circuit is its ability to operate usefully at neutron counting rates of several hundred kHz. A portable electronics package incorporating the new coincidence logic and support circuits is also described

  18. Thermal neutron counts and derivated charts | Odusote | Journal of ...

    African Journals Online (AJOL)

    thermal neutrons)” approximation. The resulting equation was applied with a mixing index, , for various formation matrices and porosities. The ratio of counts from two different detectors was plotted as a function of porosity for these formations.

  19. 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

  20. In core measurement and monitoring of reactor (neutron) radiation field

    International Nuclear Information System (INIS)

    Erben, O.

    1985-01-01

    A survey is presented of in core radiation detectors. The principles are described of activation detectors, fission chambers, self-powered neutron detectors and thermal sensors. Systems of in core measurement for WWER nuclear power plants, nuclear reactors of power plants operated by KWU, Babcock and Wilcox, Combustion Engineering and FRAMATOME are described. (E.S.)

  1. METHOD OF TESTING THERMAL NEUTRON FISSIONABLE MATERIAL FOR PURITY

    Science.gov (United States)

    Fermi, E.; Anderson, H.L.

    1961-01-24

    A process is given for determining the neutronic purity of fissionable material by the so-called shotgun test. The effect of a standard neutron absorber of known characteristics and amounts on a neutronic field also of known characteristics is measured and compared with the effect which the impurities derived from a known quantity of fissionable material has on the same neutronic field. The two readings are then made the basis of calculation from which the amount of impurities can be computed.

  2. Neutron stress measurement of W-fiber reinforced Cu composite

    CERN Document Server

    Nishida, M; Ikeuchi, Y; Minakawa, N

    2003-01-01

    Stress measurement methods using neutron and X-ray diffraction were examined by comparing the surface stresses with internal stresses in the continuous tungsten-fiber reinforced copper-matrix composite. Surface stresses were measured by X-ray stress measurement with the sin sup 2 psi method. Furthermore, the sin sup 2 psi method and the most common triaxal measurement method using Hooke's equation were employed for internal stress measurement by neutron diffraction. On the other hand, microstress distributions developed by the difference in the thermal expansion coefficients between these two phases were calculated by FEM. The weighted average strains and stresses were compared with the experimental results. The FEM results agreed with the experimental results qualitatively and confirmed the importance of the triaxial stress analysis in the neutron stress measurement. (Abstract Copyright [2003], Wiley Periodicals, Inc.)

  3. Neutron stress measurement of W-fiber reinforced Cu composite

    International Nuclear Information System (INIS)

    Nishida, M.; Hanabusa, T.; Ikeuchi, Y.; Minakawa, N.

    2003-01-01

    Stress measurement methods using neutron and X-ray diffraction were examined by comparing the surface stresses with internal stresses in the continuous tungsten-fiber reinforced copper-matrix composite. Surface stresses were measured by X-ray stress measurement with the sin 2 ψ method. Furthermore, the sin 2 ψ method and the most common triaxal measurement method using Hooke's equation were employed for internal stress measurement by neutron diffraction. On the other hand, microstress distributions developed by the difference in the thermal expansion coefficients between these two phases were calculated by FEM. The weighted average strains and stresses were compared with the experimental results. The FEM results agreed with the experimental results qualitatively and confirmed the importance of the triaxial stress analysis in the neutron stress measurement. (Abstract Copyright [2003], Wiley Periodicals, Inc.)

  4. Neutron-diffraction measurements of stress

    International Nuclear Information System (INIS)

    Holden, T.M.

    1995-01-01

    Experiments on bent steam-generator tubing have shown that different diffraction peaks, (1 1 1) or (0 0 2), give different results for the sign and magnitude of the stress and strain. From an engineering standpoint, the macroscopic stress field cannot be both positive and negative in the same volume, so this difference must be due to intergranular effects superposed on the macroscopic stress field. Uniaxial tensile test experiments with applied stresses beyond the 0.2% offset yield stress, help to understand this anomaly, by demonstrating the different strain response to applied stress along different crystallographic axes.When Zr-alloys are cooled from elevated temperatures, thermal stresses always develop, so that it is difficult to obtain a stress-free lattice spacing from which residual strains may be derived. From measurements of the temperature dependence of lattice spacing, the temperature at which the thermal stresses vanish may be found. From the lattice spacing at this temperature the stress-free lattice spacings at room temperature can be obtained readily.To interpret the measured strains in terms of macroscopic stress fields it is necessary to know the diffraction elastic constants. Neutron diffraction measurements of the diffraction elastic constants in a ferritic steel for the [1 1 0], [0 0 2] and [2 2 2] crystallographic axes, in directions parallel and perpendicular to the applied stress are compared with theoretical diffraction elastic constants. (orig.)

  5. A high performance neutron spectrometer for planetary hydrogen measurement

    Science.gov (United States)

    Naito, Masayuki; Hasebe, Nobuyuki; Nagaoka, Hiroshi; Ishii, Junya; Aoki, Daisuke; Shibamura, Eido; Kim, Kyeong J.; Matias-Lopes, José A.; Martínez-Frías, Jesús

    2017-08-01

    The elemental composition and its distribution on planetary surface provide important constraints on the origin and evolution of the planetary body. The nuclear spectrometer consisting of a neutron spectrometer and a gamma-ray spectrometer obtains elemental compositions by remote sensing. Especially, the neutron spectrometer is able to determine the hydrogen concentration, a piece of information that plays an important role in thermal history of the planets. In this work, numerical and experimental studies on the neutron spectrometer for micro-satellite application were conducted. It is found that background count rate of neutron produced from micro-satellite is very small, which enables to obtain successful results in short time observation. The neutron spectrometer combining a lithium-6 glass scintillator with a boron loaded plastic scintillator was used to be able to detect neutrons in different energy ranges. It was experimentally confirmed that the neutron signals from these scintillators were successfully discriminated by the difference of scintillation decay time between two detectors. The measurement of neutron count rates of two scintillators is found to determine hydrogen concentration on the planetary surfaces in the future missions.

  6. 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})

  7. Overview of soil moisture measurements with neutrons

    Science.gov (United States)

    Hendriks, Aagje; Steele-Dunne, Susan; van de Giesen, Nick

    2014-05-01

    Soil moisture measurements are useful for hydrological and agricultural applications. Soil moisture can be measured with a range of in-situ sensors in the soil, such as probes based on the difference in dielectric permittivity of wet and dry soil. At a large scale of tenths of kilometres, soil moisture can be measured with microwave remote sensing from satellites. At the intermediate scale, detection methods such as GPS reflectometry and the use of cosmic rays have been developed recently. One of the principles that can be used to measure soil moisture, is the difference in behaviour of neutrons in wet and dry soil. Neutrons are massive, electrically neutral particles that transfer their energy easily to light atoms, such as hydrogen. Therefore, in wet soil, neutrons lose their energy quickly. In dry soil, they scatter elastically from the heavy atoms and can be detected. The amount of detected neutrons is therefore inversely correlated with the amount of hydrogen in the soil. In this research we look for an overview of the possibilities to measure soil moisture with neutrons and how neutrons can be detected. Neutrons can be used to measure at the point scale and at a larger scale of approximately 1 km. We discuss in-situ measurements, in which a neutron source is put into the soil. Immediately next to the source is a detector, that counts the amount of neutrons that scatters back if the soil is dry. At a larger scale or measurement volume, we discuss the measurement of soil moisture with neutrons from cosmic rays. Cosmic rays are charged particles, accelerated by astrophysical sources (such as a Supernova). When the particles enter the atmosphere, they interact with the atmospheric atoms and form a shower. At sea level, we find several types of particles, such as muons and neutrons. We discuss why neutrons would be more useful for soil moisture measurements than other particles and how the use of cosmic-ray neutrons influences the measurement volume. Here we

  8. INTENSE THERMAL NEUTRON FIELDS FROM A MEDICAL-TYPE LINAC: THE E_LIBANS PROJECT.

    Science.gov (United States)

    Costa, M; Durisi, E; Ferrero, M; Monti, V; Visca, L; Anglesio, S; Bedogni, R; Gomez-Ros, J M; Romano, M; Planell, O Sans; Treccani, M; Bortot, D; Pola, A; Alikaniotis, K; Giannini, G

    2017-12-22

    The e_LiBANS project aims at producing intense thermal neutron fields for diverse interdisciplinary irradiation purposes. It makes use of a reconditioned medical electron LINAC, recently installed at the Physics Department and INFN in Torino, coupled to a dedicated photo-converter, developed within this collaboration, that uses (γ,n) reaction within high Z targets. Produced neutrons are then moderated to thermal energies and concentrated in an irradiation volume. To measure and to characterize in real time the intense field inside the cavity new thermal neutron detectors were designed with high radiation resistance, low noise and very high neutron-to-photon discrimination capability. This article offers an overview of the e_LiBANS project and describes the results of the benchmark experiment. © The Author(s) 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

    Science.gov (United States)

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

    2013-10-01

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

  10. Feasibility study on using imaging plates to estimate thermal neutron fluence in neutron-gamma mixed fields.

    Science.gov (United States)

    Fujibuchi, Toshioh; Tanabe, Yu; Sakae, Takeji; Terunuma, Toshiyuki; Isobe, Tomonori; Kawamura, Hiraku; Yasuoka, Kiyoshi; Matsumoto, Tetsuro; Harano, Hideki; Nishiyama, Jun; Masuda, Akihiko; Nohtomi, Akihiro

    2011-11-01

    In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field.

  11. Neutron moderation theory with thermal motion of the moderator nuclei

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-15

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

  12. Measuring the Neutron Lifetime using Magnetically Trapped Ultracold Neutrons

    Science.gov (United States)

    Mumm, H. P.; Coakley, K. J.; Dewey, M. S.; Huber, M. G.; Hughes, P. P.; Thompson, A. K.; Golub, R.; Huffer, C. R.; Huffman, P. R.; O'Shaughnessy, C. M.; Schelhammer, K. W.

    2010-11-01

    The neutron beta-decay lifetime is important in both theoretical predictions of the primordial abundance of ^4He and providing a strong unitarity test of the CKM mixing matrix. We have previously demonstrated trapping of Ultracold Neutrons (UCN) in a magnetic trap, and, though statistically limited, measured a lifetime consistent with the world average. A major upgrade of the apparatus has now been completed at NIST. In our unique approach, a 0.89 nm neutron beam is incident on a superfluid ^4He target within the minimum field region of an Ioffe-type magnetic trap. Neutrons are downscattered by single phonon scattering in liquid helium to near rest and trapped; at sufficiently low temperatures, the low phonon density in the helium suppresses upscatter. The electron accompanying neutron decay produces scintillation in the superfluid helium and can be detected in real time. Previous statistical limitations as well as systematics related to neutron material bottling will be reduced by significant increases in field strength and trap volume. Details of analyses of the systematics as well as the initial performance benchmarks of the new apparatus will be presented.

  13. Thermal Neutron Imaging Using A New Pad-Based Position Sensitive Neutron Detector

    Energy Technology Data Exchange (ETDEWEB)

    Dioszegi I.; Vanier P.E.; Salwen C.; Chichester D.L.; Watson S.M.

    2016-10-29

    Thermal neutrons (with mean energy of 25 meV) have a scattering mean free path of about 20 m in air. Therefore it is feasible to find localized thermal neutron sources up to ~30 m standoff distance using thermal neutron imaging. Coded aperture thermal neutron imaging was developed in our laboratory in the nineties, using He-3 filled wire chambers. Recently a new generation of coded-aperture neutron imagers has been developed. In the new design the ionization chamber has anode and cathode planes, where the anode is composed of an array of individual pads. The charge is collected on each of the individual 5x5 mm2 anode pads, (48x48 in total, corresponding to 24x24 cm2 sensitive area) and read out by application specific integrated circuits (ASICs). The high sensitivity of the ASICs allows unity gain operation mode. The new design has several advantages for field deployable imaging applications, compared to the previous generation of wire-grid based neutron detectors. Among these are the rugged design, lighter weight and use of non-flammable stopping gas. For standoff localization of thermalized neutron sources a low resolution (11x11 pixel) coded aperture mask has been fabricated. Using the new larger area detector and the coarse resolution mask we performed several standoff experiments using moderated californium and plutonium sources at Idaho National Laboratory. In this paper we will report on the development and performance of the new pad-based neutron camera, and present long range coded-aperture images of various thermalized neutron sources.

  14. Independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239Pu

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Forman, L.

    1975-01-01

    The relative independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239 Pu have been measured on line using a mass spectrograph and thermalized neutrons from a burst reactor. Independent yields were derived by normalizing the measurements to products of chain yields and fractional independent yields, estimating the latter from measured cumulative yields of Kr and Xe. Comparing the independent yields with those from 238 U fission, the 239 Pu results show shifts in isotopic yield distribution toward lower mass for both Rb and Cs and also toward the production of more Cs and less Rb when 239 Pu is fissioned

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

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2000-01-01

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

  16. Measurement of the Neutron Spectrum of a DD Electronic Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Chichester; J. T. Johnson; E. H. Seabury

    2010-08-01

    A Cuttler-Shalev (C-S) 3He proportional counter has been used to measure the energy spectrum of neutrons from a portable deuterium-deuterium electronic neutron generator. To improve the analysis of results from the C-S detector digital pulse shape analysis techniques have been used to eliminate neutron recoil artifacts in the recorded data. Data was collected using a 8-GHz, 10-bit waveform digitizer with its full scale corresponding to approximately 6-MeV neutrons. The measurements were made with the detector axis perpendicular to the direction of ions in the ENG in a plane 0.5-m to the side of the ENG, measuring neutrons emitted at an angle from 87.3? to 92.7? with respect to the path of ions in the ENG. The system demonstrated an energy resolution of approximately 0.040 MeV for the thermal peak and approximately 0.13 MeV at the DD neutron energy. In order to achieve the ultimate resolution capable with this type of detector it is clear that a higher-precision digitizer will be needed.

  17. Neutron energy response measurement of scintillation detectors

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  18. A study on measurement of neutrons generated in radiation therapy – Measurement of neurons in CR-39 detection method

    International Nuclear Information System (INIS)

    Park, Cheol-Soo; Cho, Jae-Hwan; Lee, Hae-Kag; Lee, Sun-Yeob; Jang, Hyon-Chol; Dong, Kyung-Rae; Chung, Woon-Kwan; Jin, Lee; Moon, Deog-Hwan; Lee, Kwang-Sung; Yang, Nam-Oh; Cho, Moo-Seong

    2013-01-01

    Highlights: ► To measure the neutrons generated in a linear accelerator. ► Both fast neutrons and thermal neutrons produced an increase in the dose of neutrons generated with increasing irradiation dose. ► The generation of neutrons increased when a wedge filter was used. ► When the SRS cone that required a high dose was used, more neutrons were detected. -- Abstract: The CR-39 [diethylene glycol bis-(allylcarbonate)] neuron detection method was used to measure the dose of neutrons generated in X-ray (photon) therapy conducted in a linear accelerator, and to use high-energy photons as part of the clinical applications to examine the problems associated with the dose for patients caused by the generation of neutrons from high-energy photons used for cancer therapy. According to the experimental results, 0.35 mSv, 0.65 mSv 1.82 mSv of fast neutrons on average were generated from 1 Gy, 2 Gy and 5 Gy of photon irradiation, respectively, whereas 0.26 mSv, 0.56 mSv and 1.23 mSv of thermal neutrons were generated. Both fast neutrons and thermal neutrons produced an increase in the dose of neutrons generated with increasing irradiation dose. With in regard to the dose generated within and around the irradiation area of the photon rays, it was confirmed that more neutrons were generated within the irradiation area. A wedge filer was used to measure the generation of neutrons. According to the measurement results, the generation of neutrons increased when a wedge filter was used. When the SRS cone that required a high dose was used, more neutrons were detected than those in the previous experiment. When fast neutrons were used, 2.85 mSv neutrons on average were generated from 5 Gy of photon irradiation. When thermal neutrons were used, 1.37 mSv neutrons on average were generated from 5 Gy of photon irradiation. Overall, approximately 1.6 times and 1.12 times more fast and thermal neutrons, respectively, were generated than in the case of a general treatment with 5 Gy

  19. Geant4 Analysis of a Thermal Neutron Real-Time Imaging System

    Science.gov (United States)

    Datta, Arka; Hawari, Ayman I.

    2017-07-01

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

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

    Science.gov (United States)

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

    2018-01-01

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

  1. A neutron counter based on measurement of prompt gamma rays

    International Nuclear Information System (INIS)

    Alfassi, Z.B; Zlatin, T.; Manor, O.; Dubinsky, S.; German, U.

    2004-01-01

    A neutron dosimeter based on measurement of prompt gamma rays is composed of three main elements: a moderator of the fast neutrons, a converter which transforms the thermal neutrons into gamma rays (mostly by (n, γ) reaction), and the detector of gamma rays. Chung and co-workers studied the possibility to use a Germanium detector to measure the neutron dose equivalent rate in a mixed neutron-gamma field. They reported that both thermal and fast neutron doses could be evaluated by measuring the photo-peaks at 596 keV and 691 keV due to the reactions 72 Ge(n th , γ) and 73 Ge(n f , n' γ) respectively. Chao and Niu used a Ge detector covered with moderating material. Fast neutrons were moderated in a polyethylene cylinder and then captured in the germanium crystal, where they created the 596 keV γ photons, that were counted by the same Ge crystal. Another approach was taken by Ghanbari and Mohageghi, who used 10 B loaded polyethylene as moderator and converter. They measured the 478 keV photons, which are emitted from the excited state of the 7 Li produced by the 10 B (n, γ) 7 Li reaction. In all those studies, where converters were either Ge or 10 B, relatively low energy photons were produced and measured, which are in the range of high background. There are converters that can produce high energy (in the range from 4 MeV to 7 MeV), but the efficiency of the detectors in this energy range is very low. An optimal energy range considering the two contradicting requirements of low background and high counting efficiency is estimated to be over 1 MeV, up to about 2.5 MeV. The purpose of the present work was to find an improved combination of converter-detector system with maximum efficiency and signal to noise ratio

  2. Activation method for measuring the neutron spectra parameters. Computer software

    International Nuclear Information System (INIS)

    Efimov, B.V.; Ionov, V.S.; Konyaev, S.I.; Marin, S.V.

    2005-01-01

    The description of mathematical statement of a task for definition the spectral characteristics of neutron fields with use developed in RRC KI unified activation detectors (UKD) is resulted. The method of processing of results offered by authors activation measurements and calculation of the parameters used for an estimation of the neutron spectra characteristics is discussed. Features of processing of the experimental data received at measurements of activation with using UKD are considered. Activation detectors UKD contain a little bit specially the picked up isotopes giving at irradiation peaks scale of activity in the common spectrum scale of activity. Computing processing of results of the measurements is applied on definition of spectrum parameters for nuclear reactor installations with thermal and close to such power spectrum of neutrons. The example of the data processing, the measurements received at carrying out at RRC KI research reactor F-1 is resulted [ru

  3. Visualization and measurement of gas-liquid metal two-phase flow with large density difference using thermal neutrons as microscopic probes

    International Nuclear Information System (INIS)

    Saito, Y.; Hibiki, T.; Mishima, K.; Nishihara, H.; Yamamoto, A.; Kanda, K.; Tobita, Y.; Konishi, K.; Matsubayashi, M.

    1998-01-01

    In a core melt accident of a fast breeder reactor there is a possibility of boiling of fuel-steel mixture in the containment pool. In relation to safety evaluation on severe accident, it is indispensable to evaluate the recriticality of melted core. Gas-liquid two-phase flow with a large density difference is formed due to the boiling of fuel-steel mixture. Although the large density difference may affect the basic characteristics of two-phase flow, little work has been performed so far on two-phase flow with large density difference has not been performed well. In this study, visualization and void fraction measurement of gas-liquid metal two-phase flow were performed by using neutron radiography. The effect of the large density difference between gas and liquid phases on the basic flow characteristics of two-phase flow was clarified. (author)

  4. Measurement of the neutron capture cross section of 99Tc using ANNRI at J-PARC

    Science.gov (United States)

    Katabuchi, Tatsuya; Mizumoto, Motoharu; Igashira, Masayuki; Terada, Kazushi; Kimura, Atsushi; Nakamura, Shoji; Nakao, Taro; Iwamoto, Osamu; Iwamoto, Nobuyuki; Mizuyama, Kazuhito; Harada, Hideo; Hori, Jun-ich; Kino, Koichi

    2017-09-01

    The neutron capture cross section of 99Tc was measured using NaI(Tl) detectors of the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Japan Proton Accelerator Research Complex (J-PARC) in the energy range from thermal to the keV energy region. Preliminary results were presented and compared with previous measurements and evaluations.

  5. Neutron flux measurement utilizing Campbell technique

    International Nuclear Information System (INIS)

    Kropik, M.

    2000-01-01

    Application of the Campbell technique for the neutron flux measurement is described in the contribution. This technique utilizes the AC component (noise) of a neutron chamber signal rather than a usually used DC component. The Campbell theorem, originally discovered to describe noise behaviour of valves, explains that the root mean square of the AC component of the chamber signal is proportional to the neutron flux (reactor power). The quadratic dependence of the reactor power on the root mean square value usually permits to accomplish the whole current power range of the neutron flux measurement by only one channel. Further advantage of the Campbell technique is that large pulses of the response to neutrons are favoured over small pulses of the response to gamma rays in the ratio of their mean square charge transfer and thus, the Campbell technique provides an excellent gamma rays discrimination in the current operational range of a neutron chamber. The neutron flux measurement channel using state of the art components was designed and put into operation. Its linearity, accuracy, dynamic range, time response and gamma discrimination were tested on the VR-1 nuclear reactor in Prague, and behaviour under high neutron flux (accident conditions) was tested on the TRIGA nuclear reactor in Vienna. (author)

  6. Determination of neutron generation time in miniature neutron source reactor by measurement of neutronics transfer function

    International Nuclear Information System (INIS)

    Hainoun, A.; Khamis, I.

    2000-01-01

    The prompt neutron generation time Λ and the total effective fraction of delayed neutrons (including the effect of photoneutrons) β have been experimentally determined for the miniature neutron source reactor (MNSR) of Syria. The neutron generation time was found by taking measurements of the reactor open-loop transfer function using newly devised reactivity-step-ejection method by the reactor pneumatic rabbit system. Small reactivity perturbations i.e. step changes of reactivity starting from steady state, were introduced into the reactor during operation at low power level i.e. zero-power. Relative neutron flux and reactivity versus time were obtained. Using transfer function analysis as well as least square fitting techniques and measuring the delayed neutrons fraction, the neutron generation time was determined to be 74.6±1.57 μs. Using the prompt jump approximation of neutron flux, the total effective fraction of delayed neutrons was measured and found to be 0.00783±0.00017. Measured values of Λ and β were found to be very consistent with calculated ones reported in the safety analysis report. (orig.)

  7. Lethal action of thermal neutron to living cells

    International Nuclear Information System (INIS)

    Kawai, Kenichi; Akahoshi, Mitsuhiko

    1975-01-01

    Using Amoeba Proteus (DNA content was 5x10 -10 g/cell), action mechanism of thermal neutron was described by using death of cells as an index. Using mononuclear type (M-type) and polynuclear type (P-type) of Amoeba, investigation was made as to whether the number of nucleus contained in cell was related to death of cell by radiation. As compared to gamma-ray, marked difference between sensitivity of both types of Amoeba was not recognized. It was suggested that nucleus was important as a place of death of cell by thermal neutron. In order to demonstrate that reaction of 31 P(n, γ) 32 Pwas a main cause of lethality, the following experiment was carried out. 1) In order to investigate that radical reaction was not a main cause of lethal action by thermal neutron, increase and decrease of radical reaction were observed by adding AET, Cysteine and BudR. As to thermal neutron action, effect of its chemical factors did not appeared. 2) In order to demonstrate that nuclear reaction of constituent atom of living body played an important part in death of cell, increase and decrease of nuclear reaction within cell were observed by taking Na 2 , B 4 and O 7 into Amoeba. It was suggested that nuclear reaction of cell constituent atom itself was important in death of cell. 3) Potentiation effect of gamma-ray and thermal neutron was investigated and its mechanism was discussed. (K. Serizawa)

  8. Inventory verification measurements using neutron multiplicity counting

    International Nuclear Information System (INIS)

    Ensslin, N.; Foster, L.A.; Harker, W.C.; Krick, M.S.; Langner, D.G.

    1998-01-01

    This paper describes a series of neutron multiplicity measurements of large plutonium samples at the Los Alamos Plutonium Facility. The measurements were corrected for bias caused by neutron energy spectrum shifts and nonuniform multiplication, and are compared with calorimetry/isotopics. The results show that multiplicity counting can increase measurement throughput and yield good verification results for some inventory categories. The authors provide recommendations on the future application of the technique to inventory verification

  9. Ultra Low Level Environmental Neutron Measurements Using Superheated Droplet Detectors

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  10. Measurements of anomalous neutron transport in bulk graphite

    International Nuclear Information System (INIS)

    Bowman, C.D.; Smith, G.A.; Vogelaar, B.; Howell, C.R.; Bilpuch, E.G.; Tornow, W.

    2003-01-01

    The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)

  11. Thermal neutron scattering studies of condensed matter under high pressures

    International Nuclear Information System (INIS)

    Carlile, C.J.; Salter, D.C.

    1978-01-01

    Although temperature has been used as a thermodynamic variable for samples in thermal neutron scattering experiments since the inception of the neutron technique, it is only in the last decade that high pressures have been utilised for this purpose. In the paper the problems particular to this field of work are outlined and a review is made of the types of high-pressure cells used and the scientific results obtained from the experiments. 103 references. (author)

  12. Delayed neutron measurements for 232Th neutron-induced fission

    Science.gov (United States)

    Ledoux, X.; Doré, D.; Mosconi, M.; Nolte, R.; Roettger, S.

    2010-10-01

    Delayed neutrons (DN) play an important role in nuclear reactor physics. Innovative critical reactor studies bring to light the need of new DN yields data. For the Th fuel cycle, according to the OECD recommendation, DN of the 232Th is needed with an accuracy of 5%. In the literature, significant discrepancies were observed for energies below 4 MeV and data are dispersed around 14 MeV. Therefore, a programme has been undertaken by CEA in order to measure DN yields from 232Th with incident neutron energies from 2 to 16 MeV. In this paper, the experimental setup will be described and preliminary results obtained at the PTB Ion Accelerator Facility of Braunschweig for incident neutron beam energies of 2, 3, 4, 6, 7, 10 and 16 MeV will be presented.

  13. Three-dimensional boron particle loaded thermal neutron detector

    Science.gov (United States)

    Nikolic, Rebecca J.; Conway, Adam M.; Graff, Robert T.; Kuntz, Joshua D.; Reinhardt, Catherine; Voss, Lars F.; Cheung, Chin Li; Heineck, Daniel

    2014-09-09

    Three-dimensional boron particle loaded thermal neutron detectors utilize neutron sensitive conversion materials in the form of nano-powders and micro-sized particles, as opposed to thin films, suspensions, paraffin, etc. More specifically, methods to infiltrate, intersperse and embed the neutron nano-powders to form two-dimensional and/or three-dimensional charge sensitive platforms are specified. The use of nano-powders enables conformal contact with the entire charge-collecting structure regardless of its shape or configuration.

  14. Thermal neutron dose calculation in synovium membrane for BNCS

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  15. Dual neutron flux/temperature measurement sensor

    Science.gov (United States)

    Mihalczo, John T.; Simpson, Marc L.; McElhaney, Stephanie A.

    1994-01-01

    Simultaneous measurement of neutron flux and temperature is provided by a single sensor which includes a phosphor mixture having two principal constituents. The first constituent is a neutron sensitive 6LiF and the second is a rare-earth activated Y203 thermophosphor. The mixture is coated on the end of a fiber optic, while the opposite end of the fiber optic is coupled to a light detector. The detected light scintillations are quantified for neutron flux determination, and the decay is measured for temperature determination.

  16. Introduction to the theory of thermal neutron scattering

    CERN Document Server

    Squires, G L

    2012-01-01

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

  17. Targets for bulk hydrogen analysis using thermal neutrons

    CERN Document Server

    Csikai, J; Buczko, C M

    2002-01-01

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

  18. Neutron Imaging Calibration to Measure Void Fraction

    Energy Technology Data Exchange (ETDEWEB)

    Geoghegan, Patrick J [ORNL; Bilheux, Hassina Z [ORNL; Sharma, Vishaldeep [ORNL; Fricke, Brian A [ORNL

    2015-01-01

    Void fraction is an intuitive parameter that describes the fraction of vapor in a two-phase flow. It appears as a key variable in most heat transfer and pressure drop correlations used to design evaporating and condensing heat exchangers, as well as determining charge inventory in refrigeration systems. Void fraction measurement is not straightforward, however, and assumptions on the invasiveness of the measuring technique must be made. Neutron radiography or neutron imaging has the potential to be a truly non-invasive void fraction measuring technique but has until recently only offered qualitative descriptions of two-phase flow, in terms of flow maldistributions, for example. This paper describes the calibration approach necessary to employ neutron imaging to measure steady-state void fraction. Experiments were conducted at the High Flux Isotope Reactor (HFIR) Cold Guide 1D neutron imaging facility at Oak Ridge National Laboratory (ORNL), Oak Ridge, TN, USA.

  19. Isotopic characterization and thermal neutron flux determination of a PuBe neutron source.

    Science.gov (United States)

    Purty, Ravi Ankit; Akanchha; Prasad, Shikha

    2017-07-01

    The Indian Institute of Technology Kanpur (IIT Kanpur) possesses a PuBe neutron source facility with an initial activity of 5 Ci, dated September 1966 (nearly 50 years ago). An understanding of the present activity and the rate of its change will allow implementation of proper radiological safety procedures and future radiological safety planning. Knowing the absolute neutron flux will help us in future neutron activation studies. These details are also important to ensure proper security precautions. In our work, we attempt to identify the isotopic composition to determine the rate of change of the source and the absolute thermal neutron flux of plutonium beryllium (PuBe) sample at IIT Kanpur. We have used gamma-ray spectroscopy for determining the isotopic composition of the PuBe neutron source. After utilizing gamma-ray spectroscopy it is found that the source is composed of 239 Pu and a small amount of 241 Am is present as an impurity. The mass ratio of 241 Am to 239 Pu is found to be approximately 18.1µg/g with an uncertainty of 1.39%. Delayed gamma neutron activation analysis (DGNAA) is used to determine the thermal neutron flux of the same PuBe neutron source using copper, cobalt, nickel and cadmium samples. The average thermal neutron flux as calculated from DGNAA is approximately 1.27×10 3 n/(cm 2 -s) at 1cm above the PuBe neutron source. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Neutron background measurements at China Jinping underground laboratory with a Bonner multi-sphere spectrometer

    Science.gov (United States)

    Hu, Qingdong; Ma, Hao; Zeng, Zhi; Cheng, Jianping; Chen, Yunhua; He, Shengming; Li, Junli; Shen, Manbin; Wu, Shiyong; Yue, Qian; Yue, Jianfeng; Zhang, Hui

    2017-07-01

    The neutron background spectrum from thermal neutron to 20 MeV fast neutron was measured at the first experimental hall of China Jinping underground laboratory with a Bonner multi-sphere spectrometer. The measurement system was validated by a 252Cf source and inconformity was corrected. Due to micro charge discharge, the dataset was screened and background from the steel of the detectors was estimated by MC simulation. Based on genetic algorithm we obtained the energy distribution of the neutron and the total flux of neutron was (2.69±1.02) ×10-5 cm-2 s-1.

  1. Neutron Flux Measurement Produced by BNCT Target using Proton Beam

    International Nuclear Information System (INIS)

    Ha, Jang Ho; Kim, Yong Kyun; Chai, Jong Seo; Kim, Jong Kyung

    2005-01-01

    We are investigating neutron production target system performance for boron captured neutron therapy (BNCT). The epithermal neutron is useful for this therapy and in present study we performed a simple method to measure neutron flux and energy, which are important for the accurate cancer therapy. The simple method and result of neutron flux and energy measurement experiment are presented

  2. Neutron nuclear data measurements for criticality safety

    Science.gov (United States)

    Guber, Klaus; Paradela, Carlos; Heyse, Jan; Kopecky, Stefan; Schillebeeckx, Peter; Siegler, Peter

    2017-09-01

    To support the US Department of Energy Nuclear Criticality Safety Program, neutron-induced cross section experiments were performed at the Geel Electron Linear Accelerator of the Joint Research Center Site Geel, European Union. Neutron capture and transmission measurements were carried out using metallic natural cerium and vanadium samples. Together with existing data, the measured data will be used for a new evaluation and will be submitted with covariances to the ENDF/B nuclear data library.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-06-24

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

  4. Gel dosimeters as useful dose and thermal-fluence detectors in Boron Neutron Capture Therapy (BNCT)

    International Nuclear Information System (INIS)

    Gambarini, G.; Valente, M.; Moss, R.L.; Daquino, G.G.; Nievaart, V.A.; Mariani, M.; Vanossi, E.; Carrara, M.

    2006-01-01

    The dosimetry method based on Fricke-Xylenol-Orange-infused gels in form of layers has shown noticeable potentiality for in-phantom or in-free-beam dose and thermal flux profiling and imaging in the high fluxes of thermal or epithermal neutrons utilised for boron neutron capture therapy (BNCT). Gel-dosimeters in form of layers give the possibility not only of obtaining spatial dose distributions but also of achieving measurements of each dose contribution in neutron fields. The discrimination of the various dose components is achieved by means of pixel-to-pixel manipulations of pairs of images obtained with gel-dosimeters having different isotopic composition. It is possible to place large dosimeters, detecting in such a way large dose images, because the layer geometry of dosimeters avoids sensitive variation of neutron transport due to the gel isotopic composition. Some results obtained after the last improvements of the method are reported. (Author)

  5. Epithermal neutron multiplicity counter (ENMC) - summary of measurement results

    Energy Technology Data Exchange (ETDEWEB)

    Henzlova, Daniela [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Rael, Carlos D [Los Alamos National Laboratory; Martinez, Isaac P [Los Alamos National Laboratory; Marlow, Johnna B [Los Alamos National Laboratory

    2010-10-20

    This presentation summarizes results acquired within LLNL-LANL collaboration on fast and thermal neutron detection techniques. Results acquired using Epithermal Neutron Multiplicity Counter (ENMC) are presented. Variety of sources relevant for nuclear safeguards were used in this comparison.

  6. Fast neutron measurements with 7Li and 6Li enriched CLYC scintillators

    International Nuclear Information System (INIS)

    Giaz, A.; Blasi, N.; Boiano, C.; Brambilla, S.; Camera, F.; Cattadori, C.; Ceruti, S.; Gramegna, F.; Marchi, T.; Mattei, I.; Mentana, A.; Million, B.; Pellegri, L.; Rebai, M.; Riboldi, S.; Salamida, F.; Tardocchi, M.

    2016-01-01

    The recently developed Cs 2 LiYCl 6 :Ce (CLYC) crystals are interesting scintillation detectors not only for their gamma energy resolution (<5% at 662 keV) but also for their capability to identify and measure the energy of both gamma rays and fast/thermal neutrons. The thermal neutrons were detected by the 6 Li(n,α)t reaction while for the fast neutrons the 35 Cl(n,p) 35 S and 35 Cl(n,α) 32 P neutron-capture reactions were exploited. The energy of the outgoing proton or α particle scales linearly with the incident neutron energy. The kinetic energy of the fast neutrons can be measured using both the Time Of Flight (TOF) technique and using the CLYC energy signal. In this work, the response to monochromatic fast neutrons (1.9–3.8 MeV) of two CLYC 1″×1″ crystals was measured using both the TOF and the energy signal. The observables were combined to identify fast neutrons, to subtract the thermal neutron background and to identify different fast neutron-capture reactions on 35 Cl, in other words to understand if the detected particle is an α or a proton. We performed a dedicated measurement at the CN accelerator facility of the INFN Legnaro National Laboratories (Italy), where the fast neutrons were produced by impinging a proton beam (4.5, 5.0 and 5.5 MeV) on a 7 LiF target. We tested a CLYC detector 6 Li-enriched at about 95%, which is ideal for thermal neutron measurements, in parallel with another CLYC detector 7 Li-enriched at more than 99%, which is suitable for fast neutron measurements.

  7. 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.

  8. Measurement of neutron dose in the compensator IMRT treatment.

    Science.gov (United States)

    Rezaian, Abbas; Nedaie, Hassan Ali; Banaee, Nooshin

    2017-10-01

    A radiation treatment delivery technique, intensity modulated radiation therapy (IMRT), has found widespread use in the treatment of cancers. One of IMRT implementing methods is IMRT compensator based, which the modulation are done by high Z materials. When photons with energies higher than 8MV interact with high Z material in path, Photoneutrons are produced. In this study, the effect of compensator on photoneutron production was investigated. The Monte Carlo code MCNPX was used to calculate the neutron dose equivalent as a function of the depth in phantom with and without compensator. Measurements were made using CR-39 track-etched detectors. CR-39 detectors, were cut in dimensions of 2.5×2.5 cm 2 by laser, placed in different depths of slab phantom and then irradiated by 18MV photons. Same procedure was performed with the compensator present and absent. The measured data were compared with MCNP calculations. In both experimental and simulation results, neutron dose equivalent when compensator used, was less than non-compensator field. The calculated neutron dose equivalent was maximum at surface and decreased exponentially by increasing depth, but in experimental data, the neutron dose equivalent reached a maximum at approximately 3cm depth in the phantom and beyond which decreased with depth.CR-39 calibration was carried out in air, by considering that neutron energy spectrum changes toward thermal neutrons by depth in phantom increasing, it is suggested that for measuring equivalent neutron dose at phantom depth, should have proper neutron calibration in terms of energy spectrum. Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. The fluctuating ribosome: thermal molecular dynamics characterized by neutron scattering

    Science.gov (United States)

    Zaccai, Giuseppe; Natali, Francesca; Peters, Judith; Řihová, Martina; Zimmerman, Ella; Ollivier, J.; Combet, J.; Maurel, Marie-Christine; Bashan, Anat; Yonath, Ada

    2016-11-01

    Conformational changes associated with ribosome function have been identified by X-ray crystallography and cryo-electron microscopy. These methods, however, inform poorly on timescales. Neutron scattering is well adapted for direct measurements of thermal molecular dynamics, the ‘lubricant’ for the conformational fluctuations required for biological activity. The method was applied to compare water dynamics and conformational fluctuations in the 30 S and 50 S ribosomal subunits from Haloarcula marismortui, under high salt, stable conditions. Similar free and hydration water diffusion parameters are found for both subunits. With respect to the 50 S subunit, the 30 S is characterized by a softer force constant and larger mean square displacements (MSD), which would facilitate conformational adjustments required for messenger and transfer RNA binding. It has been shown previously that systems from mesophiles and extremophiles are adapted to have similar MSD under their respective physiological conditions. This suggests that the results presented are not specific to halophiles in high salt but a general property of ribosome dynamics under corresponding, active conditions. The current study opens new perspectives for neutron scattering characterization of component functional molecular dynamics within the ribosome.

  10. Measurements of residual radioactivity in neutron transmutation doped thermistors

    International Nuclear Information System (INIS)

    Alessandrello, A.; Brofferio, C.; Camin, D.V.; Cremonesi, O.; Fiorini, E.; Giuliani, A.; Pavan, M.; Pessina, G.; Previtali, E.; Zanotti, L.

    1994-01-01

    Germanium wafers exposed to intense neutron beams from a nuclear reactor to produce neutron transmutation doped thermistors have been analysed for residual radioactivity in view of their application in experiments on rare decays. Measurements have been routinely carried out for more than three years with germanium spectrometers of low intrinsic radioactivity operating in the Gran Sasso Underground Laboratory in order to suppress the background due to cosmic rays. Four long lived nuclei produced during the irradiation by the fast neutron component of the flux have been identified. It is shown that two of them will contribute considerably to the background in experiments on rare events carried out with thermal detectors, especially in the direct search of dark matter. These measurements also enable investigation of the presence of contaminants in the germanium wafer before irradiation. The sensitivity can be as low as 10 -14 g/g for some elements. ((orig.))

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

    Science.gov (United States)

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

    2018-03-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Science.gov (United States)

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

    2017-05-16

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

  14. Research on measurement of neutron flux in irradiation channels of research reactor

    International Nuclear Information System (INIS)

    Yin Zhitao; Lv Zheng; Wang Yulin; Zheng Wuqin

    2014-01-01

    Relative distribution of thermal neutron flux in the irradiation channel is measured by classical activation foil method. After that, on a representative point in the irradiation channel, neutron temperature and absolute neutron flux are also measured. Cadmium ratio correction method is used to check the experiment result in the end. Comparative analysis shows that the results from two different methods are agreed pretty well, which adds the credibility of experiment results. (authors)

  15. Visualization and measurement of gas-liquid metal two-phase flow with large density difference using thermal neutrons as microscopic probes

    International Nuclear Information System (INIS)

    Mishima, K.; Hibiki, T.; Saito, Y.; Nishihara, H.; Tobita, Y.; Konishi, K.; Matsubayashi, M.

    1999-01-01

    In a core melt accident of a fast breeder reactor, there is a possibility of boiling of the fuel-steel mixture in the containment pool. In relation to safety evaluation on severe accident, it is indispensable to evaluate the possibility of re-criticality of melted core. Gas-liquid two-phase flow with a large liquid-to-gas density ratio is formed due to the boiling of fuel-steel mixture. Although it is anticipated that the large density ratio may affect the basic characteristics of two-phase flow, little work has been performed so far on two-phase flow with a large liquid-to-gas density ratio. In this study, visualization and void fraction measurement of gas-liquid metal two-phase flow were performed by using neutron radiography and image processing techniques. Then, the effect of large density difference between gas and liquid phases on the basic flow characteristics of two-phase flow was clarified

  16. Measurements with the new PHE neutron survey instrument

    International Nuclear Information System (INIS)

    Eakins, J.S.; Tanner, R.J.; Hager, L.G.

    2014-01-01

    A novel design of survey instrument has been developed to accurately estimate ambient dose equivalent from neutrons with energies in the range from thermal to 20 MeV. The device features moderating and attenuating layers to ease measurement of fast and intermediate energy neutrons, combined with guides that channel low-energy neutrons to the single, central detector. A prototype of this device has been constructed and exposed to a set of calibration fields: the resulting measured responses are presented and discussed here, and compared against Monte Carlo data. A simple simulated workplace neutron field has also been developed to test the device. Measured response data have been determined for a prototype design of neutron survey instrument, using facilities at PHE and NPL. In general, the results demonstrated good directional invariance and agreed well with data obtained by Monte Carlo modelling, raising confidence in the accuracy of the response characteristics expected for the device. A simple simulated workplace field has also been developed and characterised, and the performance of the device assessed in it: agreement between measured and modelled results suggests that the device would behave as anticipated in real workplace fields. These performances will be investigated further in the future, as the design makes the transition from a research prototype to a commercially available instrument. (authors)

  17. MEASURING NEUTRON STAR RADII VIA PULSE PROFILE MODELING WITH NICER

    Energy Technology Data Exchange (ETDEWEB)

    Özel, Feryal; Psaltis, Dimitrios; Bauböck, Michi [Astronomy Department, University of Arizona, 933 N. Cherry Ave., Tucson, AZ 85721 (United States); Arzoumanian, Zaven [Center for Research and Exploration in Space Science and Technology/USRA, NASA Goddard Space Flight Center, Code 662, Greenbelt, MD 20771 (United States); Morsink, Sharon, E-mail: fozel@email.arizona.edu [Department of Physics, University of Alberta, 11455 Saskatchewan Drive, Edmonton, AB T6G 2E1 (Canada)

    2016-11-20

    The Neutron-star Interior Composition Explorer is an X-ray astrophysics payload that will be placed on the International Space Station . Its primary science goal is to measure with high accuracy the pulse profiles that arise from the non-uniform thermal surface emission of rotation-powered pulsars. Modeling general relativistic effects on the profiles will lead to measuring the radii of these neutron stars and to constraining their equation of state. Achieving this goal will depend, among other things, on accurate knowledge of the source, sky, and instrument backgrounds. We use here simple analytic estimates to quantify the level at which these backgrounds need to be known in order for the upcoming measurements to provide significant constraints on the properties of neutron stars. We show that, even in the minimal-information scenario, knowledge of the background at a few percent level for a background-to-source countrate ratio of 0.2 allows for a measurement of the neutron star compactness to better than 10% uncertainty for most of the parameter space. These constraints improve further when more realistic assumptions are made about the neutron star emission and spin, and when additional information about the source itself, such as its mass or distance, are incorporated.

  18. VHTR core modeling: coupling between neutronic and thermal-hydraulics

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  19. Neutron Energy Measurements in Emergency Response Applications

    Energy Technology Data Exchange (ETDEWEB)

    Sanjoy Mukhopadhyay, Paul Guss, Michael Hornish, Scott Wilde, Tom Stampahar, Michael Reed

    2009-09-11

    We present significant results in recent advances in the measurement of neutron energy. Neutron energy measurements are a small but significant part of radiological emergency response applications. Mission critical information can be obtained by analyzing the neutron energy given off from radioactive materials. In the case of searching for special nuclear materials, neutron energy information from an unknown source can be of importance. At the Remote Sensing Laboratory (RSL) of National Security Technologies, LLC, a series of materials, viz., liquid organic scintillator (LOS), Lithium Gadolinium Borate (LGB) or Li6Gd(BO3)3 in a plastic matrix, a recently developed crystal of Cesium Lithium Yttrium Chloride, Cs2LiYCl6: Ce (called CLYC)[1], and normal plastic scintillator (BC-408) with 3He tubes have been used to study their effectiveness as a portable neutron energy spectrometer. Comparisons illustrating the strengths of the various materials will be provided. Of these materials, LGB offers the ability to tailor its response to the neutron spectrum by varying the isotopic composition of the key constituents (Lithium, Gadolinium [Yttrium], and Boron). All three of the constituent elements possess large neutron capture cross section isotopes for highly exothermic reactions. These compounds of composition Li6Gd(Y)(BO3)3 can be activated by Cerium ions Ce3+. CLYC, on the other hand, has a remarkable gamma response in addition to superb neutron discrimination, comparable to that of Europium-doped Lithium Iodide (6LiI: Eu). Comparing these two materials, CLYC has higher light output (4500 phe/MeV) than that from 6LiI: Eu and shows better energy resolution for both gamma and neutron pulse heights. Using CLYC, gamma energy pulses can be discriminated from the neutron signals by simple pulse height separation. For the cases of both LGB and LOS, careful pulse shape discrimination is needed to separate the gamma energy signals from neutron pulses. Both analog and digital

  20. Thermal diffusivity effect in opto-thermal skin measurements

    International Nuclear Information System (INIS)

    Xiao, P; Imhof, R E; Cui, Y; Ciortea, L I; Berg, E P

    2010-01-01

    We present our latest study on the thermal diffusivity effect in opto-thermal skin measurements. We discuss how thermal diffusivity affects the shape of opto-thermal signal, and how to measure thermal diffusivity in opto-thermal measurements of arbitrary sample surfaces. We also present a mathematical model for a thermally gradient material, and its corresponding opto-thermal signal. Finally, we show some of our latest experimental results of this thermal diffusivity effect study.

  1. Measurement of fast neutron background in SAGE

    CERN Document Server

    Abdurashitov, J N; Kalikhov, A V; Matushko, V L; Shikhin, A A; Yants, V E; Zaborskaia, O S

    2002-01-01

    The spectrometer intended for direct measurements of ultra low fluxes of fast neutrons is described. It is sensitive to neutron fluxes of 10 sup - sup 7 cm sup - sup 2 s sup - sup 1 and lower. The detection efficiency of fast neutrons with simultaneous energy measurement was determined from Monte-Carlo simulation to be equal to 0.11 +- 0.01. The background counting rate in the detector corresponds to a neutron flux of (6.5 +- 2.1) x 10 sup - sup 7 cm sup - sup 2 s sup - sup 1 in the range 1.0-11.0 MeV. The natural neutron flux from the surrounding mine rock at the depth of 4700 meters of water equivalent was measured to be (7.3 +- 2.4) x 10 sup - sup 7 cm sup - sup 2 s sup - sup 1 in the range 1.0-11.0 MeV. The flux of fast neutrons in the SAGE main room was measured to be < 2.3 x 10 sup - sup 7 cm sup - sup 2 s sup - sup 1 in 1.0-11.0 MeV energy range.

  2. Lifetime measurement of prompt neutrons using the neutronic noise analysis

    International Nuclear Information System (INIS)

    Ortiz Servin, J.J.

    1992-01-01

    The purpose of this work is to estimate the life of the prompt neutrons, i, of a nuclear reactor utilizing the neutron noise analysis. This technique carry to development of mathematical model that is valid for lower powers reactor. The equation resulting convey to the observation about power spectrum behaviour respect to the frecquency. In this case, the reactor in study is the Triga Mark III of Nuclear Center of Mexico that it was provided of fission chambers for register the neutron fluxes. These fluxes was digitized and storage in computer disc as signals dependents of time, for later apply the Fourier Transformation and obtain the spectras. The spectras measured to different reactor powers were adjusted to the development equation before, using the method of square minimum and so estimate the parameter i. The analysis of results throw a value of 22.73 +/- 0.92 μs. On the other hand, the calculate value to the resolve the kinetic equation of reactor defer in lower than 4 % about the estimate. Of this, it concludes that the model utilized is trusty with a good mistake margin, moreover of that the technique of Neutron Noise analysis demonstrate be competitive (Author)

  3. D-D fusion neutron-spectra measurements and ion-temperature determination at Alcator C

    International Nuclear Information System (INIS)

    Fisher, W.A.

    1983-02-01

    A neutron spectrometer system has been designed, assembled, and used to measure the D-D neutron spectrum at Alcator C. The design of the shielding and collimation was critical to the successful measurement of the spectrum and involved an integral approach in which the neutronics of the Alcator C was exploited to obtain a successful system. The system consists of a 3 He ionization chamber mounted in a multi-component shield system. Essentially the outermost part of the shield and collimator has been designed to moderate the MeV-range neutrons to thermal energies. The inner part of the shield is designed to capture the thermalized neutrons with a minimum of gamma production. As a result, measurements during plasma discharges indicate that the ratio of the number of counts in the 2.45-MeV peak to the total number of neutron counts in the ion chamber is 1.67

  4. Residual stress measurement at Budapest Neutron Center

    International Nuclear Information System (INIS)

    Gyula, T.

    2005-01-01

    The use of residual stress measurements of different construction element and recent possibilities of Budapest Neutron Centre are presented. The details investigated already: gas turbine wheel, axial compressor blade, turbine blade and plastically deformed stainless steel. We demonstrated the use of a neutron scattering (SANS, residual stress, diffraction) for the materials behavior investigation in order to analyze the processes going on under the different mechanical loading. The direction of possible instrumental development is presented. (author)

  5. The Response of Alanine Dosimeters in Thermal Neutron Fields

    DEFF Research Database (Denmark)

    Schmitz, T.; Bassler, Niels; Sharpe, P.

    Purpose: Boron Neutron Capture Therapy (BNCT) is a special kind of particle therapy, based on the neutron induced fission of the boron isotope 10B [1]. We have performed dosimetry experiments on the mixed neutron and gamma fields at the TRIGA Mark II research reactor in Mainz. Commonly, dosimetry...... in such fields is realized by foil activation and ion chambers [2]. Here we investigate alanine as an easier and more robust alternative dosimeter. Methods: We have performed four phantom experiments at the TRIGA Mark II research reactor in Mainz [3], in a predominantly thermal neutron field with a strong gamma...... response of all pellets could be reproduced by calculations within a uncertainty of 5 %. For all experiments three dose components have been separated. A proton dose is generated in the 14N(n,p)14C reaction. Secondary gammas are generated by various (n,γ) reactions, dominated by the 2.2 MeV photon from...

  6. Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Liu, H.B.

    1996-01-01

    The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 10 10 n/cm 2 ·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10 -11 cGy·cm 2 /n th ) and fast neutron (0.58 x 10 -11 cGy·cm 2 /n th ) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed

  7. Specimen environments in thermal neutron scattering experiments

    International Nuclear Information System (INIS)

    Cebula, D.J.

    1980-11-01

    This report is an attempt to collect into one place outline information concerning the techniques used and basic design of sample environment apparatus employed in neutron scattering experiments. Preliminary recommendations for the specimen environment programme of the SNS are presented. The general conclusion reached is that effort should be devoted towards improving reliability and efficiency of operation of specimen environment apparatus and developing systems which are robust and easy to use, rather than achieving performance at the limits of technology. (author)

  8. Neutron flux measurement by mobile detectors

    International Nuclear Information System (INIS)

    Verchain, M.

    1987-01-01

    Various incore instrumentation systems and their technological evolution are first reviewed. Then, for 1300 MWe PWR nuclear power plant, temperature and neutron flux measurement are described. Mobile fission chambers, with their large measuring range and accurate location allow a good knowledge of the core. Other incore measures are possible because of flux detector thimble tubes inserted in the reactor core [fr

  9. Neutron-capture-activation cross sections of 9496Zr and 98100Mo at thermal and 30 keV energy

    International Nuclear Information System (INIS)

    Wyrick, J.M.; Poenitz, W.P.

    1982-01-01

    Neutron-capture cross sections of 94 96 Zr and 98 100 Mo were measured relative to the standard-capture cross section of gold at thermal and 30 keV neutron energies using the activation technique. The reported values are based upon available decay-scheme information

  10. Prompt gamma-rays from thermal-neutron capture

    International Nuclear Information System (INIS)

    Lone, M.A.; Leavitt, R.A.; Harrison, D.A.; Lemmel, H.D.

    1989-04-01

    This document describes format and contents of a nuclear data library on magnetic tape which lists prompt gamma rays from thermal-neutron capture evaluated by M.A. Lone et al. The magnetic tape is available, costfree, from the IAEA Nuclear Data Section. (author)

  11. Thermalization time in a model of neutron star

    OpenAIRE

    Ducomet, B.; Nečasová, Š. (Šárka)

    2011-01-01

    We consider an initial boundary value problem for the equation describing heat conduction in a spherical model of neutron star considered by Lattimer et al. We estimate the asymptotic decay of the solution, which provides a plausible estimate for a "thermalization time" for the system.

  12. Response of neutron-irradiated RPV steels to thermal annealing

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

    1997-03-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels.

  13. Thermal neutron detection using alkali halide scintillators with Li-6 and pulse shape discrimination

    Energy Technology Data Exchange (ETDEWEB)

    Brubaker, Erik [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dibble, Dean C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mengesha, Wondwosen [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Yang, Pin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2013-09-01

    An ideal 3He detector replacement for the near- to medium-term future will use materials that are easy to produce and well understood, while maintaining thermal neutron detection efficiency and gamma rejection close to the 3He standard. Toward this end, we investigated the use of standard alkali halide scintillators interfaced with 6Li and read out with photomultiplier tubes (PMTs). Thermal neutrons are captured on 6Li with high efficiency, emitting high-energy and triton (3H) reaction products. These particles deposit energy in the scintillator, providing a thermal neutron signal; discrimination against gamma interactions is possible via pulse shape discrimination (PSD), since heavy particles produce faster pulses in alkali halide crystals. We constructed and tested two classes of detectors based on this concept. In one case 6Li is used as a dopant in polycrystalline NaI; in the other case a thin Li foil is used as a conversion layer. In the configurations studied here, these systems are sensitive to both gamma and neutron radiation, with discrimination between the two and good energy resolution for gamma spectroscopy. We present results from our investigations, including measurements of the neutron efficiency and gamma rejection for the two detector types. We also show a comparison with Cs2LiYCl6:Ce (CLYC), which is emerging as the standard scintillator for simultaneous gamma and thermal neutron detection, and also allows PSD. We conclude that 6Li foil with CsI scintillating crystals has near-term promise as a thermal neutron detector in applications previously dominated by 3He detectors. The other approach, 6Li-doped alkali halides, has some potential, but require more work to understand material properties and improve fabrication processes.

  14. The transmission of thermal neutrons along air filled ducts in water

    International Nuclear Information System (INIS)

    Piercey, D.C.

    1962-06-01

    Predictions and measurements of thermal neutron transmission along air filled ducts in water have been made. To aid the analysis, the ducts were shielded in various ways using cadmium sheet. The predictions were, in general, in agreement to better than a factor 4 for ducts up to 200 duct radii in length. (author)

  15. Measurement and simulation for a complementary imaging with the neutron and X-ray beams

    Directory of Open Access Journals (Sweden)

    Hara Kaoru Y.

    2017-01-01

    Full Text Available By using a composite source system, we measured radiographs of the thermal neutron and keV X-ray in the 45-MeV electron linear accelerator facility at Hokkaido University. The source system provides the alternative beam of neutron and X-ray by switching the production target onto the electron beam axis. In the measurement to demonstrate a complementary imaging, the detector based on a vacuum-tube type neutron color image intensifier was applied to the both beams for dual-purpose. On the other hand, for reducing background in a neutron transmission spectrum, test measurements using a gadolinium-type neutron grid were performed with a cold neutron source at Hokkaido University. In addition, the simulations of the neutron and X-ray transmissions for various substances were performed using the PHITS code. A data analysis procedure for estimating the substance of sample was investigated through the simulations.

  16. Measurement and simulation for a complementary imaging with the neutron and X-ray beams

    Science.gov (United States)

    Hara, Kaoru Y.; Sato, Hirotaka; Kamiyama, Takashi; Shinohara, Takenao

    2017-09-01

    By using a composite source system, we measured radiographs of the thermal neutron and keV X-ray in the 45-MeV electron linear accelerator facility at Hokkaido University. The source system provides the alternative beam of neutron and X-ray by switching the production target onto the electron beam axis. In the measurement to demonstrate a complementary imaging, the detector based on a vacuum-tube type neutron color image intensifier was applied to the both beams for dual-purpose. On the other hand, for reducing background in a neutron transmission spectrum, test measurements using a gadolinium-type neutron grid were performed with a cold neutron source at Hokkaido University. In addition, the simulations of the neutron and X-ray transmissions for various substances were performed using the PHITS code. A data analysis procedure for estimating the substance of sample was investigated through the simulations.

  17. Guidelines on calibration of neutron measuring devices

    International Nuclear Information System (INIS)

    Burger, G.

    1988-01-01

    The International Atomic Energy Agency and the World Health Organization have agreed to establish an IAEA/WHO Network of Secondary Standard Dosimetry Laboratories (SSDLs) in order to improve accuracy in applied radiation dosimetry throughout the world. These SSDLs must be equipped with, and maintain, secondary standard instruments, which have been calibrated against primary standards, and must be nominated by their governments for membership of the network. The majority of the existing SSDLs were established primarily to work with photon radiation (X-rays and gamma rays). Neutron sources are, however, increasingly being applied in industrial processes, research, nuclear power development and radiation biology and medicine. Thus, it is desirable that the SSDLs in countries using neutron sources on a regular basis should also fulfil the minimum requirements to calibrate neutron measuring devices. It is the primary purpose of this handbook to provide guidance on calibration of instruments for radiation protection. A calibration laboratory should also be in a position to calibrate instrumentation being used for the measurement of kerma and absorbed dose and their corresponding rates. This calibration is generally done with photons. In addition, since each neutron field is usually contaminated by photons produced in the source or by scatter in the surrounding media, neutron protection instrumentation has to be tested with respect to its intrinsic photon response. The laboratory will therefore need to possess equipment for photon calibration. This publication deals primarily with methods of applying radioactive neutron sources for calibration of instrumentation, and gives an indication of the space, manpower and facilities needed to fulfil the minimum requirements of a calibration laboratory for neutron work. It is intended to serve as a guide for centres about to start on neutron dosimetry standardization and calibration. 94 refs, 8 figs, 12 tabs

  18. Measurements in support of a neutron radiography facility for the SLOWPOKE-2 at RMC

    Science.gov (United States)

    Lewis, W. J.; Andrews, W. S.; Bennett, L. G. I.; Beeley, P. A.

    1990-12-01

    The feasibility of using the small (20 kWh) SLOWPOKE-2 research reactor for neutron radiography has been investigated. Although designed primarily for neutron activation analysis (NAA) and radioisotope production, the SLOWPOKE-2 at RMC was installed with a thermal column of heavy water in a sector of the water gap between the beryllium reflector and the reactor container. The thermal-neutron flux in the reactor pool, just beyond the reactor container, has been measured to be a factor of 2.7 higher than in similar locations remote from the thermal column. Placed in this location was a prototype neutron radiography facility, consisting of a beam tube (or collimator), vertically tangential to the reactor core, and a beam stop. Once the feasibility of using a SLOWPOKE-2 for neutron radiography was demonstrated, subsequent investigations were carried out to optimize the quality of the obtainable radiographs. Both neutron radiographic and thermal-neutron flux measurements were undertaken to determine the optimum placement and arrangement of the beam tube. A Category III (as defined by the ASTM Standard E545-86) neutron radiography facility was obtained, although Category I or II were indicated as feasible. Based on this prototype design and experimentation, a permanent neutron radiography facility will be installed. The design calculations have been finalized, construction blueprints have been prepared, and work is proceeding with the construction, installation and commissioning of the facility.

  19. Measurements in support of a neutron radiography facility for the SLOWPOKE-2 at RMC

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W.J.; Andrews, W.S.; Bennett, L.G.I.; Beeley, P.A. (Royal Military Coll. of Canada, Kingston, ON (Canada). SLOWPOKE-2 Facility Royal Military Coll. of Canada, Kingston, ON (Canada). Dept. of Chemistry and Chemical Engineering)

    1990-12-20

    The feasibility of using the small (20 kWh) SLOWPOKE-2 research reactor for neutron radiography has been investigated. Although designed primarily for neutron activation analysis (NAA) and radioisotope production, the SLOWPOKE-2 at RMC was installed with a thermal column of heavy water in a sector of the water gap between the beryllium reflector and the reactor container. The thermal-neutron flux in the reactor pool, just beyond the reactor container, has been measured to be a factor of 2.7 higher than in similar locations remote from the thermal column. Placed in this location was a prototype neutron radiography facility, consisting of a beam tube (or collimator), vertically tangential to the reactor core, and a beam stop. Once the feasibility of using a SLOWPOKE-2 for neutron radiography was demonstrated, subsequent investigations were carried out to optimize the quality of the obtainable radiographs. Both neutron radiographic and thermal-neutron flux measurements were undertaken to determine the optimum placement and arrangement of the beam tube. A Category III (as defined by the ASTM Standard E545-86) neutron radiography facility was obtained, although Category I or II were indicated as feasible. Based on this prototype design and experimentation, a permanent neutron radiography facility will be installed. The design calculations have been finalized, construction blueprints have been prepared, and work is proceeding with the construction, installation and commissioning of the facility. (orig.).

  20. Measurements in support of a neutron radiography facility for the SLOWPOKE-2 at RMC

    International Nuclear Information System (INIS)

    Lewis, W.J.; Andrews, W.S.; Bennett, L.G.I.; Beeley, P.A.; Royal Military Coll. of Canada, Kingston, ON

    1990-01-01

    The feasibility of using the small (20 kWh) SLOWPOKE-2 research reactor for neutron radiography has been investigated. Although designed primarily for neutron activation analysis (NAA) and radioisotope production, the SLOWPOKE-2 at RMC was installed with a thermal column of heavy water in a sector of the water gap between the beryllium reflector and the reactor container. The thermal-neutron flux in the reactor pool, just beyond the reactor container, has been measured to be a factor of 2.7 higher than in similar locations remote from the thermal column. Placed in this location was a prototype neutron radiography facility, consisting of a beam tube (or collimator), vertically tangential to the reactor core, and a beam stop. Once the feasibility of using a SLOWPOKE-2 for neutron radiography was demonstrated, subsequent investigations were carried out to optimize the quality of the obtainable radiographs. Both neutron radiographic and thermal-neutron flux measurements were undertaken to determine the optimum placement and arrangement of the beam tube. A Category III (as defined by the ASTM Standard E545-86) neutron radiography facility was obtained, although Category I or II were indicated as feasible. Based on this prototype design and experimentation, a permanent neutron radiography facility will be installed. The design calculations have been finalized, construction blueprints have been prepared, and work is proceeding with the construction, installation and commissioning of the facility. (orig.)

  1. Field Measurement of Thermal Inertia

    Science.gov (United States)

    Kahle, A. B.; Schieldge, J. P.; Marsh, S. E.

    1983-01-01

    Radiometric measurements determine thermal inertia for geologic materials. Measurements are correlated with data obtained by remote sensing, for discriminating varieties of rock encountered when exploring for minerals by aircraft or by satellites equipped with infrared scanners.

  2. Scattering of thermal neutron by the water molecule

    International Nuclear Information System (INIS)

    Rosa, L.P.

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

  3. Neutron interrogation of actinides with a 17 MeV electron accelerator and first results from photon and neutron interrogation non-simultaneous measurements combination

    Energy Technology Data Exchange (ETDEWEB)

    Sari, A., E-mail: adrien.sari@cea.fr [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, 91191 Gif-sur-Yvette Cedex (France); Carrel, F.; Lainé, F. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, 91191 Gif-sur-Yvette Cedex (France); Lyoussi, A. [CEA, DEN, 13108 Saint-Paul-Lez-Durance Cedex (France)

    2013-10-01

    In this article, we demonstrate the feasibility of neutron interrogation using the conversion target of a 17 MeV linear electron accelerator as a neutron generator. Signals from prompt neutrons, delayed neutrons, and delayed gamma-rays, emitted by both uranium and plutonium samples were analyzed. First results from photon and neutron interrogation non-simultaneous measurements combination are also reported in this paper. Feasibility of this technique is shown in the frame of the measurement of uranium enrichment. The latter was carried out by combining detection of prompt neutrons from thermal fission and delayed neutrons from photofission, and by combining delayed gamma-rays from thermal fission and delayed gamma-rays from photofission.

  4. Thermally optimized lithium neutron producing target design for accelerator-based BNCT

    International Nuclear Information System (INIS)

    Park Shane; Joo Hyeong Min; Jang Byeong Ill; Jeun, Gyoodong; Kim Jong Kyung; Chai, Jong Seo

    2006-01-01

    In accelerator-based Boron Neutron Capture Therapy (BNCT), 7 Li(p,n) 7 Be reaction is prevalently used as a neutron source. However, lithium has a very low melting point and poor thermal conductivity. Thus lithium target needs an efficient cooling. In this study, ways of increasing proton beam diameter and slanting target are proposed to reduce the heat density of lithium target. Thermal analysis on the lithium target design shows that water cooling is feasible if the proton beam diameter and target slopes are in the available range of the contour plots generated from this study. On the basis of the thermal analysis, the prototype of target system was designed and manufactured. Full-model thermal analysis and temperature measuring experiment were subsequently performed. The calculated temperature distribution coincided with the contour plots and the experimental results. These results will be used in the manufacture of the prototype accelerator-based BNCT facility at Hanyang University. (author)

  5. Proportional counter measurements in neutron therapy beams

    International Nuclear Information System (INIS)

    Menzel, H.G.

    1984-01-01

    Dosimetry for clinical neutron therapy requires a characterization of radiation quality in addition to the specification of absorbed dose. Generally, a very simple approach has been adopted which consists in separating total absorbed dose into neutron and photon fractions. This is explained by the requirement of clinical dosimetry to apply methods suitable for routine measurements, by the lack of generally accepted improved alternatives, and by the fact that radiation quality is only one of several problems in neutron therapy not sufficiently solved. Spectra measured with low-pressure tissue-equivalent proportional counters (experimental microdosimetry) provide a detailed description of the physical properties of the radiation field at neutron therapy facilities. These descriptions are suitable for explaining the influence of different parameters (collimation, field size, phantom) on radiation quality. Although the physical properties of the radiation field as described by the measured microdosimetric distributions and quantities are not the only properties relevant for radiation effects, in general there are reasons to believe that they provide a suitable radiation quality characterization for the limited range of applications in neutron therapy. (author)

  6. ZZ THERMGAM, Thermal Neutron Capture Gamma Spectroscopical Data Library

    International Nuclear Information System (INIS)

    1990-01-01

    Description of program or function: - Format: special format (ASCII) defined in documentation; - Nuclides: Z = 1 to 83 (except for Pm-61), 90 and 94. - Origin: Data published between 1968 and March 1980. This catalog of prompt gamma rays from thermal-neutron capture in natural elements is designed primarily as a convenient tool for analytical prompt gamma-ray spectroscopy. Other potential uses are the identification of background gamma rays and of chemical trace impurities in target materials used for (n,gamma) studies and for prediction of secondary gamma rays from materials irradiated by neutrons

  7. Method for manufacturing solid-state thermal neutron detectors with simultaneous high thermal neutron detection efficiency (>50%) and neutron to gamma discrimination (>1.0E4)

    Science.gov (United States)

    Nikolic, Rebecca J.; Conway, Adam M.; Heineck, Daniel; Voss, Lars F.; Wang, Tzu Fang; Shao, Qinghui

    2013-10-15

    Methods for manufacturing solid-state thermal neutron detectors with simultaneous high thermal neutron detection efficiency (>50%) and neutron to gamma discrimination (>10.sup.4) are provided. A structure is provided that includes a p+ region on a first side of an intrinsic region and an n+ region on a second side of the intrinsic region. The thickness of the intrinsic region is minimized to achieve a desired gamma discrimination factor of at least 1.0E+04. Material is removed from one of the p+ region or the n+ region and into the intrinsic layer to produce pillars with open space between each pillar. The open space is filed with a neutron sensitive material. An electrode is placed in contact with the pillars and another electrode is placed in contact with the side that is opposite of the intrinsic layer with respect to the first electrode.

  8. Thermal neutron scattering cross sections of beryllium and magnesium oxides

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  9. Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kunc, M.

    1974-01-01

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

  10. Method of measuring neutron spectra in JMTR exclusively used for irradiation and their evaluation

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi

    1983-01-01

    In the core of the Japan Materials Testing Reactor, about 60 capsules are irradiated. These are the material capsules for irradiating reactor materials, the fuel capsules for irradiating reactor fuel, the RI capsules for producing radioisotopes and so on. In the irradiation experiment using a reactor, the information on the neutron fluence is indispensable, and the neutron fluence in the irradiated specimen part is evaluated with a dosimeter or the nuclear calculation for the core of the JMTR. At the time of irradiating reactor materials, the dosimeter Fe-54 (n,p) Mn-54 is generally used for evaluating the neutron fluence more than 1 MeV. In the case of fuel irradiation, the thermal neutron fluence is evaluated with the dosimeter Co-59 (n,γ) Co-60. It is important to examine in detail neutron spectra by both calculation and experiment in the reactors exclusively used for irradiation such as the JMTR. The neutron irradiation field in the JMTR, neutron spectrum measuring experiment, the neutron flux monitors for standardizing data, the measurement of X-ray and gamma ray, neutron guess spectrum, the compilation of neutron cross section for SAND 2, and the unfolding of neutron spectra are reported. The degree of agreement of the neutron fluence more than 1 MeV by measurement and calculation was +- 10 to 20 %. (Kako, I.)

  11. Thermal conductivity of beryllium under low temperature high dose neutron irradiation

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  12. Residual stress measurement by neutron diffraction

    International Nuclear Information System (INIS)

    Akita, Koichi; Suzuki, Hiroshi

    2010-01-01

    Neutron diffraction method has great advantages, allowing us to determine the residual stress deep present within the bulk materials and components nondestructively. Therefore, the method has been applied to confirm the structural integrity of the actual mechanical components and structures and to improve the manufacturing process and strength reliability of the products. This article reviews the residual stress measurement methodology of neutron diffraction. It also refers to the appropriate treatments of diffraction plane, stress-free lattice spacing, coarse grain and surface error to obtain reliable results. Finally, a few applications are introduced to show the capabilities of the neutron stress measurement method for the studies on the strength and elasto-plastic behaviors of crystalline materials. (author)

  13. Thermal measurements and inverse techniques

    CERN Document Server

    Orlande, Helcio RB; Maillet, Denis; Cotta, Renato M

    2011-01-01

    With its uncommon presentation of instructional material regarding mathematical modeling, measurements, and solution of inverse problems, Thermal Measurements and Inverse Techniques is a one-stop reference for those dealing with various aspects of heat transfer. Progress in mathematical modeling of complex industrial and environmental systems has enabled numerical simulations of most physical phenomena. In addition, recent advances in thermal instrumentation and heat transfer modeling have improved experimental procedures and indirect measurements for heat transfer research of both natural phe

  14. Method for simultaneous measurement of borehole and formation neutron decay-times employing iterative fitting

    International Nuclear Information System (INIS)

    Schultz, W.E.

    1982-01-01

    A method is described of making in situ measurements of the thermal neutron decay time of earth formations in the vicinity of a wellbore. The borehole and earth formations in its vicinity are repetitively irradiated with pulsed fast neutrons and, during the intervals between pulses, capture gamma radiation is measured in at least four, non-overlapping, contiguous time intervals. A background radiation measurement is made between successive pulses and used to correct count-rates representative of thermal neutron populations in the borehole and the formations, the count-rates being generated during each of the time intervals. The background-corrected count-rate measurements are iteratively fitted to exponential curves using a least squares technique to simultaneously derive signals representing borehole component and formation component of the thermal neutron decay time. The signals are recorded as a function of borehole depth. (author)

  15. 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.)

  16. Prompt gamma-ray analysis using cold and thermal guided neutron beams

    International Nuclear Information System (INIS)

    Yonezawa, C.; Magara, M.; Hoshi, M.; Tachikawa, E.; Sawahata, H.; Ito, Y.

    1995-01-01

    A permanent and stand-alone neutron-induced prompt γ-ray analysis (PGA) system, usable at both cold and thermal neutron beam guides of JRR-3M has been constructed. The characteristics of the system, including neutron beam and γ-ray spectrometer were measured. Owing to the absence of fast neutrons and the low γ-ray background, analytical sensitivities and detection limits better than those in other PGA systems have been achieved. Analytical results of ten elements in Standard Reference Material of Coal Fly Ash agreed well with those obtained by other methods. Isotopic analysis of Ni and its application to accurate and precise determination of Ni by stable isotope dilution method were performed. (author) 14 refs.; 4 figs.; 1 tab

  17. Measurement of the C / H ratio using neutrons

    International Nuclear Information System (INIS)

    Martinelli, P.

    1960-01-01

    A probe made up of a Ra (α, n) Be neutron source and a proportional counter filled with boron trifluoride has been used to measure the C/H ratio in hydrocarbons. The intensity of the thermal neutron flux in the neighbourhood of the detector increases with the concentration of the hydrocarbon hydrogen surrounding it. By measuring the density it is possible to deduce the C/H ratio. It is thus possible to evaluate the C/H ratio with a precision equal to that given by the β-ray transmission method. The errors arising from the chemical nature of the hydrocarbon can be reduced to a minimum. This method has the advantage of allowing the measurement of the C/H ratio of hydrocarbons contained in recipients or thick steel tubing by means an independent portable apparatus. (author) [fr

  18. Measuring surface temperature of isolated neutron stars and related problems

    Science.gov (United States)

    Teter, Marcus Alton

    New and exciting results for measuring neutron star surface temperatures began with the successful launch of the Chandra X-ray observatory. Among these results are new detections of neutron star surface temperatures which have made it possible to seriously test neutron star thermal evolution theories. The important new temperature determination of the Vela pulsar (Pavlov, et al., 2001a) requires a non-standard cooling scenario to explain it. Apart from this result, we have measured PSR B1055-52's surface temperature in this thesis, determining that it can be explained by standard cooling with heating. Our spectral fit of the combined data from ROSAT and Chandra have shown that a three component model, two thermal blackbodies and an non-thermal power-law, is required to explain the data. Furthermore, our phase resolved spectroscopy has begun to shed light on the geometry of the hot spot on PSR B1055-52's surface as well as the structure of the magnetospheric radiation. Also, there is strong evidence for a thermal distribution over its surface. Most importantly, the fact that PSR B1055-52 does not have a hydrogen atmosphere has been firmly established. To reconcile these two key observations, on the Vela pulsar and PSR B1055-52, we tested neutron star cooling with neutrino processes including the Cooper pair neutrino emission process. Overall, it has been found that a phase change associated with pions being present in the cores of more massive neutron stars explains all current of the data. A transition from neutron matter to pion condensates in the central stellar core explains the difference between standard and non-standard cooling scenarios, because the superfluid suppression of pion cooling will reduce the emissivity of the pion direct URCA process substantially. A neutron star with a mass of [Special characters omitted.] with a medium stiffness equation of state and a T72 type neutron superfluid models the standard cooling case well. A neutron star of [Special

  19. Handheld dual thermal neutron detector and gamma-ray spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Stowe, Ashley C.; Burger, Arnold; Bhattacharya, Pijush; Tupitsyn, Yevgeniy

    2017-05-02

    A combined thermal neutron detector and gamma-ray spectrometer system, including: a first detection medium including a lithium chalcopyrite crystal operable for detecting neutrons; a gamma ray shielding material disposed adjacent to the first detection medium; a second detection medium including one of a doped metal halide, an elpasolite, and a high Z semiconductor scintillator crystal operable for detecting gamma rays; a neutron shielding material disposed adjacent to the second detection medium; and a photodetector coupled to the second detection medium also operable for detecting the gamma rays; wherein the first detection medium and the second detection medium do not overlap in an orthogonal plane to a radiation flux. Optionally, the first detection medium includes a .sup.6LiInSe.sub.2 crystal. Optionally, the second detection medium includes a SrI.sub.2(Eu) scintillation crystal.

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

    Science.gov (United States)

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

    2017-12-01

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

  1. Method for simultaneous measurement of borehole and formation neutron decay-times

    International Nuclear Information System (INIS)

    Smith, H.D.; Arnold, D.M.

    1982-01-01

    A method is described of making in situ measurements of the thermal neutron decay time of earth formations in the vicinity of a wellbore. The borehole and earth formations are irradiated, with pulsed fast neutrons and, during the interval between neutron pulses, capture gamma radiation is measured in at least four, non-overlapping, contiguous time intervals. Count-rates representative of thermal neutron populations in the borehole and the formations are made during each of the time intervals. A background radiation measurement for correcting the count-rates is preferably also periodically made. The count-rates are combined to derive simultaneously the formation and borehole neutron lifetime components which are recorded as a function of borehole depth. (author)

  2. Measurements of neutron capture cross sections

    International Nuclear Information System (INIS)

    Nakajima, Yutaka

    1984-01-01

    A review of measurement techniques for the neutron capture cross sections is presented. Sell transmission method, activation method, and prompt gamma-ray detection method are described using examples of capture cross section measurements. The capture cross section of 238 U measured by three different prompt gamma-ray detection methods (large liquid scintillator, Moxon-Rae detector, and pulse height weighting method) are compared and their discrepancies are resolved. A method how to derive the covariance is described. (author)

  3. Neutron residual stress measurements in linepipe

    Science.gov (United States)

    Law, Michael; Gnaëpel-Herold, Thomas; Luzin, Vladimir; Bowie, Graham

    2006-11-01

    Residual stresses in gas pipelines are generated by manufacturing and construction processes and may affect the subsequent pipe integrity. In the present work, the residual stresses in eight samples of linepipe were measured by neutron diffraction. Residual stresses changed with some coating processes. This has special implications in understanding and mitigating stress corrosion cracking, a major safety and economic problem in some gas pipelines.

  4. Neutron residual stress measurements in linepipe

    International Nuclear Information System (INIS)

    Law, Michael; Gnaepel-Herold, Thomas; Luzin, Vladimir; Bowie, Graham

    2006-01-01

    Residual stresses in gas pipelines are generated by manufacturing and construction processes and may affect the subsequent pipe integrity. In the present work, the residual stresses in eight samples of linepipe were measured by neutron diffraction. Residual stresses changed with some coating processes. This has special implications in understanding and mitigating stress corrosion cracking, a major safety and economic problem in some gas pipelines

  5. Measurement of the neutron electric dipole moment

    International Nuclear Information System (INIS)

    Dress, W.B.; Perrin, P.; Miller, P.D.; Pendlebury, J.M.; Ramsey, N.F.

    1975-01-01

    Experiments have been performed in view of improving the accuracy in measuring the electric dipole moment of the neutron (EDM). This EDM is written as eD where e is the electron charge and D the dipole length. The analysis of the data indicates that /D/ 24 cm with 90% confidence [fr

  6. Automated measurement of thermal conductivity

    Science.gov (United States)

    Hale, D. V.

    1969-01-01

    Testing technique permits accurate measurement of temperature-dependent thermal conductivity, by virtue of the small temperature differential required across a specimen. The permissible mean insulation temperature ranges from cryogenic to 10 degrees F for the insulation under test.

  7. Large subcriticality measurement by pulsed neutron method

    International Nuclear Information System (INIS)

    Yamane, Y.; Yoshida, A.; Nishina, K.; Kobayashi, K.; Kanda, K.

    1985-01-01

    To establish the method determining large subcriticalities in the field of nuclear criticality safety, the authors performed pulsed neutron experiments using the Kyoto University Critical Assembly (KUCA) at Research Reactor Institute, Kyoto University and the Cockcroft-Walton type accelerator attached to the assembly. The area-ratio method proposed by Sjoestrand was employed to evaluate subcriticalities from neutron decay curves measured. This method has the shortcomings that the neutron component due to a decay of delayed neutrons remarkably decreases as the subcriticality of an objective increases. To overcome the shortcoming, the authors increased the frequency of pulsed neutron generation. The integral-version of the area-ratio method proposed by Kosaly and Fisher was employed in addition in order to remove a contamination of spatial higher modes from the decay curve. The latter becomes significant as subcriticality increases. The largest subcriticality determined in the present experiments was 125.4 dollars, which was equal to 0.5111 in a multiplication factor. The calculational values evaluated by the computer code KENO-IV with 137 energy groups based on the Monte Carlo method agreed well with those experimental values

  8. Factors affecting neutron measurements and calculations. Part B. Effects of local environment

    International Nuclear Information System (INIS)

    Santoro, Robert T.; Kerr, George D.; Pace, Joseph V. III; Lillie, Richard A.

    2005-01-01

    One of the difficulties encountered by Hiroshima and Nagasaki atomic-bomb analysts was achieving agreement between measured and calculated neutron activation data, particularly for the cases of materials that were activated by thermal and epithermal neutrons. In DS86 (Kerr et al. 1987) and the work reported here, the adopted standard was to calculate neutron kerma and neutron activation at 1 m above flat, wet ground as a function of distance from the burst. This two-dimensional air-over-ground model does not account for local effects, because the air-ground interface is a flat surface that offers no local shielding by terrain fluctuations, houses, walls, etc. In addition, the ground composition is assumed to be the same at all ground ranges. The composition of the surface and local environment conditions can have profound impact on the thermal neutron response at 1 m above the ground surface. Depending on the composition of the surface materials and assumptions made for the presence of buildings and other structures, calculated responses might increase or decrease the neutron activation of materials by thermal and epithermal neutrons. The examination of measured neutron activation responses suggests that wide variations did indeed occur. Measured and calculated activation data were, in some cases, found to differ by as much as 10% to 50% depending on location and distance from the hypocenter. In other cases, measured activation data taken at the same ground range often showed differences among themselves. (J.P.N.)

  9. Rotational energy conversion and thermal evolution of neutron stars

    Science.gov (United States)

    Zhu, Cui; Zhou, Xia; Wang, Na

    2017-12-01

    Pulsars are rapidly spinning, strongly magnetized neutron stars. Their electromagnetic dipole radiation is usually assumed to be at the expense of the rotational energy. In this work, we consider a new channel through which rotational energy could be radiated away directly via neutrinos. With this new energy conversion channel, we can improve the chemical heating mechanism that originates in the deviation from β equilibrium due to spin-down compression. The improved chemical and thermal evolution equations with different magnetic field strengths are solved numerically. The results show that the new energy conversion channel could raise the surface temperature of neutron stars, especially for weak field stars at later stages of their evolution. Moreover, our results indicate that the new energy conversion channel induced by the non-equilibrium reaction processes should be taken into account in the study of thermal evolution. Supported by National Natural Science Foundation of China (11373006) and the West Light Foundation of Chinese Academy of Sciences (ZD201302)

  10. Neutron thermalization in quality control of asphalts content in mixtures for paving. Adaptation of nuclear densimeters for this purpose

    International Nuclear Information System (INIS)

    Bravo R, T.; Montanez M, P.O.

    1995-01-01

    This paper shows how the neutron source of the nuclear densimeters, used for measure the humidity, can be used for measuring and making the quality control of the asphalt percentage in mixtures used for street paving. The measures are based in the neutronic thermalization processes, because the hydrogen is the main part of chemical composition of the asphalts. A calibration method for the equipment is presented. (author). 6 refs, 3 figs, 3 tabs

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    Directory of Open Access Journals (Sweden)

    SANG IN KIM

    2014-04-01

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

  13. Activation experiment for concrete blocks using thermal neutrons

    Directory of Open Access Journals (Sweden)

    Okuno Koichi

    2017-01-01

    Full Text Available Activation experiments for ordinary concrete, colemanite-peridotite concrete, B4C-loaded concrete, and limestone concrete are carried out using thermal neutrons. The results reveal that the effective dose for gamma rays from activated nuclides of colemanite-peridotite concrete is lower than that for the other types of concrete. Therefore, colemanite-peridotite concrete is useful for reducing radiation exposure for workers.

  14. Electric and thermal conductivities of quenched neutron star crusts

    Science.gov (United States)

    Ogata, Shuji; Ichimaru, Setsuo

    1990-01-01

    The electric and thermal conductivities in the outer crustal matter of a neutron star quenched into a solid state by cooling are estimated using a Monte Carlo simulation of freezing transition for dense plasmas. The conductivities are calculated by the precise evaluation of the scattering integrals, using the procedure of Ichimaru et al. (1983) and Iyetomi and Ichimaru (1983). The results predict the conductivities lower, by a factor of about 3, than those with the single-phonon approximation.

  15. Optimization of Moderator Size of Thermal and Epithermal Neutron Source Based on a Compact Accelerator for Neutron Imaging

    Science.gov (United States)

    Hasemi, Hiroyuki; Kamiyama, Takashi; Sato, Hirotaka; Kino, Koichi; Kiyanagi, Yoshiaki; Nakajima, Ken

    A compact accelerator-driven neutron source has some advantages over a large accelerator facility in terms of accessibility and usability. Recently, the project to develop a non-destructive testing system for nuclear fuels by neutron imaging using a compact accelerator-driven neutron source has launched in Japan. In this project, the traditional neutron radiography and temperature imaging by neutron resonance absorption spectroscopy (N-RAS) have been studied. From the viewpoint of L/D, a high-brightness moderator is desirable for the neutron imaging. In this study, we investigated the dependence of moderator size on the source brightness and the pulse characteristics of the neutron by simulation calculations to design the moderator for imaging using thermal and epithermal neutrons. As a result, the optimal size of the moderator for the neutron imaging was 6∼7 cm in the energy region from 5 meV to 100 eV.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  17. Extended use of alanine irradiated in experimental reactor for combined gamma-and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of C-14 by LSC

    Czech Academy of Sciences Publication Activity Database

    Bartoníček, B.; Kučera, Jan; Světlík, Ivo; Viererbl, L.; Lahodová, Z.; Tomášková, Lenka; Cabalka, M.

    2014-01-01

    Roč. 93, NOV (2014), s. 52-56 ISSN 0969-8043 R&D Projects: GA TA ČR TA02010218 Institutional support: RVO:61389005 Keywords : reactor rediation * alanine/ESR dosimeter * C-14 * LSC simultaneous gamma * neutron dose * assessment Subject RIV: JF - Nuclear Energetics Impact factor: 1.231, year: 2014

  18. A technique of measuring neutron spectrum

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  19. Transparent lithiated polymer films for thermal neutron detection

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-11

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

  20. Neutron kinetics for system thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1996-01-01

    There is general agreement that for many light water reactor (LWR) calculations for licensing safety analysis, probabilistic risk assessment, operational support, and training, it is necessary to use a multidimensional neutron kinetics model coupled to a thermal-hydraulics model in order to obtain satisfactory results. This need coincides with the fact that in recent years there has been considerable research and development in this field, with modelers taking advantage of the increase in computing power that has become available. This progress has now led to coupling multidimensional neutron kinetics models to the nuclear steam supply system thermal hydraulics. This is not new since some coupled codes have always been available. What is new is that the coupling can now be done with very sophisticated models, and the planning of this coupling and the requisite modeling can take advantage of the experience of many code developers in many countries. The U.S. Nuclear Regulatory Commission and other organizations are in the process of reviewing the state of the art and making recommendations for future development. This paper summarizes one contribution to this review process: a review of the multidimensional neutron kinetics modeling, and ancillary modeling, which would be used in conjunction with system thermal-hydraulic models to perform core dynamics calculations

  1. Epi-thermal neutron diffraction of Gd alloy, (3)

    International Nuclear Information System (INIS)

    Watanabe, Noboru; Ishikawa, Yoshikazu; Sekine, Hisashi; Suzuki, Michio; Fujii, Hironobu.

    1975-01-01

    Epi-thermal neutron diffraction of Gd alloy was studied. A single crystal of GdZn was made by the Bridgman's method, and the degree of order was 1.00. One of the epi-thermal neutron diffraction patterns from the GdZn single crystal is shown. The nuclear scattering amplitude of Gd was obtained as bsub(Gd)=(1.15 +- 0.15) x 10 -12 cm, which is in good agreement with the value obtained from the comparison with the magnetic diffraction of Gdsub(0.9)Ysub(0.1). The epi-thermal neutron diffraction of Gdsub(0.85)Thsub(0.15) was also studied. The magnetic satellite line was clearly seen at low temperature. Higher order satellite line could not be observed. The magnetic structures in which the satellite is observed at are proper screw, cycloid, transverse sine wave and conical structures. The possibility of the conical structure was eliminated by the observed diffraction patterns at room temperature and at 80 0 K. The magnetic structure factors for the other three structures were obtained. At present, it is unable to determine the magnetic structure for Gdsub(0.85)Thsub(0.15). (Kato, T.)

  2. Velocity-space sensitivity of neutron spectrometry measurements

    DEFF Research Database (Denmark)

    Jacobsen, Asger Schou; Salewski, Mirko; Eriksson, J.

    2015-01-01

    Neutron emission spectrometry (NES) measures the energies of neutrons produced in fusion reactions. Here we present velocity-space weight functions for NES and neutron yield measurements. Weight functions show the sensitivity as well as the accessible regions in velocity space for a given range...... of the neutron energy spectrum. Combined with a calculated fast-ion distribution function, they determine the part of the distribution function producing detectable neutrons in a given neutron energy range. Furthermore, we construct a forward model based on weight functions capable of rapidly calculating neutron...

  3. Thermal-neutron fission cross section of 26. 1-min /sup 235/U/sup m/

    Energy Technology Data Exchange (ETDEWEB)

    Talbert W.L. Jr.; Starner, J.W.; Estep, R.J.; Balestrini, S.J.; Attrep M. Jr.; Efurd, D.W.; Roensch, F.R.

    1987-11-01

    The thermal-neutron fission cross section of /sup 235/U/sup m/ has been measured relative to the ground-state cross section. A rapid radiochemical separation procedure was developed to provide sizeable (10/sup 10/ to 10/sup 11/ atom) samples that were reasonably free of the parent /sup 239/Pu. From a series of eight measurements, the value of 1.42 +- 0.04 was obtained for the ratio sigma/sub m//sigma/sub g/.

  4. Temperature measurement by neutron resonance radiography

    International Nuclear Information System (INIS)

    Mayers, J.; Baciocco, G.; Hannon, A.C.

    1988-03-01

    We present a new data analysis technique for obtaining temperatures from neutron resonance radiography measurements. The technique is applied to measurements on tantalum and rhenium foils, a high temperature engineering alloy and a model catalytic system. Temperatures have been obtained to an accuracy of better than ± 1 0 C in the temperature range 21 0 C to 600 0 C. The results are in good agreement with theoretical simulations. (author)

  5. In-situ neutron diffraction measurements of temperature and stresses during friction stir welding of 6061-T6 aluminum alloy

    International Nuclear Information System (INIS)

    Woo, Wan Chuck; Feng, Zhili; Wang, Xun-Li; Brown, D.W.; Clausen, B.; An, Ke; Choo, Hahn; Hubbard, Camden R.; David, Stan A.

    2007-01-01

    The evolution of temperature and thermal stresses during friction stir welding of Al6061-T6 was investigated by means of in-situ, time-resolved neutron diffraction technique. A method is developed to deconvolute the temperature and stress from the lattice spacing changes measured by neutron diffraction. The deep penetration capability of neutrons made it possible for the first time to obtain the temperature and thermal stresses inside a friction stir weld

  6. Detection of low caloric power of coal by pulse fast-thermal neutron analysis

    International Nuclear Information System (INIS)

    Gu De-shan; Sang Hai-feng; Qiao Shuang; Liu Yu-ren, Liu Lin-mao; Jing Shi-wei; Chinese Academy of Sciences, Changchun

    2004-01-01

    Analysis method and principle of pulse fast-thermal neutron analysis (PFTNA) are introduced. A system for the measurement of low caloric power of coal by PFTNA is also presented. The 14 MeV pulse neutron generator and BGO detector and 4096 MCA were applied in this system. A multiple linear regression method applied to the data solved the interferential problem of multiple elements. The error of low caloric power between chemical analysis and experiment was less than 0.4 MJ/kg. (author)

  7. Neutron-Induced Failures in Semiconductor Devices

    Energy Technology Data Exchange (ETDEWEB)

    Wender, Stephen Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-04-06

    This slide presentation explores single event effect, environmental neutron flux, system response, the Los Alamos Neutron Science Center (LANSCE) neutron testing facility, examples of SEE measurements, and recent interest in thermal neutrons.

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

    International Nuclear Information System (INIS)

    Ritenour, R.L.

    1989-01-01

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

  9. A study on the linearity characteristics of neutron power measurement system for Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Tai Ki; Kim, Young Ki; Lee, Byung Chul; Park, Sang Jun

    1999-06-01

    It is briefly described the general principles of neutron detection and the method of neutron measurement in the nuclear reactor which neutron flux varies widely and gamma radiation also exists. Wide-range Fission Chamber System which is excellent in electrical and mechanical performances has been selected for neutron power measurement system for Hanaro. The linearity characteristics of neutron power signals is a critical factor of the reliability in reactor power control. In particular , the linearity of the log power signal, which covers 10 decade form 10 {sup -8} %FP to 200 %FP was a matter of primary concern during commissioning. In case of the linear power signal for reactor control at high power condition, the output signals were additionally analyzed in connection with the reactor thermal power and the delayed neutron signal from the primary pipe as well asthe output signal from the compensated ion chamber as a reference signal. (author). 13 refs., 7 tabs., 33 figs.

  10. A study on the linearity characteristics of neutron power measurement system for Hanaro

    International Nuclear Information System (INIS)

    Kang, Tai Ki; Kim, Young Ki; Lee, Byung Chul; Park, Sang Jun

    1999-06-01

    It is briefly described the general principles of neutron detection and the method of neutron measurement in the nuclear reactor which neutron flux varies widely and gamma radiation also exists. Wide-range Fission Chamber System which is excellent in electrical and mechanical performances has been selected for neutron power measurement system for Hanaro. The linearity characteristics of neutron power signals is a critical factor of the reliability in reactor power control. In particular , the linearity of the log power signal, which covers 10 decade form 10 -8 %FP to 200 %FP was a matter of primary concern during commissioning. In case of the linear power signal for reactor control at high power condition, the output signals were additionally analyzed in connection with the reactor thermal power and the delayed neutron signal from the primary pipe as well as the output signal from the compensated ion chamber as a reference signal. (author). 13 refs., 7 tabs., 33 figs

  11. Neutron-multiplication measurement instrument

    International Nuclear Information System (INIS)

    Nixon, K.V.; Dowdy, E.J.; France, S.W.; Millegan, D.R.; Robba, A.A.

    1982-01-01

    The Advanced Nuclear Technology Group of the Los Alamos National Laboratory is now using intelligent data-acquisition and analysis instrumentation for determining the multiplication of nuclear material. Earlier instrumentation, such as the large NIM-crate systems, depended on house power and required additional computation to determine multiplication or to estimate error. The portable, battery-powered multiplication measurement unit, with advanced computational power, acquires data, calculates multiplication, and completes error analysis automatically. Thus, the multiplication is determined easily and an available error estimate enables the user to judge the significance of results

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  13. 242Pu absolute neutron-capture cross section measurement

    Science.gov (United States)

    Buckner, M. Q.; Wu, C. Y.; Henderson, R. A.; Bucher, B.; Chyzh, A.; Bredeweg, T. A.; Baramsai, B.; Couture, A.; Jandel, M.; Mosby, S.; O'Donnell, J. M.; Ullmann, J. L.

    2017-09-01

    The absolute neutron-capture cross section of 242Pu was measured at the Los Alamos Neutron Science Center using the Detector for Advanced Neutron-Capture Experiments array along with a compact parallel-plate avalanche counter for fission-fragment detection. During target fabrication, a small amount of 239Pu was added to the active target so that the absolute scale of the 242Pu(n,γ) cross section could be set according to the known 239Pu(n,f) resonance at En,R = 7.83 eV. The relative scale of the 242Pu(n,γ) cross section covers four orders of magnitude for incident neutron energies from thermal to ≈ 40 keV. The cross section reported in ENDF/B-VII.1 for the 242Pu(n,γ) En,R = 2.68 eV resonance was found to be 2.4% lower than the new absolute 242Pu(n,γ) cross section.

  14. 242Pu absolute neutron-capture cross section measurement

    Directory of Open Access Journals (Sweden)

    Buckner M.Q.

    2017-01-01

    Full Text Available The absolute neutron-capture cross section of 242Pu was measured at the Los Alamos Neutron Science Center using the Detector for Advanced Neutron-Capture Experiments array along with a compact parallel-plate avalanche counter for fission-fragment detection. During target fabrication, a small amount of 239Pu was added to the active target so that the absolute scale of the 242Pu(n,γ cross section could be set according to the known 239Pu(n,f resonance at En,R = 7.83 eV. The relative scale of the 242Pu(n,γ cross section covers four orders of magnitude for incident neutron energies from thermal to ≈ 40 keV. The cross section reported in ENDF/B-VII.1 for the 242Pu(n,γ En,R = 2.68 eV resonance was found to be 2.4% lower than the new absolute 242Pu(n,γ cross section.

  15. A novel detector assembly for detecting thermal neutrons, fast neutrons and gamma rays

    Energy Technology Data Exchange (ETDEWEB)

    Cester, D., E-mail: davide.cester@gmail.com [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Lunardon, M.; Moretto, S. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); INFN Sezione di Padova, Via Marzolo 8, I-35131 Padova (Italy); Nebbia, G. [INFN Sezione di Padova, Via Marzolo 8, I-35131 Padova (Italy); Pino, F. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Sajo-Bohus, L. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Laboratorio de Fisica Nuclear, Universidad Simon Bolivar, Apartado 89000, 1080 A Caracas (Venezuela, Bolivarian Republic of); Stevanato, L.; Bonesso, I.; Turato, F. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy)

    2016-09-11

    A new composite detector has been developed by combining two different commercial scintillators. The device has the capability to detect gamma rays as well as thermal and fast neutrons; the signal discrimination between the three types is performed on-line by means of waveform digitizers and PSD algorithms. This work describes the assembled detector and its discrimination performance to be employed in the applied field.

  16. Safeguards and Physics Measurements: Neutron Dosimetry

    International Nuclear Information System (INIS)

    Vanhavere, F.

    2000-01-01

    The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations as well as to investigate the charcteristics of bubble detectors in order to be able to use them as direct-readiong neutron dosemeters

  17. A Canadian perspective on engineering strain measurements by neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Holden, T.M., E-mail: holdent@magma.ca [Northern Stress Technologies, Deep River, ON (Canada)

    2010-10-15

    The origins of the method, which relies on the high penetration of thermal neutrons through most industrial materials, in the late 1970s and 1980s and its development up to the present time are described. The basic steps from the measurement of lattice spacing to the calculation of stress are outlined. The length scales on which strains and stresses are manifest have a bearing on the interpretation of strain in terms of stress. A number of examples of macroscopic strain or stress fields are described where these scale considerations, as well as gradients of solute concentration that affect the interpretation of strain, matter. The strains on the length scale of the grains provide useful information about the elasticity and plasticity of materials, particularly in materials of lower symmetry than cubic. Finally, a number of future directions are suggested, which are enabled by the new intense neutron sources and improved equipment built in the last few years. (author)

  18. Measurement for skyshine of neutron generated by the K-600 neutron generator

    International Nuclear Information System (INIS)

    Zhen Huazhi; Li Guisheng; Wu Jingmin; Li Jianping

    1988-01-01

    The attenuation low of neutron scattering in atmosphere that generated by K-600 neutron generator at IMP was measured in order to evaluate the effect of the neutron generator to surroundings. The attenuation lenth λ = 396m was obtained and this result is in aggreement with the measured data at some laboratories abroad

  19. A militarily fielded thermal neutron activation sensor for landmine detection

    Energy Technology Data Exchange (ETDEWEB)

    Clifford, E.T.H. [Bubble Technology Industries, Chalk River (Canada); McFee, J.E. [Defence R and D Canada-Suffield, Medicine Hat (Canada)], E-mail: john.mcfee@drdc-rddc.gc.ca; Ing, H.; Andrews, H.R.; Tennant, D.; Harper, E. [Bubble Technology Industries, Chalk River (Canada); Faust, A.A. [Defence R and D Canada-Suffield, Medicine Hat (Canada)

    2007-08-21

    The Canadian Department of National Defence has developed a teleoperated, vehicle-mounted, multi-sensor system to detect anti-tank landmines on roads and tracks in peacekeeping operations. A key part of the system is a thermal neutron activation (TNA) sensor which is placed above a suspect location to within a 30 cm radius and confirms the presence of explosives via detection of the 10.835 MeV gamma ray associated with thermal neutron capture on {sup 14}N. The TNA uses a 100{mu}g{sup 252}Cf neutron source surrounded by four 7.62cmx7.62cm NaI(Tl) detectors. The system, consisting of the TNA sensor head, including source, detectors and shielding, the high-rate, fast pulse processing electronics and the data processing methodology are described. Results of experiments to characterize detection performance are also described. The experiments have shown that anti-tank mines buried 10 cm or less can be detected in roughly a minute or less, but deeper mines and mines significantly displaced horizontally take considerably longer time. Mines as deep as 30 cm can be detected for long count times (1000 s). Four TNA detectors are now in service with the Canadian Forces as part of the four multi-sensor systems, making it the first militarily fielded TNA sensor and the first militarily fielded confirmation sensor for landmines. The ability to function well in adverse climatic conditions has been demonstrated, both in trials and operations.

  20. Neutron measuring instruments for radiation protection

    International Nuclear Information System (INIS)

    Heinzelmann, M.; Schneider, W.; Hoefert, M.; Kuehn, H.; Jahr, R.; Wagner, S.; Piesch, E.

    1979-09-01

    The present report deals with selected topics from the field of neutron dosimetry for radiation protection connected with the work of the subcommittee 6802 in the Standards Committee on Radiology (NAR) of the German Standards Institute (DIN). It is a sort of material collection. The topics are: 1. Measurement of the absorbed-energy dose by a) ionization chambers in fields of mixed radiation and b) recoil-proton proportional counting tubes. 2. Measurement of the equivalent dose, neutron monitors, combination methods by a) rem-meters, b) recoil-proton counting tubes, c) recombination method, tissue-equivalent proportional counters, activation methods for high energies in fields of mixed radiation, d) personnel dosimetry by means of ionization chambers and counting tubes, e) dosimetry by means of activation methods, nuclear track films, nonphotographic nuclear track detectors and solid-state dosimeters. (orig./HP) [de

  1. Current advances in neutron diffraction stress measurement

    International Nuclear Information System (INIS)

    Suzuki, Hiroshi; Tomota, Yo

    2007-01-01

    Neutron diffraction has been employed for stress evaluation of various mechanical components. The hkl lattice plane spacings in a diffraction gauge volume are measured, then elastic strains are calculated and finally stresses are determined by using the Hooke's law. Since the real engineering mechanical parts are so complicated that more sophisticated method must be progressed to obtain stress distribution in the inside of a sample. Current advances on this issue are reviewed. (author)

  2. Measurement of absolute neutron flux in LWSCR based on the nuclear track method

    International Nuclear Information System (INIS)

    Sadeghzadeh, J.; Nassiri Mofakham, N.; Khajehmiri, Z.

    2012-01-01

    Highlights: ► Up to now the spectral parameters of thermal neutrons are measured with activation foils that are not always reliable in low flux systems. ► We applied a solid state nuclear track detector to measure the absolute neutron flux in the light water sub-critical reactor (LWSCR). ► Experiments concerning fission track detecting were performed and were investigated using the Monte Carlo code MCNP. ► The neutron fluxes obtained in experiment are in fairly good agreement with the results obtained by MCNP. - Abstract: In the present paper, a solid state nuclear track detector is applied to measure the absolute neutron flux in the light water sub-critical reactor (LWSCR) in Nuclear Science and Technology Research Institute (NSTRI). Up to now, the spectral parameters of thermal neutrons have been measured with activation foils that are not always reliable in low flux systems. The method investigated here is the irradiation method. Experiments concerning fission track detecting were performed. The experiment including neutron flux calculation method has also been investigated using the Monte Carlo code MCNP. The analysis shows that the values of neutron flux obtained by experiment are in fairly good agreement with the results obtained by MCNP. Thus, this method may be able to predict the absolute value of neutron flux at LWSCR and other similar reactors.

  3. Analysis of phosphate rock samples for vanadium using accelerator-based thermal neutrons

    International Nuclear Information System (INIS)

    Khiari, F.Z.; Aksoy, A.; Sahin, A.

    2004-01-01

    Extensive sedimentary phosphate deposits exist in the Sirhan-Turayf basin in northwestern Saudi Arabia containing significant amounts of uranium, thorium, vanadium and rare earth elements. The determination of the concentration and pattern of distribution of some of these elements is essential for economic aspects. Analysis of vanadium in selected phosphate rock samples from the basin using accelerator-based thermal neutrons activation analysis (TNAA) is reported. Samples were irradiated in a thermal neutron flux of 2.5 x 10 6 n x cm -2 x s -1 . The induced activity was measured with a HP-GMX detector coupled to a PC-based data acquisition and analysis system. The facility was calibrated using certified standards of vanadium. The minimum detection limit of vanadium was about 1 mg. Three independent measurements on each sample yielded comparable results indicating the reliability of the technique. The vanadium concentrations in the samples vary from 23 to 457 ppm. (author)

  4. Development of a dual phantom technique for measuring the fast neutron component of dose in boron neutron capture therapy.

    Science.gov (United States)

    Sakurai, Yoshinori; Tanaka, Hiroki; Kondo, Natsuko; Kinashi, Yuko; Suzuki, Minoru; Masunaga, Shinichiro; Ono, Koji; Maruhashi, Akira

    2015-11-01

    Research and development of various accelerator-based irradiation systems for boron neutron capture therapy (BNCT) is underway throughout the world. Many of these systems are nearing or have started clinical trials. Before the start of treatment with BNCT, the relative biological effectiveness (RBE) for the fast neutrons (over 10 keV) incident to the irradiation field must be estimated. Measurements of RBE are typically performed by biological experiments with a phantom. Although the dose deposition due to secondary gamma rays is dominant, the relative contributions of thermal neutrons (below 0.5 eV) and fast neutrons are virtually equivalent under typical irradiation conditions in a water and/or acrylic phantom. Uniform contributions to the dose deposited from thermal and fast neutrons are based in part on relatively inaccurate dose information for fast neutrons. This study sought to improve the accuracy in the dose estimation for fast neutrons by using two phantoms made of different materials in which the dose components can be separated according to differences in the interaction cross sections. The development of a "dual phantom technique" for measuring the fast neutron component of dose is reported. One phantom was filled with pure water. The other phantom was filled with a water solution of lithium hydroxide (LiOH) capitalizing on the absorbing characteristics of lithium-6 (Li-6) for thermal neutrons. Monte Carlo simulations were used to determine the ideal mixing ratio of Li-6 in LiOH solution. Changes in the depth dose distributions for each respective dose component along the central beam axis were used to assess the LiOH concentration at the 0, 0.001, 0.01, 0.1, 1, and 10 wt. % levels. Simulations were also performed with the phantom filled with 10 wt. % 6LiOH solution for 95%-enriched Li-6. A phantom was constructed containing 10 wt. % 6LiOH solution based on the simulation results. Experimental characterization of the depth dose distributions of the

  5. 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

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

    OpenAIRE

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

    1985-01-01

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

  7. Measuring the neutron lifetime using magnetically trapped neutrons

    Energy Technology Data Exchange (ETDEWEB)

    O' Shaughnessy, C.M.; Golub, R.; Schelhammer, K.W.; Swank, C.M.; Seo, P.-N. [North Carolina State University, 2401 Stinson Drive, Raleigh, NC (United States); Huffman, P.R., E-mail: Paul_Huffman@ncsu.ed [North Carolina State University, 2401 Stinson Drive, Raleigh, NC (United States); Dzhosyuk, S.N.; Mattoni, C.E.H.; Yang, L.; Doyle, J.M. [Harvard University, 17 Oxford Street, Cambridge, MA (United States); Coakley, K.J.; Thompson, A.K.; Mumm, H.P. [National Institute of Standards and Technology, 100 Bureau Drive, Gaithersburg, MD (United States); Lamoreaux, S.K.; McKinsey, D.N. [Yale University, 217 Prospect Street, New Haven, CT (United States); Yang, G. [University of Maryland, College Park, MD (United States)

    2009-12-11

    The neutron beta-decay lifetime plays an important role both in understanding weak interactions within the framework of the Standard Model and in theoretical predictions of the primordial abundance of {sup 4}He in Big Bang Nucleosynthesis. In previous work, we successfully demonstrated the trapping of ultracold neutrons in a conservative potential magnetic trap. A major upgrade of the apparatus is nearing completion at the National Institute of Standards and Technology Center for Neutron Research (NCNR). In our approach, a beam of 0.89 nm neutrons is incident on a superfluid {sup 4}He target within the minimum field region of an Ioffe-type magnetic trap. A fraction of the neutrons is downscattered in the helium to energies <200neV, and those in the appropriate spin state become trapped. The inverse process is suppressed by the low phonon density of helium at temperatures less than 200 mK, allowing the neutron to travel undisturbed. When the neutron decays the energetic electron ionizes the helium, producing scintillation light that is detected using photomultiplier tubes. Statistical limitations of the previous apparatus will be alleviated by significant increases in field strength and trap volume resulting in twenty times more trapped neutrons.

  8. Optimising polarised neutron scattering measurements--XYZ and polarimetry analysis

    International Nuclear Information System (INIS)

    Cussen, L.D.; Goossens, D.J.

    2002-01-01

    The analytic optimisation of neutron scattering measurements made using XYZ polarisation analysis and neutron polarimetry techniques is discussed. Expressions for the 'quality factor' and the optimum division of counting time for the XYZ technique are presented. For neutron polarimetry the optimisation is identified as analogous to that for measuring the flipping ratio and reference is made to the results already in the literature

  9. Optimising polarised neutron scattering measurements--XYZ and polarimetry analysis

    CERN Document Server

    Cussen, L D

    2002-01-01

    The analytic optimisation of neutron scattering measurements made using XYZ polarisation analysis and neutron polarimetry techniques is discussed. Expressions for the 'quality factor' and the optimum division of counting time for the XYZ technique are presented. For neutron polarimetry the optimisation is identified as analogous to that for measuring the flipping ratio and reference is made to the results already in the literature.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  13. Long Range Active Detection of HEU Based on Thermal Neutron Multiplication

    Energy Technology Data Exchange (ETDEWEB)

    Forman L.; Dioszegi I.; Salwen, C.; and Vanier, P.E.

    2010-05-24

    We report on the results of measurements of proton irradiation on a series of targets at Brookhaven National Laboratory’s (BNL) Alternate Gradient Synchrotron Facility (AGS), in collaboration with LANL and SNL. We examined the prompt radiation environment in the tunnel for the DTRA-sponsored series (E 972), which investigated the penetration of air and subsequent target interaction of 4 GeV proton pulses. Measurements were made by means of an organic scintillator with a 500 MHz bandwidth system. We found that irradiation of a depleted uranium (DU) target resulted in a large gamma-ray signal in the 100-500 µsec time region after the proton flash when the DU was surrounded by polyethylene, but little signal was generated if it was surrounded by boron-loaded polyethylene. Subsequent Monte Carlo (MCNPX) calculations indicated that the source of the signal was consistent with thermal neutron capture in DU. The MCNPX calculations also indicated that if one were to perform the same experiment with a highly enriched uranium (HEU) target there would be a distinctive fast neutron yield in this 100-500 µsec time region from thermal neutron-induced fission. The fast neutrons can be recorded by the same direct current system and differentiated from gamma ray pulses in organic scintillator by pulse shape discrimination.

  14. Monte Carlo simulation of a very high resolution thermal neutron detector composed of glass scintillator microfibers.

    Science.gov (United States)

    Song, Yushou; Conner, Joseph; Zhang, Xiaodong; Hayward, Jason P

    2016-02-01

    In order to develop a high spatial resolution (micron level) thermal neutron detector, a detector assembly composed of cerium doped lithium glass microfibers, each with a diameter of 1 μm, is proposed, where the neutron absorption location is reconstructed from the observed charged particle products that result from neutron absorption. To suppress the cross talk of the scintillation light, each scintillating fiber is surrounded by air-filled glass capillaries with the same diameter as the fiber. This pattern is repeated to form a bulk microfiber detector. On one end, the surface of the detector is painted with a thin optical reflector to increase the light collection efficiency at the other end. Then the scintillation light emitted by any neutron interaction is transmitted to one end, magnified, and recorded by an intensified CCD camera. A simulation based on the Geant4 toolkit was developed to model this detector. All the relevant physics processes including neutron interaction, scintillation, and optical boundary behaviors are simulated. This simulation was first validated through measurements of neutron response from lithium glass cylinders. With good expected light collection, an algorithm based upon the features inherent to alpha and triton particle tracks is proposed to reconstruct the neutron reaction position in the glass fiber array. Given a 1 μm fiber diameter and 0.1mm detector thickness, the neutron spatial resolution is expected to reach σ∼1 μm with a Gaussian fit in each lateral dimension. The detection efficiency was estimated to be 3.7% for a glass fiber assembly with thickness of 0.1mm. When the detector thickness increases from 0.1mm to 1mm, the position resolution is not expected to vary much, while the detection efficiency is expected to increase by about a factor of ten. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Carbon filter property detection with thermal neutron technique

    International Nuclear Information System (INIS)

    Deng Zhongbo; Han Jun; Li Wenjie

    2003-01-01

    The paper discussed the mechanism that the antigas property of the carbon filter will decrease because of its carbon bed absorbing water from the air while the carbon filter is being stored, and introduced the principle and method of detection the amount of water absorption with thermal neutron technique. Because some certain relation between the antigas property of the carbon filter and the amount of water absorption exists, the decrease degree of the carbon filter antigas property can be estimated through the amount of water absorption, offering a practicable facility technical pathway to quickly non-destructively detect the carbon filter antigas property

  16. Measurement of the 242Pu neutron capture cross section

    Science.gov (United States)

    Buckner, M. Q.; Wu, C. Y.; Henderson, R. A.; Bucher, B.; Bredeweg, T. A.; Baramsai, B.; Couture, A.; Jandel, M.; Mosby, S.; O'Donnell, J. M.; Ullmann, J. L.; Chyzh, A.; Dance Collaboration

    2015-10-01

    Precision (n,f) and (n, γ) cross sections are important for the network calculations of the radiochemical diagnostic chain for the U.S. DOE's Stockpile Stewardship Program. 242Pu(n, γ) cross section is relevant to the network calculations of Pu and Am. Additionally, new reactor concepts have catalyzed considerable interest in the measurement of improved cross sections for neutron-induced reactions on key actinides. To date, little or no experimental data has been reported on 242Pu(n, γ) for incident neutron energy below 50 keV. A new measurement of the 242Pu(n, γ) reaction was performed with the DANCE together with an improved PPAC for fission-fragment detection at LANSCE during FY14. The relative scale of the 242Pu(n, γ) cross section spans four orders of magnitude for incident neutron energies from thermal to ~ 30 keV. The absolute scale of the 242Pu(n, γ) cross section is set according to the measured 239Pu(n,f) resonance at 7.8 eV; the target was spiked with 239Pu for this measurement. The absolute 242Pu(n, γ) neutron capture cross section is ~ 30% higher than the cross section reported in ENDF for the 2.7 eV resonance. Latest results to be reported. Funded by U.S. DOE Contract No. DE-AC52-07NA27344 (LLNL) and DE-AC52-06NA25396 (LANL). U.S. DOE/NNSA Office of Defense Nuclear Nonproliferation Research and Development. Isotopes (ORNL).

  17. Development and Applications of Residual Stress Measurements Using Neutron Beams

    International Nuclear Information System (INIS)

    2014-01-01

    The deep penetration and selective absorption of neutrons make them a powerful tool for the non-destructive testing of large samples of material or large objects. Residual stress that is formed in a material during manufacturing, welding, utilization or repair can be measured by means of neutron diffraction. In fact, neutron diffraction is the only non-destructive testing method which can facilitate three dimensional mapping of residual stress in a bulk component. Stress measurement using neutron beams is a technique that enables this kind of high quality non-destructive investigation, and provides insight into the material strain and stress state deep within engineering components and structures under various conditions representative of those which might be experienced in service. Such studies are of importance to improve the quality of industrial components in production and to optimize design criteria in applications. Anisotropies in macroscopic properties such as thermal and electrical conductivities, for instance of fuel elements, and mechanical properties of materials depend on the textures developed during their preparation or thermal treatment. Such textures also can be studied using neutron diffraction techniques. There is currently substantial scientific and industrial demand for high quality non-destructive residual stress measurements, and the continuing competitive drive to optimize performance and minimize weight in many applications indicates that this demand will continue to grow. As such, the neutron diffraction technique is an increasingly important tool for mechanical and materials engineering in the search for improved manufacturing processes to reduce stress and distortion. Considering this trend, and in accordance with its purpose of promoting the peaceful use of nuclear applications, in 2006-2009 the IAEA organized a Coordinated Research Project on the Development and Application of the Techniques of Residual Stress Measurements in Materials

  18. Development and Applications of Residual Stress Measurements Using Neutron Beams

    OpenAIRE

    ABRIOLA S. A.; BALAGUROV A.; BASHIR J.; DAS A.; EDWARDS L.; GNAEUPEL-HEROLD T.; GOH B.; IONITA I.; MIKULA P.; OHMS Carsten; PELD N.; SCHNEIDER Rainer; SUTIARSO S.; TOROK G.; VENTER A.

    2012-01-01

    The deep penetration and selective absorption of neutrons make them a powerful tool in nondestructive testing of materials with large samples or objects. Residual stress formed in a material during manufacturing, welding, utilization or repairs can be measured by means of neutron diffraction. In fact, neutron diffraction is the only non-destructive testing method, which can facilitate 3-D mapping of residual stress in a bulk component. Stress measurement using neutron beams is a technique ...

  19. Spallation neutron spectra measured at Saturne

    International Nuclear Information System (INIS)

    Boyard, J.L.; Bouyer, P.; Brochard, F.; Duchazeaubeneix, J.C.; Durand, J.M.; Leray, S.; Milleret, G.; Plouin, F.; Uematsu, M.; Whittal, D.M.; Martinez, E.; Beau, M.; Boue, F.; Crespin, S.; Drake, D.; Frehaut, J.; Lochard, J.P.; Patin, Y.; Petibon, E.; Legrain, R.; Terrien, Y.

    1995-01-01

    Good knowledge of spallation reactions is necessary to design accelerator-based transmutation systems. An extensive program has begun at Saturne to measure energy and angular distributions of neutrons produced by incident protons or deuterons of up to 2 GeV on several thin targets. Our measurements will extend the available data to higher energies than the present limit of 800 MeV enabling improvements to the codes which are sometimes in poor agreement with the data. (Authors). 7 refs., 7 figs

  20. Thermal analysis of titanium drive-in target for D-D neutron generation.

    Science.gov (United States)

    Jung, N S; Kim, I J; Kim, S J; Choi, H D

    2010-01-01

    Thermal analysis was performed for a titanium drive-in target of a D-D neutron generator. Computational fluid dynamics code CFX-5 was used in this study. To define the heat flux term for the thermal analysis, beam current profile was measured. Temperature of the target was calculated at some of the operating conditions. The cooling performance of the target was evaluated by means of the comparison of the calculated maximum target temperature and the critical temperature of titanium. Copyright 2009 Elsevier Ltd. All rights reserved.

  1. Ascorbic acid reduced mutagenicity at the HPRT locus in CHO cells against thermal neutron radiation

    International Nuclear Information System (INIS)

    Kinashi, Yuko; Sakurai, Yoshinori; Masunaga, Shinichiro; Suzuki, Minoru; Nagata, Kenji; Ono, Koji

    2004-01-01

    We investigated the biological effects of the long-lived radicals induced following neutron irradiation. It has been reported that radiation-induced long-lived radicals were scavenged by post-irradiation treatment of ascorbic acid (Koyama, 1998). We studied the effects of ascorbic acid acting as a long-lived radical scavenger on cell killing and mutagenicity in Chinese hamster ovary cells against thermal neutrons produced at the Kyoto University Research reactor. Ascorbic acid was added to cells 30 min after neutron irradiation and removed 150 min after irradiation. The biological end point of cell survival was measured by colony formation assay. The mutagenicity was measured by the mutant frequency in the HPRT locus. The post-irradiation treatment of ascorbic acid did not alter the cell killing effect of neutron radiation. However, the mutagenicity was decreased, especially when the cells were irradiated with boron. Our results suggested that ascorbic acid scavenged long-lived radicals effectively and caused apparent protective effects against mutagenicity of boron neutron capture therapy

  2. The diffusional pulsed cooling of the thermal neutron flux in small two-region systems. Monte Carlo Simulation

    International Nuclear Information System (INIS)

    Wiacek, U.

    2006-06-01

    The thermal neutron transport in small unhomogeneous system and namely in two- layers where the first one -outer moderator is of hydride type (polyethylene or plexiglas) and the second one - inner is made with other materials is investigated. The diffusional cooling of neutrons has been calculated by means of monte Carlo simulations using MCPN code. Because of un consistency of calculated and measured data the MCPN code library has been modified for polyethylene and plexiglas

  3. Measurement of neutron yield by 62 MeV proton beam on a thick beryllium target

    Energy Technology Data Exchange (ETDEWEB)

    Osipenko, M., E-mail: osipenko@ge.infn.it [INFN, sezione di Genova, 16146 Genova (Italy); Ripani, M. [INFN, sezione di Genova, 16146 Genova (Italy); Alba, R. [INFN, Laboratori Nazionali del Sud, 95123 Catania (Italy); Ricco, G. [INFN, sezione di Genova, 16146 Genova (Italy); Schillaci, M. [INFN, Laboratori Nazionali del Sud, 95123 Catania (Italy); Barbagallo, M. [INFN, sezione di Bari, 70126 Bari (Italy); Boccaccio, P. [INFN, Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Celentano, A. [Dipartimento di Fisica dell' Università di Genova, 16146 Genova (Italy); Colonna, N. [INFN, sezione di Bari, 70126 Bari (Italy); Cosentino, L.; Del Zoppo, A.; Di Pietro, A. [INFN, Laboratori Nazionali del Sud, 95123 Catania (Italy); Esposito, J. [INFN, Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Figuera, P.; Finocchiaro, P. [INFN, Laboratori Nazionali del Sud, 95123 Catania (Italy); Kostyukov, A. [Moscow State University, Moscow 119992 (Russian Federation); Maiolino, C.; Santonocito, D.; Scuderi, V. [INFN, Laboratori Nazionali del Sud, 95123 Catania (Italy); Viberti, C.M. [Dipartimento di Fisica dell' Università di Genova, 16146 Genova (Italy)

    2013-09-21

    The design of a low-power prototype of neutron amplifier recently proposed within the INFN-E project indicated the need for more accurate data on the neutron yield produced by a proton beam with energy of about 70 MeV impinging on a thick beryllium target. Such measurement was performed at the LNS superconducting cyclotron, covering a wide angular range from 0° to 150° and a complete neutron energy interval from thermal to beam energy. Neutrons with energy above 0.5 MeV were measured by liquid scintillators exploiting their time of flight to determine the kinetic energy. For lower energy neutrons, down to thermal energy, a {sup 3}He detector was used. The obtained data are in good agreement with previous measurements at 0° using 66 MeV proton beam, covering neutron energies >10MeV, as well as with measurements at few selected angles using protons of 46, 55 and 113 MeV energy. The present results extend the neutron yield data in the 60–70 MeV beam energy range. A comparison of measured yields to MCNP, FLUKA and Geant4 Monte Carlo simulations was performed.

  4. Measurement of neutron yield by 62 MeV proton beam on a thick beryllium target

    Science.gov (United States)

    Osipenko, M.; Ripani, M.; Alba, R.; Ricco, G.; Schillaci, M.; Barbagallo, M.; Boccaccio, P.; Celentano, A.; Colonna, N.; Cosentino, L.; Del Zoppo, A.; Di Pietro, A.; Esposito, J.; Figuera, P.; Finocchiaro, P.; Kostyukov, A.; Maiolino, C.; Santonocito, D.; Scuderi, V.; Viberti, C. M.

    2013-09-01

    The design of a low-power prototype of neutron amplifier recently proposed within the INFN-E project indicated the need for more accurate data on the neutron yield produced by a proton beam with energy of about 70 MeV impinging on a thick beryllium target. Such measurement was performed at the LNS superconducting cyclotron, covering a wide angular range from 0° to 150° and a complete neutron energy interval from thermal to beam energy. Neutrons with energy above 0.5 MeV were measured by liquid scintillators exploiting their time of flight to determine the kinetic energy. For lower energy neutrons, down to thermal energy, a 3He detector was used. The obtained data are in good agreement with previous measurements at 0° using 66 MeV proton beam, covering neutron energies >10 MeV, as well as with measurements at few selected angles using protons of 46, 55 and 113 MeV energy. The present results extend the neutron yield data in the 60-70 MeV beam energy range. A comparison of measured yields to MCNP, FLUKA and Geant4 Monte Carlo simulations was performed.

  5. Measurement of neutron yield by 62 MeV proton beam on a thick beryllium target

    International Nuclear Information System (INIS)

    Osipenko, M.; Ripani, M.; Alba, R.; Ricco, G.; Schillaci, M.; Barbagallo, M.; Boccaccio, P.; Celentano, A.; Colonna, N.; Cosentino, L.; Del Zoppo, A.; Di Pietro, A.; Esposito, J.; Figuera, P.; Finocchiaro, P.; Kostyukov, A.; Maiolino, C.; Santonocito, D.; Scuderi, V.; Viberti, C.M.

    2013-01-01

    The design of a low-power prototype of neutron amplifier recently proposed within the INFN-E project indicated the need for more accurate data on the neutron yield produced by a proton beam with energy of about 70 MeV impinging on a thick beryllium target. Such measurement was performed at the LNS superconducting cyclotron, covering a wide angular range from 0° to 150° and a complete neutron energy interval from thermal to beam energy. Neutrons with energy above 0.5 MeV were measured by liquid scintillators exploiting their time of flight to determine the kinetic energy. For lower energy neutrons, down to thermal energy, a 3 He detector was used. The obtained data are in good agreement with previous measurements at 0° using 66 MeV proton beam, covering neutron energies >10MeV, as well as with measurements at few selected angles using protons of 46, 55 and 113 MeV energy. The present results extend the neutron yield data in the 60–70 MeV beam energy range. A comparison of measured yields to MCNP, FLUKA and Geant4 Monte Carlo simulations was performed

  6. The neutronic method for measuring soil moisture

    International Nuclear Information System (INIS)

    Couchat, Ph.

    1967-01-01

    The three group diffusion theory being chosen as the most adequate method for determining the response of the neutron soil moisture probe, a mathematical model is worked out using a numerical calculation programme with Fortran IV coding. This model is fitted to the experimental conditions by determining the effect of different parameters of measuring device: channel, fast neutron source, detector, as also the soil behaviour under neutron irradiation: absorbers, chemical binding of elements. The adequacy of the model is tested by fitting a line through the image points corresponding to the couples of experimental and theoretical values, for seven media having different chemical composition: sand, alumina, line stone, dolomite, kaolin, sandy loam, calcareous clay. The model chosen gives a good expression of the dry density influence and allows α, β, γ and δ constants to be calculated for a definite soil according to the following relation which gives the count rate of the soil moisture probe: N = (α ρ s +β) H v +γ ρ s + δ. (author) [fr

  7. Investigation on coupling characteristics of neutronics/thermal-hydraulics of PWR NPP core

    International Nuclear Information System (INIS)

    Zheng Yong; Peng Minjun; Xia Genglei; Liu Xinkai

    2014-01-01

    In this paper, an integrated neutronics/thermal-hydraulic model for the reactor of Qinshan Phase n NPP project was developed, using the RELAP5-HD as core coupled computational code. Based on the coupled model, the steady state calculation and the rod drop transient simulation were performed. The results show that the values obtained from RELAP5-HD calculation agree well with the available measured data, and the calculated accident curves can predict all major parameters trends of the transient with good accuracy. Both steady state and transient calculation results are in accordance with the theoretical analysis from the feedback aspect of coupled reactor neutronics/thermal-hydraulics, this demonstrates that a successful coupled model of Qinshan Phase n NPP core has been developed, and the established model provides a good foundation for further simulation analysis of the nuclear power plant system. (authors)

  8. Double helix boron-10 powder thermal neutron detector

    Science.gov (United States)

    Wang, Zhehui; Morris, Christopher L.; Bacon, Jeffrey D.

    2015-06-02

    A double-helix Boron-10 powder detector having intrinsic thermal neutron detection efficiency comparable to 36'' long, 2-in diameter, 2-bar Helium-3 detectors, and which can be used to replace such detectors for use in portal monitoring, is described. An embodiment of the detector includes a metallic plate coated with Boron-10 powder for generating alpha and Lithium-7 particles responsive to neutrons impinging thereon supported by insulators affixed to at least two opposing edges; a grounded first wire wound in a helical manner around two opposing insulators; and a second wire having a smaller diameter than that of the first wire, wound in a helical manner around the same insulators and spaced apart from the first wire, the second wire being positively biased. A gas, disposed within a gas-tight container enclosing the plate, insulators and wires, and capable of stopping alpha and Lithium-7 particles and generating electrons produces a signal on the second wire which is detected and subsequently related to the number of neutrons impinging on the plate.

  9. Measured neutron beam line shielding effectiveness of several iron/polyethylene configurations

    International Nuclear Information System (INIS)

    Legate, G.L.; Howe, M.L.; Mundis, R.L.

    1988-01-01

    Neutron and gamma-ray leakage measurements were taken at various stages of shield construction of neutron flight path 5 (the Lash-up flight path) at LANSCE, to compare the relative effectiveness of several configurations. Dose equivalent rates were determined for three categories: ''low-energy neutrons'', below 20 MeV; ''high- energy neutrons'', above 20 MeV; and gamma rays, as measured by hand-held survey instruments. The low energy neutrons were measured by activation of an indium foil in a paraffin-filled cadmium canister, sized to be generally insensitive above 20 MeV. High-energy neutrons were measured by (n,2n) production of Carbon 11 in a plastic scintillator with a 20-MeV threshold. Thermal neutrons were not measured at the shield-leakage test points. Room-scattered neutrons were observed by Albatross IV detector readings, which were taken beside the shield as a measure of variation of room background as the shield configuration changed. 1 fig., 1 tab

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  11. A slow neutron polarimeter for the measurement of parity-odd neutron rotary power

    Energy Technology Data Exchange (ETDEWEB)

    Snow, W. M.; Anderson, E.; Bass, T. D.; Dawkins, J. M.; Fry, J.; Haddock, C.; Horton, J. C.; Luo, D.; Micherdzinska, A. M.; Walbridge, S. B. [Indiana University and Center for the Exploration of Energy and Matter, 2401 Milo B. Sampson Lane, Bloomington, Indiana 47408 (United States); Barrón-Palos, L.; Maldonado-Velázquez, M. [Instituto de Física, Universidad Nacional Autónoma de México, Ciudad de México, D.F. 04510, México (Mexico); Bass, C. D. [LeMoyne College, 1419 Salt Springs Road, Syracuse, New York 13214 (United States); Crawford, B. E. [Gettysburg College, 300 North Washington Street, Gettysburg, Pennsylvania 17325 (United States); Crawford, C. [University of Kentucky, 177 Chem.-Phys. Building, 505 Rose Street, Lexington, Kentucky 40506-0055 (United States); Esposito, D. [University of Dayton, 300 College Park, Dayton, Ohio 45469 (United States); Gardiner, H. [Louisiana State University, Baton Rouge, Louisiana 70803 (United States); Gan, K. [The George Washington University, 2121 I Street N.W., Washington, District of Columbia 20052 (United States); Heckel, B. R.; Swanson, H. E., E-mail: swanson@npl.washington.edu [University of Washington/Center for Experimental Nuclear Physics and Astrophysics, Box 354290, Seattle, Washington 98195 (United States); and others

    2015-05-15

    We present the design, description, calibration procedure, and an analysis of systematic effects for an apparatus designed to measure the rotation of the plane of polarization of a transversely polarized slow neutron beam as it passes through unpolarized matter. This device is the neutron optical equivalent of a crossed polarizer/analyzer pair familiar from light optics. This apparatus has been used to search for parity violation in the interaction of polarized slow neutrons in matter. Given the brightness of existing slow neutron sources, this apparatus is capable of measuring a neutron rotary power of dϕ/dz = 1 × 10{sup −7} rad/m.

  12. Surface physics with cold and thermal neutron reflectometry

    International Nuclear Information System (INIS)

    Steyerl, A.

    1991-11-01

    Three aspects of the research project ''Surface physics with cold and ultracold neutron reflectometry'' were stressed during the present first year: (1) Setup of the reflectometer facility at the research reactor of the Rhode Island Nuclear Science Center. The installation provides a narrow ''pencil beam'' analyzed by time of flight using a chopper system. Following beam characterization and a test measurement of the total cross section of copper single crystal first reflectivity measurements are currently performed using a supermirror. (2) Design stud for the ultracold neutron imaging system, with involvement of the relevant industry. Bids are available for several components indicating that it will be very difficult to build the entire system unless further funds become available. (3) Analysis of features of neutron reflection from surfaces with special emphasis on the effect of surface roughness both on the specular beam and the diffusely reflected and refracted intensity. Previous theoretical studies were supplemented by further numerical calculations of diffuse scattering distributions using different models. Application of ultracold and cold neutron reflectometry to the study of liquid-vapor phase transition were discussed. The theoretical work also includes the development of tentative ideas for novel fundamental physics experiments

  13. Measuring planetary neutron albedo fluxes by remote gamma-ray sensing

    Science.gov (United States)

    Haines, E. L.; Metzger, A. E.

    1984-01-01

    In order to measure the planetary neutron albedo fluxes, a neutron-absorbing shield which emits gamma rays of characteristic energy and serves as a neutron detector, is added to a gamma-ray spectrometer (GRS). The gamma rays representing the neutron flux are observed against interference consisting of cosmic gamma rays, planetary continuum and line emission, and gamma rays arising from the interaction of cosmic rays with the GRS and the spacecraft. The uncertainty and minimum detection limits in neutron albedo fluxes are calculated for two missions, a lunar orbiter and a comet nucleus rendezvous. A GRS on a lunar orbiter at 100 km altitude detects a thermal neutron albedo flux as low as 0.002/sq cm/s and an expected flux of about 0.6/sq cm/s is measured with an uncertainty of 0.001/sq cm/s, for a 100 h observation period. For the comet nucleus, again in a 100 h observing period, a thermal neutron albedo flux is detected at a level of 0.006/sq cm/s and an expected flux of about 0.4/sq cm/s is measured with an uncertainty of 0.004/sq cm/s. The expanded geological capabilities made possible by this technique include improvements in H sensitivity, spatial resolution, and measurement depth; and an improved model of induced gamma-ray emission.

  14. Thermal-neutron capture gamma-rays. Volume 1

    International Nuclear Information System (INIS)

    Tuli, J.K.

    1997-05-01

    The energy and photon intensity of gamma rays as seen in thermal-neutron capture are presented in ascending order of gamma energy. All those gamma-rays with intensity of ≥ 2% of the strongest transition are included. The two strongest transitions seen for the target nuclide are indicated in each case. Where the target nuclide mass number is indicated as nat the natural target was used. The gamma energies given are in keV. The gamma intensities given are relative to 100 for the strongest transition. All data for A > 44 are taken from Evaluated Nuclear Structure Data File (4/97), a computer file of evaluated nuclear structure data maintained by the National Nuclear Data Center, Brookhaven National Laboratory, on behalf of the Nuclear Structure and Decay and Decay Data network, coordinated by the International Atomic Energy Agency, Vienna. These data are published in Nuclear Data Sheets, Academic Press, San Diego, CA. The data for A ≤ 44 is taken from ''Prompt Gamma Rays from Thermal-Neutron Capture,'' M.A. Lone, R.A. Leavitt, D.A. Harrison, Atomic Data and Nuclear Data Tables 26, 511 (1981)

  15. 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.

  16. Development of a scattering probability method for accurate vapor fraction measurements by neutron radiography

    CERN Document Server

    Joo, H

    1999-01-01

    Recent test results indicated drawbacks associated with the simple exponential attenuation method (SEAM) as currently applied to neutron radiography measurements to determine vapor fractions in a hydrogenous two-phase flow in a metallic conduit. The scattering component of the neutron beam intensity exiting the flow system is not adequately accounted for by SEAM, and this leads to inaccurate results. To properly account for the scattering effect, a neutron scattering probability method (SPM) is developed. The method applies a neutron-hydrogen scattering kernel to scattered thermal neutrons that leave the incident beam in narrow conduits but eventually show up elsewhere in the measurements. The SPM has been tested with known vapor (void) distributions within an acrylic disk and a water/vapor channel. The vapor (void) fractions deduced by SPM are in good agreement with the known exact values. Details of the scattering correction method and the test results are discussed.

  17. Neutron fluxes and spectra of the neutron guide tubes in the upgraded JRR-3

    International Nuclear Information System (INIS)

    Kawabata, Y.; Suzuki, M.; Tsuruno, A.; Onishi, N.

    1992-01-01

    A cold neutron source with two thermal and three cold neutron guide tubes has been installed in the upgraded JRR-3. Neutron fluxes, time-of-flight spectra and the gain of the cold neutron source were measured at the end of the neutron guide tubes. The neutron beam distributions were found by neutron radiography films. (orig.)

  18. An alternative method for the measurement of neutron flux

    Indian Academy of Sciences (India)

    Abstract. A simple and easy method for measuring the neutron flux is presented. This paper deals with the experimental verification of neutron dose rate–flux relationship for a non-dissipative medium. Though the neutron flux cannot be obtained from the dose rate in a dissipative medium, experimental result shows that for ...

  19. An alternative method for the measurement of neutron flux

    Indian Academy of Sciences (India)

    A simple and easy method for measuring the neutron flux is presented. This paper deals with the experimental verification of neutron dose rate–flux relationship for a non-dissipative medium. Though the neutron flux cannot be obtained from the dose rate in a dissipative medium, experimental result shows that for ...

  20. EDM: Neutron electric dipole moment measurement

    Directory of Open Access Journals (Sweden)

    Peter Fierlinger

    2016-02-01

    Full Text Available An electric dipole moment (EDM of the neutron would be a clear sign of new physics beyond the standard model of particle physics. The search for this phenomenon is considered one of the most important experiments in fundamental physics and could provide key information on the excess of matter versus antimatter in the universe. With high measurement precision, this experiment aims to ultimately achieve a sensitivity of 10-28 ecm, a 100-fold improvement in the sensitivity compared to the state-of-the-art. The EDM instrument is operated by an international collaboration based at the Technische Universität München.

  1. Neutron Spectrum Measurements from Irradiations at NCERC

    Energy Technology Data Exchange (ETDEWEB)

    Jackman, Kevin Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Michelle A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bredeweg, Todd Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchens, Gregory Joe [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-04-15

    Several irradiations have been conducted on assemblies (COMET/ZEUS and Flattop) at the National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site (NNSS). Configurations of the assemblies and irradiated materials changed between experiments. Different metallic foils were analyzed using the radioactivation method by gamma-ray spectrometry to understand/characterize the neutron spectra. Results of MCNP calculations are shown. It was concluded that MCNP simulated spectra agree with experimental measurements, with the caveats that some data are limited by statistics at low-energies and some activation foils have low activities.

  2. Control of thermal-neutron pulses under conditions of heterogeneous poisoning of the moderator

    International Nuclear Information System (INIS)

    Alekseev, N.I.

    1987-01-01

    Monochromatic thermal neutrons are often used to study effects of the nuclear, molecular, and crystal structures of materials. The most widespread method for obtaining monochromatic neutrons is the time-of-flight method which requires an impulsive neutron beam. In this paper thermal neutron sources consisting of moderators and targets are considered from the standpoints of the relationship between neutron energy and pulse duration in experiments based on the time-of-flight method and with heterogeneous poisoning of the moderator in order to arrive at a pulse control scenario for optimizing the monochromaticity and energy resolution of the beam

  3. Neutron Energy Measurements in Radiological Emergency Response Applications

    Energy Technology Data Exchange (ETDEWEB)

    Sanjoy Mukhopadhyay, Paul Guss, Michael Hornish, Scott Wilde, Tom Stampahar, Michael Reed

    2009-04-30

    We present significant results in recent advances in the determination of neutron energy. Neutron energy measurements are a small but very significant part of radiological emergency response applications. Mission critical information can be obtained by analyzing the neutron energy given off from radioactive materials. In the case of searching for special nuclear materials, neutron energy information from an unknown source can be of paramount importance.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  5. Analysis of prompt fission neutrons in 235U(nth,f) and fission fragment distributions for the thermal neutron induced fission of 234U

    Science.gov (United States)

    Al-Adili, A.; Tarrío, D.; Hambsch, F.-J.; Göök, A.; Jansson, K.; Solders, A.; Rakopoulos, V.; Gustafsson, C.; Lantz, M.; Mattera, A.; Oberstedt, S.; Prokofiev, A. V.; Vidali, M.; Österlund, M.; Pomp, S.

    2016-06-01

    This paper presents the ongoing analysis of two fission experiments. Both projects are part of the collaboration between the nuclear reactions group at Uppsala and the JRC-IRMM. The first experiment deals with the prompt fission neutron multiplicity in the thermal neutron induced fission of 235U(n,f). The second, on the fission fragment properties in the thermal fission of 234U(n,f). The prompt fission neutron multiplicity has been measured at the JRC-IRMM using two liquid scintillators in coincidence with an ionization chamber. The first experimental campaign focused on 235U(nth,f) whereas a second experimental campaign is foreseen later for the same reaction at 5.5 MeV. The goal is to investigate how the so-called sawtooth shape changes as a function of fragment mass and excitation energy. Some harsh experimental conditions were experienced due to the large radiation background. The solution to this will be discussed along with preliminary results. In addition, the analysis of thermal neutron induced fission of 234U(n,f) will be discussed. Currently analysis of data is ongoing, originally taken at the ILL reactor. The experiment is of particular interest since no measurement exist of the mass and energy distributions for this system at thermal energies. One main problem encountered during analysis was the huge background of 235U(nth,f). Despite the negligible isotopic traces in the sample, the cross section difference is enormous. Solution to this parasitic background will be highlighted.

  6. Development of a portable thermal neutron detector based on a boron rich heterodiode

    Science.gov (United States)

    Tomov, R.; Venn, R.; Owens, A.; Peacock, A.

    2008-10-01

    Results are presented on the development of a portable detector suitable for detection of individual thermal neutrons. The device is based on direct absorption of neutrons in an absorber film containing 10B. The resultant charge arising from the capture products is detected by a p-n junction partly formed from this absorber and internal to the device. When a small bias voltage is applied (typically a few volts) a current pulse is observed due to the movement of this charge in the electric field of the p-n junction. For each detected neutron the charge pulse height, rise time and time of detection are recorded. Device performance, in terms of efficiency and spectral response, is explored as a function of neutron absorber thickness, geometry and overall diode electrical characteristics and validated against neutron source measurements at the UK National Physical Laboratory (NPL). The diodes have a natural background suppression capability through traditional pulse height and pulse rise time discrimination. The manufacturing process permits fabrication of arrays of diodes, with typical areas of ~15 mm2, thus increasing the collecting area and the signal to noise ratio, albeit with increased readout complexity. The associated multi-channel readout electronics is standard, however, and commonly used in existing X-ray sensors. Simple portable sensors based on these heterodiodes are expected to have applications in the detection of nuclear materials in a variety of security related situations.

  7. Correlated γ rays following capture of thermal neutrons on 113Cd

    Science.gov (United States)

    Rusev, G.; Jandel, M.; Arnold, C. W.; Bredeweg, T. A.; Couture, A.; Mosby, S. M.; Ullmann, J. L.; Krtička, M.

    2013-10-01

    Natural cadmium is often used as the shielding against thermal neutrons and component in detectors sensitive to neutrons, because of the large cross section of 113Cd for capture of neutrons with energies below 1 eV. Investigation of the neutron-capture γ rays from the 113Cd (n , γ) reaction is of importance for these applications. We report the intensity distributions of these cascade γ-ray transitions. The neutron-capture experiment on 113Cd has been carried out at LANL's LANSCE using the 4 π BaF2 DANCE array. The measured two-dimensional spectrum of counts vs. γ-ray energy vs. γ-ray multiplicity from the strongest resonance in the 113Cd (n , γ) reaction at 0.178 eV has been compared with predictions from the statistical model using the code DICEBOX. Work supported by the NNSA Office of Nonproliferation and Verification Research and Development performed under the Department of Energy contract DE-AC52-06NA25396.

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

    CERN Document Server

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

    1999-01-01

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

  9. Miniature neutron sources: Thermal neutron sources and their users in the academic field

    International Nuclear Information System (INIS)

    Egelstaff, P.A.

    1992-01-01

    The three levels of thermal neutron sources are introduced - University laboratory sources infrastructure sources and world-class sources - and the needs for each kind and their inter-dependence will be emphasized. A description of the possibilities for University sources based on α-Be reactions or spontaneous fission emission is given, and current experience with them is described. A new generation of infrastructure sources is needed to continue the regional programs based on small reactors. Some possibilities for accelerator sources that could meet this need are considered

  10. 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).

  11. Capability of NIPAM polymer gel in recording dose from the interaction of 10B and thermal neutron in BNCT

    International Nuclear Information System (INIS)

    Khajeali, Azim; Reza Farajollahi, Ali; Kasesaz, Yaser; Khodadadi, Roghayeh; Khalili, Assef; Naseri, Alireza

    2015-01-01

    The capability of N-isopropylacrylamide (NIPAM) polymer gel to record the dose resulting from boron neutron capture reaction in BNCT was determined. In this regard, three compositions of the gel with different concentrations of 10 B were prepared and exposed to gamma radiation and thermal neutrons. Unlike irradiation with gamma rays, the boron-loaded gels irradiated by neutron exhibited sensitivity enhancement compared with the gels without 10 B. It was also found that the neutron sensitivity of the gel increased by the increase of concentration of 10 B. It can be concluded that NIPAM gel might be suitable for the measurement of the absorbed dose enhancement due to 10 B and thermal neutron reaction in BNCT. - Highlights: • Three compositions of NIPAM gel with different concentration of 10 B have been exposed by gamma and thermal neutron. • The vials containing NIPAM gel have been irradiated by an automatic system capable of providing for dose uniformity. • Suitability of NIPAM polymer gel in measuring radiation doses in BNCT has been investigated.

  12. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  13. Neutron activation system for spectral measurements of pulsed ion diode neutron production

    International Nuclear Information System (INIS)

    Hanson, D.L.; Kruse, L.W.

    1980-02-01

    A neutron energy spectrometer has been developed to study intense ion beam-target interactions in the harsh radiation environment of a relativistic electron beam source. The main component is a neutron threshold activation system employing two multiplexed high efficiency Ge(Li) detectors, an annihilation gamma coincidence system, and a pneumatic sample transport. Additional constraints on the neutron spectrum are provided by total neutron yield and time-of-flight measurements. A practical lower limit on the total neutron yield into 4π required for a spectral measurement with this system is approx. 10 10 n where the neutron yield is predominantly below 4 MeV and approx. 10 8 n when a significant fraction of the yield is above 4 MeV. Applications of this system to pulsed ion diode neutron production experiments on Hermes II are described

  14. Development of neutron measurement techniques in reactor diagnostics and determination of water content and water flow

    International Nuclear Information System (INIS)

    Avdic, Senada

    2000-09-01

    The present thesis deals with three comparatively different topics in neutron physics research. These topics are as follows: construction and experimental investigation of a new detector, capable of measuring the neutron current, and investigation of the possibility to use it for the localisation of a neutron source in a simple experimental arrangement; execution of neutron transmission measurements based on a stationary neutron generator, and the study of their suitability for determining the volume porosity of geological samples; study of the possibility for improving the accuracy of water flow measurements based on the pulsed neutron activation technique. The first subject of this thesis concerns the measurement of the neutron current by a newly constructed detector. The motivation for this work stems from a recent suggestion that the performance of core monitoring methods could be enhanced if, in addition to the scalar neutron flux, also the neutron current was measured. To this end, a current detector was based on a scintillator mounted on a fibre and a Cd layer on one side of the detector. The measurements of the 2-D neutron current were performed in an experimental system by using this detector. The efficiency of the detector in reactor diagnostics was illustrated by demonstrating that the position of a neutron source can be determined by measuring the scalar neutron flux and the neutron current in one spatial point. The results of measurement and calculation show both the suitability of the detector construction for the measurement of the neutron current vector and the use of the current in diagnostics and monitoring. The second subject of this thesis concerns fast neutron transmission measurements, based on a stationary neutron generator, for determining the volume porosity of a sample in a model experiment. Such a technique could be used in field measurements with obvious advantages in comparison with thermal neutron transmission techniques, which can

  15. Measurement of 235U fission spectrum-averaged cross sections and neutron spectrum adjusted with the activation data

    International Nuclear Information System (INIS)

    Kobayashi, Katsuhei; Kobayashi, Tooru

    1992-01-01

    The 235 U fission spectrum-averaged cross sections for 13 threshold reactions were measured with the fission plate (27 cm in diameter and 1.1 cm thick) at the heavy water thermal neutron facility of the Kyoto University Reactor. The Monte Carlo code MCNP was applied to check the deviation from the 235 U fission neutron spectrum due to the room-scattered neutrons, and it was found that the resultant spectrum was close to that of 235 U fission neutrons. Supplementally, the relations to derive the absorbed dose rates with the fission plate were also given using the calculated neutron spectra and the neutron Kerma factors. Finally, the present values of the fission spectrum-averaged cross sections were employed to adjust the 235 U fission neutron spectrum with the NEUPAC code. The adjusted spectrum showed a good agreement with the Watt-type fission neutron spectrum. (author)

  16. Measurement of spectra and neutron fluxes on artificial earth satellites from the Cosmos series

    Science.gov (United States)

    Dudkin, V. Y.; Kovalev, Y. Y.; Novikova, M. R.; Potapov, Y. V.; Skvortsov, S. S.; Smirennyy, L. N.

    1975-01-01

    In 1966-1967 measurements were carried out at the altitudes of 200 to 400 km to determine the spectra and fluxes of fast neutrons inside the hermetically sealed artificial earth satellites of the Cosmos series. The detectors used were nuclear emulsions of the B9 and BR types and an emulsion of the P9 type, filled with Li and P. Spectra and fluxes of neutrons in the range of energies from thermal energies to 10 MeV are presented. Neutron doses are also estimated.

  17. Fission neutron spectra measurements at LANSCE - status and plans

    Energy Technology Data Exchange (ETDEWEB)

    Haight, Robert C [Los Alamos National Laboratory; Noda, Shusaku [Los Alamos National Laboratory; Nelson, Ronald O [Los Alamos National Laboratory; O' Donnell, John M [Los Alamos National Laboratory; Devlin, Matt [Los Alamos National Laboratory; Chatillon, Audrey [CEA-FRANCE; Granier, Thierry [CEA-FRANCE; Taieb, Julien [CEA-FRANCE; Laurent, Benoit [CEA-FRANCE; Belier, Gilbert [CEA-FRANCE; Becker, John A [LLNL; Wu, Ching - Yen [LLNL

    2009-01-01

    A program to measure fission neutron spectra from neutron-induced fission of actinides is underway at the Los Alamos Neutron Science Center (LANSCE) in a collaboration among the CEA laboratory at Bruyeres-le-Chatel, Lawrence Livermore National Laboratory and Los Alamos National Laboratory. The spallation source of fast neutrons at LANSCE is used to provide incident neutron energies from less than 1 MeV to 100 MeV or higher. The fission events take place in a gas-ionization fission chamber, and the time of flight from the neutron source to that chamber gives the energy of the incident neutron. Outgoing neutrons are detected by an array of organic liquid scintillator neutron detectors, and their energies are deduced from the time of flight from the fission chamber to the neutron detector. Measurements have been made of the fission neutrons from fission of {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu. The range of outgoing energies measured so far is from 1 MeV to approximately 8 MeV. These partial spectra and average fission neutron energies are compared with evaluated data and with models of fission neutron emission. Results to date will be presented and a discussion of uncertainties will be given in this presentation. Future plans are to make significant improvements in the fission chambers, neutron detectors, signal processing, data acquisition and the experimental environment to provide high fidelity data including mea urements of fission neutrons below 1 MeV and improvements in the data above 8 MeV.

  18. Coupled neutronics - thermal-hydraulics programs for SCWRS

    Energy Technology Data Exchange (ETDEWEB)

    Reiss, T. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Muegyetem rkp. 9., 1111 Budapest (Hungary)

    2010-07-01

    The Supercritical Water Cooled Reactor (SCWR) was chosen as one of the Generation IV reactors by GIF. At the moment, a number of concepts - thermal as well as fast ones - exist. The reference parameters for a thermal SCWR have been taken from the European High Performance Light Water Reactor (HPLWR). Since the pressure is higher than the critical pressure (22.1 MPa) there is no change in the phase of the water in the core. On the other hand, due to the significant changes in the physical properties of water at supercritical pressure, the system is susceptible to local temperature, density and power oscillations. This inclination is increased by the pseudo-critical transformation of the water used as coolant. Thus, for modelling a system of this type coupled neutronics - thermal-hydraulics programs are required. Such a program system has been developed with the following main features: great modularity which allows for easy modifications, thus several SCWR concepts can be studied; detailed assembly calculations (with MCNP) and full-core analysis (with SCALE) are supported; the differential equations of xenon poisoning are implemented to study xenon oscillations. The program system was used to examine the assembly of the HPLWR, to design the assembly and the core of the Simplified Supercritical Water Cooled Reactor (SSCWR) and to model xenon oscillations in SCWRs. (authors)

  19. Measurements of thermal accommodation coefficients.

    Energy Technology Data Exchange (ETDEWEB)

    Rader, Daniel John; Castaneda, Jaime N.; Torczynski, John Robert; Grasser, Thomas W.; Trott, Wayne Merle

    2005-10-01

    A previously-developed experimental facility has been used to determine gas-surface thermal accommodation coefficients from the pressure dependence of the heat flux between parallel plates of similar material but different surface finish. Heat flux between the plates is inferred from measurements of temperature drop between the plate surface and an adjacent temperature-controlled water bath. Thermal accommodation measurements were determined from the pressure dependence of the heat flux for a fixed plate separation. Measurements of argon and nitrogen in contact with standard machined (lathed) or polished 304 stainless steel plates are indistinguishable within experimental uncertainty. Thus, the accommodation coefficient of 304 stainless steel with nitrogen and argon is estimated to be 0.80 {+-} 0.02 and 0.87 {+-} 0.02, respectively, independent of the surface roughness within the range likely to be encountered in engineering practice. Measurements of the accommodation of helium showed a slight variation with 304 stainless steel surface roughness: 0.36 {+-} 0.02 for a standard machine finish and 0.40 {+-} 0.02 for a polished finish. Planned tests with carbon-nanotube-coated plates will be performed when 304 stainless-steel blanks have been successfully coated.

  20. Calibration of NE213 detector in neutron measurement

    International Nuclear Information System (INIS)

    Akkurt, I

    2004-01-01

    Organic scintillator is one of the widely used materials in neutron measurement as they have good timing properties and a high hydrogen content. Calibration of the detector system is an important part of the experimental study for interpretation of the results. As the neutron uncharged, the pulse from the detector is not directly used to determine neutron kinetic energy but the detection threshold for recoil charged particles (p,d,□ etc) has to be known in order to calculate the neutron detection efficiency. In this work calibration procedure of a NE213 detector array used in neutron measurements at MaxLab (Lund,Sweden ), is described.This includes both pulse height and neutron flight time which is important in neutron energy determinations

  1. RPL-SC dosimetric system for measuring gamma and neutron irradiation in case of emergency

    International Nuclear Information System (INIS)

    Khristova, M. G.

    1993-01-01

    A RPL-SC dosimetric system is designed based on radiophotoluminescence (RPL) and on the effect of fast neutron bombardment of silicon semiconductor (SC) diodes. The experimental prototype consists of a computerized automatic measurement system and an individual dosimetric cassette accommodating RPL and SC detectors. The equipment includes: a device for measurement of the direct voltage of Si diodes and the RPL light emitted by RPL detectors; a compartment with dosimetric cassettes to be measured; a manipulator with three positions executing automatic measurement of cassettes; a computer and a printer. The system operates in both manual and automatic modes. In the manual mode each step of the manipulator is set up by the operator who changes the ranges after they have been filled to capacity and registers the results. In the automatic mode the whole process of maintaining the supply and control voltage, of manipulator's operation, measuring, data recording and data processing are controlled by a specially designed computer programme. Main technical parameters: 1) Measurement range of absorbed dose: gamma rays - 10 -3 to 10 2 Gy; thermal neutrons - 10 -3 to 10 2 Gy; fast neutrons - 10 to 30 Gy. 2) Energy range: gamma rays - 0.04 to 1.25 MeV; thermal neutrons - 0.024 eV; fast neutrons - 0.3 to 14 MeV. 3) Relative measurement error - ±15% 4) Recurrent measurement of one and the same dose. 5) Measurement time of 1 detector - 15 sec. (author)

  2. Benchmarking a first-principles thermal neutron scattering law for water ice with a diffusion experiment

    Directory of Open Access Journals (Sweden)

    Holmes Jesse

    2017-01-01

    Full Text Available The neutron scattering properties of water ice are of interest to the nuclear criticality safety community for the transport and storage of nuclear materials in cold environments. The common hexagonal phase ice Ih has locally ordered, but globally disordered, H2O molecular orientations. A 96-molecule supercell is modeled using the VASP ab initio density functional theory code and PHONON lattice dynamics code to calculate the phonon vibrational spectra of H and O in ice Ih. These spectra are supplied to the LEAPR module of the NJOY2012 nuclear data processing code to generate thermal neutron scattering laws for H and O in ice Ih in the incoherent approximation. The predicted vibrational spectra are optimized to be representative of the globally averaged ice Ih structure by comparing theoretically calculated and experimentally measured total cross sections and inelastic neutron scattering spectra. The resulting scattering kernel is then supplied to the MC21 Monte Carlo transport code to calculate time eigenvalues for the fundamental mode decay in ice cylinders at various temperatures. Results are compared to experimental flux decay measurements for a pulsed-neutron die-away diffusion benchmark.

  3. Retrospective Reactor Dosimetry for Neutron Fluence, Helium, and Boron Measurements

    International Nuclear Information System (INIS)

    Greenwood, Lawrence R.; Oliver, Brian M.

    2003-01-01

    Neutron fluences can be measured and radiation damage parameters determined by analyzing the neutron reaction products in very small samples removed from components of an operating power research reactor. This process, known as retrospective reactor dosimetry, provides precise neutron exposure parameters for establishing or validating calculations of neutron fluences, helium generation, and radiation damage to reactor materials. Correlation of the neutron fluence and helium data helps to establish and validate models of radiation damage and helium production that are needed to address important issues such as irradiation assisted stress corrosion cracking, void swelling, and weld repair of cracks. Results are presented for samples recently obtained from several operating reactors

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

    Science.gov (United States)

    Vermeeren, L.; Wéber, M.

    2003-06-01

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

  5. Thermal-neutron multiple scattering: critical double scattering

    International Nuclear Information System (INIS)

    Holm, W.A.

    1976-01-01

    A quantum mechanical formulation for multiple scattering of thermal-neutrons from macroscopic targets is presented and applied to single and double scattering. Critical nuclear scattering from liquids and critical magnetic scattering from ferromagnets are treated in detail in the quasielastic approximation for target systems slightly above their critical points. Numerical estimates are made of the double scattering contribution to the critical magnetic cross section using relevant parameters from actual experiments performed on various ferromagnets. The effect is to alter the usual Lorentzian line shape dependence on neutron wave vector transfer. Comparison with corresponding deviations in line shape resulting from the use of Fisher's modified form of the Ornstein-Zernike spin correlations within the framework of single scattering theory leads to values for the critical exponent eta of the modified correlations which reproduce the effect of double scattering. In addition, it is shown that by restricting the range of applicability of the multiple scattering theory from the outset to critical scattering, Glauber's high energy approximation can be used to provide a much simpler and more powerful description of multiple scattering effects. When sufficiently close to the critical point, it provides a closed form expression for the differential cross section which includes all orders of scattering and has the same form as the single scattering cross section with a modified exponent for the wave vector transfer

  6. Effective thermal neutron absorption cross section for heterogeneous mixture

    International Nuclear Information System (INIS)

    Gabanska, B.; Igielski, A.; Krynicka-Drozdowicz, E.; Woznicka, U.

    1989-01-01

    The first estimations (basing on Umiastowski's theory) of the influence of the sample heterogeneity of the effective thermal neutron absorption cross section were compared with the results obtained for the homogeneous mixture which components and concentration were the same as those of the heterogeneous sample. An experiment was prepared to determine how good this estimate is. Three artificial heterogeneous cylindrical samples (2R = H = 9 cm) were manufactured from pure silver cylinders embedded in plexiglass, keeping the Ag content and varying the size of cylinders (2R = H = 1.0 cm, 0.6 cm and 0.4 cm). Calculations performed show that the experimental effect of the sample heterogeneity can be significant. 5 figs., 5 tabs, 11 refs. (author)

  7. Use of pulsed neutron-neutron logging, thermal neutron-neutron logging, and gamma logging methods in classification for sand-clay sediments of Lower Cretaceous in Prikumsk oil-and-gas region according to filtration-capacitance characteristics

    International Nuclear Information System (INIS)

    Maksimenko, A.N.; Basin, Ya.N.; Novgorodov, V.A.

    1974-01-01

    To isolate reservoirs, the formation and deformation penetration zone parameters are used. They are estimated according to the false oil saturation factor and the time of the penetration zone deformation which are determined from the complex exploration of cased wells using the pulse neutron logging, thermal neutron-neutron logging and gamma logging techniques

  8. Numerical stability analysis of coupled neutronics and thermal-hydraulics schemes and new neutronic feedback-reactions model

    International Nuclear Information System (INIS)

    Guertin, Chantal

    1995-01-01

    This thesis is part of the validation process of using coupled 3D neutronics and thermal-hydraulics codes for studying accidental situations with boiling. First part is dedicated to a numerical stability analysis of neutronics and thermal-hydraulics coupled schemes. Both explicit and semi-implicit coupling schemes were applied to solve the set of equations describing the linearized neutronics and thermal-hydraulics of point reactor. Point reactor modelling was preferred to obtain analytical expressions of eigenvalues of the discretized Systems. Stability criteria, based on eigenvalues, was calculated as well as neutronic and thermalhydraulic responses of the System following insertion of a reactivity step. Results show no severe restriction of time domain, stability wise. Actual transient calculations using coupled neutronics and thermal-hydraulics codes, like COCCINELLE and THYC developed at Electricite de France, do not show stability problems. Second part introduces surface spline as a new neutronic feedback model. The cross influences of feedback parameters is now taken into account. Moderator temperature and density were modeled. This method, simple and accurate, allows an homogeneous description of cross-sections overall operating reactor situations including accidents with boiling. (author) [fr

  9. Magnetic field devices for neutron spin transport and manipulation in precise neutron spin rotation measurements

    Energy Technology Data Exchange (ETDEWEB)

    Maldonado-Velázquez, M. [Posgrado en Ciencias Físicas, Universidad Nacional Autónoma de México, 04510 (Mexico); Barrón-Palos, L., E-mail: libertad@fisica.unam.mx [Instituto de Física, Universidad Nacional Autónoma de México, Apartado Postal 20-364, 01000 (Mexico); Crawford, C. [University of Kentucky, Lexington, KY 40506 (United States); Snow, W.M. [Indiana University, Bloomington, IN 47405 (United States)

    2017-05-11

    The neutron spin is a critical degree of freedom for many precision measurements using low-energy neutrons. Fundamental symmetries and interactions can be studied using polarized neutrons. Parity-violation (PV) in the hadronic weak interaction and the search for exotic forces that depend on the relative spin and velocity, are two questions of fundamental physics that can be studied via the neutron spin rotations that arise from the interaction of polarized cold neutrons and unpolarized matter. The Neutron Spin Rotation (NSR) collaboration developed a neutron polarimeter, capable of determining neutron spin rotations of the order of 10{sup −7} rad per meter of traversed material. This paper describes two key components of the NSR apparatus, responsible for the transport and manipulation of the spin of the neutrons before and after the target region, which is surrounded by magnetic shielding and where residual magnetic fields need to be below 100 μG. These magnetic field devices, called input and output coils, provide the magnetic field for adiabatic transport of the neutron spin in the regions outside the magnetic shielding while producing a sharp nonadiabatic transition of the neutron spin when entering/exiting the low-magnetic-field region. In addition, the coils are self contained, forcing the return magnetic flux into a compact region of space to minimize fringe fields outside. The design of the input and output coils is based on the magnetic scalar potential method.

  10. Improved thermal neutron activation sensor for detection of bulk explosives

    Science.gov (United States)

    McFee, John E.; Faust, Anthony A.; Andrews, H. Robert; Clifford, Edward T. H.; Mosquera, Cristian M.

    2012-06-01

    Defence R&D Canada - Suffield and Bubble Technology Industries have been developing thermal neutron activation (TNA) sensors for detection of buried bulk explosives since 1994. First generation sensors, employing an isotopic source and NaI(Tl) gamma ray detectors, were deployed by Canadian Forces in 2002 as confirmation sensors on the ILDS teleoperated, vehicle-mounted, multi-sensor anti-tank landmine detection systems. The first generation TNA could detect anti-tank mines buried 10 cm or less in no more than a minute, but deeper mines and those significantly displaced horizontally required considerably longer times. Mines as deep as 30 cm could be detected with long counting times (1000 s). The second generation TNA detector is being developed with a number of improvements aimed at increasing sensitivity and facilitating ease of operation. Among these are an electronic neutron generator to increase sensitivity for deeper and horizontally displaced explosives; LaBr3(Ce) scintillators, to improve time response and energy resolution; improved thermal and electronic stability; improved sensor head geometry to minimize spatial response nonuniformity; and more robust data processing. This improved sensitivity can translate to either decreased counting times, decreased minimum detectable explosive quantities, increased maximum sensor-to-target displacement, or a trade off among all three. Experiments to characterize the performance of the latest generation TNA in detecting buried landmines and IEDs hidden in culverts were conducted during 2011. This paper describes the second generation system. The experimental setup and methodology are detailed and preliminary comparisons between the performance of first and second generation systems are presented.

  11. Application of neutron backscatter techniques to level measurement problems

    International Nuclear Information System (INIS)

    Leonardi-Cattolica, A.M.; McMillan, D.H.; Telfer, A.; Griffin, L.H.; Hunt, R.H.

    1982-01-01

    We have designed and built portable level detectors and fixed level monitors based on neutron scattering and detection principles. The main components of these devices, which we call neutron backscatter gauges, are a neutron emitting radioisotope, a neutron detector, and a ratemeter. The gauge is a good detector for hydrogen but is much less sensitive to most other materials. This allows level measurements of hydrogen bearing materials, such as hydrocarbons, to be made through the walls of metal vessels. Measurements can be made conveniently through steel walls which are a few inches thick. We have used neutron backscatter gauges in a wide variety of level measurement applications encountered in the petrochemical industry. In a number of cases, the neutron techniques have proven to be superior to conventional level measurement methods, including gamma ray methods

  12. The fluence research of filter material for fast neutron fluence measurement

    International Nuclear Information System (INIS)

    Tang Xiding

    2010-01-01

    When the fast neutron fluence is measured by radioactivation techniques in the nuclear reactor the fast neutron is also filtered a little by the thermal neutron filter material, and if the filter material thickness increase the filtered fast neutron increases therewith. For fast neutron fluenc measurement, there are only cadmium, boron and gadolinium three elements filtering fluence can be calculated ordinarily. In order to calculate the filtered fast neutron fluence of the all elements in the filter material, the many total cross sections of nuclides had checked out from nuclear cross section data library, converted them into the same energy group structure, then element's total cross section, compound's total cross section and multilayer filters' total cross section had calculated from these total cross sections with same energy group structure, a new cross section data library can be obtained lastly through merging these cross sections into the old cross section data library used for neutron fluence measurement. The calculation analysis indicates that the results of the unit 2 surveillance capsule U of DAYA Bay NPP and the unit 1 surveillance capsule A of the Second Nuclear Power Plant of Qinshan by considering the all elements subtracting iron are smaller about 1.5% and 2.6% respectively than the ones only to consider cadmium, boron. The old measured results accord with the new values under the measurement uncertainty, are reliable. The new results are more accuracy. (authors)

  13. Burn-up measurements coupling gamma spectrometry and neutron measurement

    International Nuclear Information System (INIS)

    Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.

    2006-01-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  14. Cosmic-ray neutron simulations and measurements in Taiwan.

    Science.gov (United States)

    Chen, Wei-Lin; Jiang, Shiang-Huei; Sheu, Rong-Jiun

    2014-10-01

    This study used simulations of galactic cosmic ray in the atmosphere to investigate the neutron background environment in Taiwan, emphasising its altitude dependence and spectrum variation near interfaces. The calculated results were analysed and compared with two measurements. The first measurement was a mobile neutron survey from sea level up to 3275 m in altitude conducted using a car-mounted high-sensitivity neutron detector. The second was a previous measured result focusing on the changes in neutron spectra near air/ground and air/water interfaces. The attenuation length of cosmic-ray neutrons in the lower atmosphere was estimated to be 163 g cm(-2) in Taiwan. Cosmic-ray neutron spectra vary with altitude and especially near interfaces. The determined spectra near the air/ground and air/water interfaces agree well with measurements for neutrons below 10 MeV. However, the high-energy portion of spectra was observed to be much higher than our previous estimation. Because high-energy neutrons contribute substantially to a dose evaluation, revising the annual sea-level effective dose from cosmic-ray neutrons at ground level in Taiwan to 35 μSv, which corresponds to a neutron flux of 5.30 × 10(-3) n cm(-2) s(-1), was suggested. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  15. Procedure for measurement of anisotropy factor for neutron sources

    International Nuclear Information System (INIS)

    Creazolla, Prycylla Gomes

    2017-01-01

    Radioisotope neutron sources allow the production of reference fields for calibration of neutron detectors for radiation protection and analysis purposes. When the emission rate of these sources is isotropic, no correction is necessary. However, variations in source encapsulation and in the radioactive material concentration produce differences in its neutron emission rate, relative to the source axis, this effect is called anisotropy. In this study, is describe a procedure for measuring the anisotropy factor of neutron sources performed in the Laboratório de Metrologia de Neutrons (LN) using a Precision Long Counter (PLC) detector. A measurement procedure that takes into account the anisotropy factor of neutron sources contributes to solve some issues, particularly with respect to the high uncertainties associated with neutron dosimetry. Thus, a bibliographical review was carried out based on international standards and technical regulations specific to the area of neutron fields, and were later reproduced in practice by means of the procedure for measuring the anisotropy factor in neutron sources of the LN. The anisotropy factor is determined as a function of the angle of 90° in relation to the cylindrical axis of the source. This angle is more important due to its high use in measurements and also of its higher neutron emission rate if compared with other angles. (author)

  16. Beryllium phonon spectrum from cold neutron measurements

    International Nuclear Information System (INIS)

    Bulat, I.A.

    1979-01-01

    The inelastic coherent scattering of neutrons with the initial energy E 0 =4.65 MeV on the spectrometer according to the time of flight is studied in polycrystalline beryllium. The measurements are made for the scattering angles THETA=15, 30, 45, 60, 75 and 90 deg at 293 K. The phonon spectrum of beryllium, i-e. g(w) is reestablished from the experimental data. The data obtained are compared with the data of model calculations. It is pointed out that the phonon spectrum of beryllium has a bit excessive state density in the energy range from 10 to 30 MeV. It is caused by the insufficient statistical accuracy of the experiment at low energy transfer

  17. Thermal neutron imaging through XRQA2 GAFCHROMIC films coupled with a cadmium radiator

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, D. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); INAIL – DIT, Via di Fontana Candida n.1, 00040 Monteporzio Catone (Italy); Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Bortot, D. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Palomba, M. [ENEA Casaccia, Via Anguillarese, 301, S. Maria di Galeria, 00123 Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); Gentile, A. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Strigari, L. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Pressello, C. [Department of Medical Physics, Azienda Ospedaliera San Camillo Forlanini, Circonvallazione Gianicolense 87, 00152 Roma (Italy); Soriani, A. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Gómez-Ros, J.M. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

    2015-10-21

    A simple and inexpensive method to perform passive thermal neutron imaging on large areas was developed on the basis of XRQA2 GAFCHROMIC films, commonly employed for quality assurance in radiology. To enhance their thermal neutron response, the sensitive face of film was coupled with a 1 mm thick cadmium radiator, forming a sandwich. By exchanging the order of Cd filter and sensitive film with respect to the incident neutron beam direction, two different configurations (beam-Cd-film and beam-film-Cd) were identified. These configurations were tested at thermal neutrons fluence values in the range 10{sup 9}–10{sup 10} cm{sup −2}, using the ex-core radial thermal neutron column of the ENEA Casaccia – TRIGA reactor. The results are presented in this work.

  18. Correction for variable moderation and multiplication effects associated with thermal neutron coincidence counting

    International Nuclear Information System (INIS)

    Baron, N.

    1978-01-01

    A correction is described for multiplication and moderation when doing passive thermal neutron coincidence counting nondestructive assay measurements on powder samples of PuO 2 mixed arbitrarily with MgO, SiO 2 , and moderating material. The multiplication correction expression is shown to be approximately separable into the product of two independent terms; F/sub Pu/ which depends on the mass of 240 Pu, and F/sub αn/ which depends on properties of the matrix material. Necessary assumptions for separability are (1) isotopic abundances are constant, and (2) fission cross sections are independent of incident neutron energy: both of which are reasonable for the 8% 240 Pu powder samples considered here. Furthermore since all prompt fission neutrons are expected to have nearly the same energy distributions, variations among different samples can be due only to the moderating properties of the samples. Relative energy distributions are provided by a thermal neutron well counter having two concentric rings of 3 He proportional counters placed symmetrically about the well. Measured outer-to-inner ring ratios raised to an empirically determined power for coincidences, (N/sup I//N/sup O/)/sup Z/, and singles, (T/sup O//T/sup I/)/sup delta/, provide corrections for moderation and F/sub αn/ respectively, and F/sub Pu/ is approximated by M 240 /sup X//M 240 . The exponents are calibration constants determined by a least squares fitting procedure using standards' data. System calibration is greatly simplified using the separability principle. Once appropriate models are established for F/sub Pu/ and F/sub αn/, only a few standards are necessary to determine the calibration constants associated with these terms. Since F/sub Pu/ is expressed as a function of M 240 , correction for multiplication in a subsequent assay demands only a measurement of F/sub αn/

  19. Measurements of neutron spectrum from uranium converter

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  1. New measurement of neutron capture resonances of 209Bi

    CERN Document Server

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

    2006-01-01

    The neutron capture cross section of Bi209 has been measured at the CERN n TOF facility by employing the pulse-height-weighting technique. Improvements over previous measurements are mainly because of an optimized detection system, which led to a practically negligible neutron sensitivity. Additional experimental sources of systematic error, such as the electronic threshold in the detectors, summing of gamma-rays, internal electron conversion, and the isomeric state in bismuth, have been taken into account. Gamma-ray absorption effects inside the sample have been corrected by employing a nonpolynomial weighting function. Because Bi209 is the last stable isotope in the reaction path of the stellar s-process, the Maxwellian averaged capture cross section is important for the recycling of the reaction flow by alpha-decays. In the relevant stellar range of thermal energies between kT=5 and 8 keV our new capture rate is about 16% higher than the presently accepted value used for nucleosynthesis calculations. At th...

  2. Neutron beta decay measurements planned for the SNS

    Science.gov (United States)

    Pocanic, Dinko

    2009-10-01

    A cold neutron beam line, dedicated to fundamental neutron physics (FnPB), is presently being completed at the Oak Ridge, TN, Spallation Neutron Source. Among other experiments, the beamline will host a comprehensive set of precise studies of the neutron beta decay. Neutron beta decay is characterised by the decay rate (or its inverse, the neutron lifetime), and a set of decay parameters describing the kinematical and spin correlations among the participating particles. Within the standard model (SM), the neutron lifetime and three decay parameters (a, A, and B) are fixed by two parameters: the Vud element of the Cabibbo-Kobayashi-Maskawa mixing matrix, and λ=GA/GV, the ratio of axial vector and vector nucleon form factors. This overdetermined system provides a unique opportunity to explore possible departures from the simple SM, as well as the nature of such departures, e.g., left-right supersymmetric extensions, leptoquarks, non-(V-A) admixtures, etc., with broad implications in subatomic physics. The FnPB neutron beta decay program will include measurements of the neutron lifetime, continuing the present NIST experiment, a measurement of a, the electron-neutrino correlation, and b, the Fierz interference term, (the ``Nab'' experiment), along with measurements of A and B, the correlations between neutron spin and electron and neutrino momenta, respectively, (the ``abBA'' experiment). Current plans for these experiments will be discussed in detail.

  3. Large solid-angle polarisation analysis at thermal neutron wavelengths using a sup 3 He spin filter

    CERN Document Server

    Heil, W; Cywinski, R; Humblot, H; Ritter, C; Roberts, T W; Stewart, J R

    2002-01-01

    The strongly spin-dependent absorption of neutrons in nuclear spin-polarised sup 3 He opens up the possibility of polarising neutrons from reactors and spallation sources over the full kinematical range of cold, thermal and hot neutrons. In this paper we describe the first large solid-angle polarisation analysis measurement using a sup 3 He neutron spin filter at thermal neutron wavelengths (lambda=2.5 A). This experiment was performed on the two-axis diffractometer D1B at the Institut Laue-Langevin using a banana-shaped filter cell (530 cm sup 3 ) filled with sup 3 He gas with a polarisation of P=52% at a pressure of 2.7 bar. A comparison is made with a previous measurement on D7 using a cold neutron beam on the same sample, i.e. amorphous ErY sub 6 Ni sub 3. Using uniaxial polarisation analysis both the nuclear and magnetic cross-sections could be extracted over the range of scattering-vectors [0.5<=Q(A sup - sup 1)<=3.5]. The results are in qualitative and quantitative agreement with the D7-data, whe...

  4. Neutron measurements in search of cold fusion

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.E.; Goulding, C.A.; Johnson, M.W.; Butterfield, K.B.; Gottesfeld, S.; Baker, D.A.; Springer, T.E.; Garzon, F.H.; Bolton, R.D.; Leonard, E.M.; Chancellor, T. (Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States))

    1991-05-10

    We have conducted a search for neutron emission from cold fusion systems of the electrochemical type and, to a lesser extent, the high-pressure gas cell type. Using a high-efficiency well counter and an NE 213 scintillator, the experiments were conducted on the earth's surface and in a shielded cave approximately 50 ft underground. After approximately 6500 h of counting time, we have obtained no evidence for cold fusion processes leading to neutron production. However, we have observed all three types of neutron data that have been presented as evidence for cold fusion: large positive fluctuations in the neutron counting rate, weak peaks near 2.5 MeV in the neutron energy spectrum, and bursts of up to 140 neutrons in 500-{mu}s intervals. The data were obtained under circumstances that clearly show our results to be data encountered as a part of the naturally occurring neutron background, which is due primarily to cosmic rays. Thus, observing these types of data does not, of itself, provide evidence for the existence of cold fusion processes. Artifacts in the data that were due to counter misbehavior were also observed to lead to long-term neutron bursts'' whose time duration varied from several hours to several days. We conclude that any experiments which attempt to observed neutron emission must include strong steps to ensure that the experiments deal adequately with both cosmic-ray processes and counter misbehavior.

  5. Neutron measurements in search of cold fusion

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.E.; Goulding, C.A.; Johnson, M.W.; Butterfield, K.B.; Gottesfeld, S.; Baker, D.A.; Springer, T.E.; Garzon, F.H.; Bolton, R.D.; Leonard, E.M.; Chancellor, T.

    1990-01-01

    We have conducted a research for neutron emission from cold fusion systems of the electrochemical type and, to a lesser extent, the high-pressure gas cell type. Using a high-efficiency well counter and an NE 213 scintillator, the experiments were conducted on the earth's surface and in a shielded cave approximately 50 ft underground. After approximately 6500 h of counting time, we have obtained no evidence for cold fusion processes leading to neutron production. However, we have observed all three types of neutron data that have been presented as evidence for cold fusion: large positive fluctuations in the neutron counting rate, weak peaks near 2.5 MeV in the neutron energy spectrum, and bursts of up to 145 neutrons in 500-{mu}s intervals. The data were obtained under circumstances that clearly show our results to be data encountered as a part of naturally occurring neutron background, which is due primarily to cosmic rays. Thus, observing these types of data does not, of itself, provide evidence for the existence of cold fusion processes. Artifacts in the data that were due to counter misbehavior were also to lead to long-term neutron bursts'' whose time duration varied from several hours to several days. We conclude that any experiments which attempt to observe neutron emission must include strong steps to ensure that the experiments deal adequately with both cosmic-ray processes and counter misbehavior. 13 refs., 14 figs.

  6. Advanced thermal neutron area detector. The development and application of an imaging plate neutron detector

    International Nuclear Information System (INIS)

    Niimura, Nobuo

    1995-01-01

    This report reviews a newly developed imaging plate neutron detector (IP-ND), along with its actual application. First, imaging plate, which is an integrating two-dimensional radiation detector using photostimulated luminescence (PSL), is briefly mentioned. Then, IP-ND is described in terms of the following: design principle, trial manufacture of IP-ND, and performance (such as dynamic range, spatial resolution, neutron detection efficiency, and PSL according to kinds of neutron converters). The application of IP-ND is outlined under the following fields: (1) neutron radiography, (2) electric noiseless detector, (3) fast neutron detector, (4) neutron diffraction, (5) neutron scattering, and (6) neutron reflector. (N.K.)

  7. Neutron emission effects on fragment mass and kinetic energy distribution from fission of 239Pu induced by thermal neutrons

    International Nuclear Information System (INIS)

    Montoya, M.; Rojas, J.; Lobato, I.

    2010-01-01

    The average of fragment kinetic energy (E-bar sign*) and the multiplicity of prompt neutrons (ν(bar sign)) as a function of fragment mass (m*), as well as the fragment mass yield (Y(m*)) from thermal neutron-induced fission of 239 Pu have been measured by Tsuchiya et al.. In that work the mass and kinetic energy are calculated from the measured kinetic energy of one fragment and the difference of time of flight of the two complementary fragments. However they do not present their results about the standard deviation σ E *(m*). In this work we have made a numerical simulation of that experiment which reproduces its results, assuming an initial distribution of the primary fragment kinetic energy (E(A)) with a constant value of the standard deviation as function of fragment mass (σ E (A)). As a result of the simulation we obtain the dependence σ E *(m*) which presents an enhancement between m* = 92 and m* = 110, and a peak at m* = 121.

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

    International Nuclear Information System (INIS)

    Nakai, Hirokazu

    1990-01-01

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

  9. High pressure sample container for thermal neutron spectroscopy and diffraction on strongly scattering fluids

    International Nuclear Information System (INIS)

    Verkerk, P.; Pruisken, A.M.M.

    1979-01-01

    A description is presented of the construction and performance of a container for thermal neutron scattering on a fluid sample with about 1.5 cm -1 macroscopic cross section (neglecting absorption). The maximum pressure is about 900 bar. The container is made of 5052 aluminium capillary with inner diameter 0.75 mm and wall thickness 0.25 mm; it covers a neutron beam with a cross section of 9 X 2.5 cm 2 . The container has been successfully used in neutron diffraction and time-of-flight experiments on argon-36 at 120 K and several pressures up to 850 bar. It is shown that during these measurements the temperature gradient over the sample as well as the error in the absolute temperature were both less than 0.05 K. Subtraction of the Bragg peaks due to container scattering in diffraction experiments may be dfficult, but seems feasible because of the small amount of aluminium in the neutron beam. Correction for container scattering and multiple scattering in time-of-flight experiments may be difficult only in the case of coherently scattering samples and small scattering angles. (Auth.)

  10. Recent Progress Towards a Measurement of the Neutron Lifetime Using Magnetically Trapped Ultracold Neutrons

    Science.gov (United States)

    Schelhammer, K. W.; Huffer, C. R.; Huffman, P. R.; Marley, D. E.; Coakley, K. J.; Huber, Michael; Hughes, P. P.; Mumm, H. P.; Thompson, A. K.; Yue, A. T.; Abrams, N. C.

    2012-03-01

    Free neutron beta decay is a fundamental process in the Standard Model that can be used to test the weak interaction as well as provide information about primordial ^4He abundance. Recent precision measurements of the neutron lifetime have led to reduced confidence in the absolute value of this parameter; due presumably to unknown systematic effects. This work seeks to measure the neutron lifetime using a different technique that employs a superconducting magnetic trap to confine ultracold neutrons. Neutrons are loaded into the trap through the superthermal technique where 1 mEv neutrons down scatter from phonons in liquid helium losing the majority of their energy. Neutrons in the appropriate spin state are then confined by the static magnetic field. During the past year, over 400 run cycles of data were collected using the upgraded apparatus. Analysis of previous data sets was limited due to large numbers of background events relative to the neutron decay signal. An increased number of trapped neutrons as well as a analysis using pulse shape discrimination allows one to significantly increase the overall precision of the measurement. Details of this ongoing analysis will be presented with preliminary results.

  11. The chemical consequences of thermal neutron capture in alkali selenates

    International Nuclear Information System (INIS)

    Duplatre, G.; Vargas, J.I.

    1977-01-01

    The initial retention of the SeO 4 2- ion after thermal neutron capture has been studied in various matrices by chemical analysis. A comparison between the thermal behaviour of the chemically analyzed Sesup(IV) and the disappearance of the E.P.R. species SeO 3 - and SeO 4 3- showed that the retention fraction would include all species with oxidation state higher or equal to VI. The retentions observed in the different matrices show the existence of four families with respective retentions of: 2.6%[K 2 SeO 4 diluted in (NH 4 ) 2 SO 4 ], 9.2% [anhydrous and hydrated Li and Ca selenates; K 2 SeO 4 diluted in NaIO 3 ; Se + implanted in K 2 SeO 4 ; Triglycine selenate], 21.5% [K 2 SeO 4 diluted in KNO 3 , K 2 SO 4 , Na 2 WO 4 and Na 2 WO 4 .2H 2 O] and 32.0% [Na,K and Cs selenates]. Whereas chemical considerations may be invoked for the (NH 4 ) 2 SO 4 matrix, a mechanical model is proposed for the three other groups. (author)

  12. 10B enriched plastic scintillators for application in thermal neutron detection

    Science.gov (United States)

    Mahl, Adam; Yemam, Henok A.; Fernando, Roshan; Koubek, Joshua T.; Sellinger, Alan; Greife, Uwe

    2018-02-01

    We report here on the synthesis and characterization of a novel 10B enriched aromatic molecule that can be incorporated into common poly(vinyltoluene) (PVT) based plastic scintillators to achieve enhanced thermal neutron detection. Starting from relatively inexpensive 10B enriched boric acid, we have prepared 4,4,5,5-tetramethyl-2-phenyl-1,3,2-dioxaborolane (MBB) in three high yield steps. MBB is soluble and compatible with PVT based formulations and results in stable plastic scintillators. Chemical synthesis, solubility limit in PVT, and the physical properties of the dopant were explored. The relevant response properties of the resulting scintillators when exposed to neutron and gamma radiation, including light yield and pulse shape discrimination properties were measured and analyzed.

  13. Bubble Detector Neutron Measurements on JET High Performance Discharges

    Energy Technology Data Exchange (ETDEWEB)

    Gherendi, M.; Craciunescu, T.; Pantea, A.; Zoita, V. [Association EURATOM-MEdC, National Institute for Laser, Plasma and Radiation Physics, Bucharest (Romania); Conroy, S.; Gatu Johnson, M.; Hellesen, C. [Association EURATOM-VR, Uppsala University, Uppsala (Sweden); Edlington, T.; Kiptily, V.; Popovichev, S. [Association EURATOM-CCFE, Culham Science Centre, Abingdon (United Kingdom); Murari, A. [Association EURATOM-ENEA, RFX, Padova (Italy)

    2011-07-01

    A neutron diagnostics technique based on the bubble detectors has been successfully used for measurements during the JET experimental campaigns of 2008 and 2009. The main aim of these measurements was the determination of the neutron field characteristics in high performance discharges which employ high neutral beam powers ({approx}20 MW) and produce neutron yields in the range (3-5)*10{sup 16} neutrons per pulse. The neutron field parameters at a specific location above a narrow collimating channel in the ceiling of the JET Torus Hall have been measured simultaneously by two independent techniques (super-heated fluid detectors or 'bubble detectors' and time-of-flight). The bubble detector measurement location is situated at the end of a vertical collimated line of sight, behind the TOFOR time-of-flight spectrometer. The field-of-view of the neutron detectors can be varied by means of a pre-collimator. Spatial (radial and toroidal) distributions of the neutron fluence have been obtained using two-dimensional arrays containing up to 10 bubble detectors. The operation of the bubble detector array as a neutron pinhole camera having a radial resolution at the JET vacuum chamber mid-plane of about 55 mm was demonstrated in measurements using various openings of the pre-collimator. The comparison of the area integrated fluence determined by the bubble detector array with the TOFOR neutron flux has shown a good correlation factor of about 0.99. (authors)

  14. Procedures for measurement of anisotropy factor of neutron sources

    International Nuclear Information System (INIS)

    Creazolla, P.G.; Camargo, A.; Astuto, A.; Silva, F.; Pereira, W.W.

    2017-01-01

    Radioisotope sources of neutrons allow the production of reference fields for calibration of neutron measurement devices for radioprotection and analysis purposes. When the emission rate of these sources is isotropic, no correction is necessary. However, variations in the source capsule material and variations in the concentration of the emitting material may produce differences in its neutron emission rate relative to the source axis, this effect is called anisotropy. A proposed procedure for measuring the anisotropy factor of the sources belonging to the IRD/LNMRI/LN Neutron Metrology Laboratory using a Precision Long Counter (PLC) detector will be presented

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  16. Note: Coincidence measurements of 3He and neutrons from a compact D-D neutron generator.

    Science.gov (United States)

    Ji, Q; Lin, C-J; Tindall, C; Garcia-Sciveres, M; Schenkel, T; Ludewigt, B A

    2017-05-01

    Tagging of neutrons (2.45 MeV) with their associated 3 He particles from deuterium-deuterium (D-D) fusion reactions has been demonstrated in a compact neutron generator setup enabled by a high brightness, microwave-driven ion source with a high fraction of deuterons. Energy spectra with well separated peaks of the D-D fusion reaction products, 3 He, tritons, and protons, were measured with a silicon PIN diode. The neutrons were detected using a liquid scintillator detector with pulse shape discrimination. By correlating the 3 He detection events with the neutron detection in time, we demonstrated the tagging of emitted neutrons with 3 He particles detected with a Si PIN diode detector mounted inside the neutron generator vacuum vessel.

  17. Note: Coincidence measurements of 3He and neutrons from a compact D-D neutron generator

    Science.gov (United States)

    Ji, Q.; Lin, C.-J.; Tindall, C.; Garcia-Sciveres, M.; Schenkel, T.; Ludewigt, B. A.

    2017-05-01

    Tagging of neutrons (2.45 MeV) with their associated 3He particles from deuterium-deuterium (D-D) fusion reactions has been demonstrated in a compact neutron generator setup enabled by a high brightness, microwave-driven ion source with a high fraction of deuterons. Energy spectra with well separated peaks of the D-D fusion reaction products, 3He, tritons, and protons, were measured with a silicon PIN diode. The neutrons were detected using a liquid scintillator detector with pulse shape discrimination. By correlating the 3He detection events with the neutron detection in time, we demonstrated the tagging of emitted neutrons with 3He particles detected with a Si PIN diode detector mounted inside the neutron generator vacuum vessel.

  18. Visualization and measurement of thermo-fluid phenomena by using neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Mishima, Kaichiro [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1998-03-01

    Recently, neutron radiography is rapidly expanding its application to various fields incollaboration with the development of electronic imaging techniques. Particularly, in the field of thermal and fluid engineering, it has drawn much attention as an innovative and non-intrusive method for fluid-visualization and measurement using neutron beam as a probe. In this paper, some examples of applications are introduced on the high frame-rate imaging, quantification method (especially, determination of void fraction), and multidimensional measurement which are much interested in fluid measurement in relation to light water reactor safety and thermohydraulics. (author)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  20. Response of different neutrons measurers with the new ICRP 60

    International Nuclear Information System (INIS)

    Bourgois, L.; Girod, M.; Ferreux, L.

    1999-01-01

    The publication 60 of the ICRP has modified the factors of fluence conversion dose equivalent for the neutrons. The majority of neutrons counters are from old generation. The comparison of the two curves can give factor 2 in the measure of dose equivalents according to neutrons energy (700 keV). It is important to check in function of energy the response of neutrons counters used in radiation protection by taking the conversion factors of the ICRP 60 given in the ICRP 74. Three neutrons measurers of the old generation and a new one have been tested. The different devices answer better the new ICRP than the ICRP 21; (the devices were Studvick, Cramal and Dineutron counters, Berthold LB 123-N for the new generation and they have been tested with neutrons sources: Am-Li, Am-F, Am-B, Am-Be). (N.C.)