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Sample records for absolute neutron spectrum

  1. An absolute measurement of 252Cf prompt fission neutron spectrum at low energy range

    International Nuclear Information System (INIS)

    Lajtai, A.; Dyachenko, P.P.; Kutzaeva, L.S.; Kononov, V.N.; Androsenko, P.A.; Androsenko, A.A.

    1983-01-01

    Prompt neutron energy spectrum at low energies (25 keV 252 Cf spontaneous fission has been measured with a time-of-flight technique on a 30 cm flight-path. Ionization chamber and lithium-glass were used as fission fragment and neutron detectors, respectively. Lithium glasses of NE-912 (containing 6 Li) and of NE-913 (containing 7 Li) 45 mm in diameter and 9.5 mm in thickness have been employed alternatively, for the registration of fission neutrons and gammas. For the correct determination of the multiscattering effects - the main difficulty of the low energy neutron spectrum measurements - a special geometry for the neutron detector was used. Special attention was paid also to the determination of the absolute efficiency of the neutron detector. The real response function of the spectrometer was determined by a Monte-Carlo calculation. The scattering material content of the ionization chamber containing a 252 Cf source was minimized. As a result of this measurement a prompt fission neutron spectrum of Maxwell type with a T=1.42 MeV parameter was obtained at this low energy range. We did not find any neutron excess or irregularities over the Maxwellian. (author)

  2. PC based system for absolute neutron flux-spectrum measurements. Final report for the period 1 April 1994 - 15 December 1997

    International Nuclear Information System (INIS)

    Dobrea, D.

    1998-01-01

    When measuring absolute neutron flux-spectra, thin detector foils are irradiated in a neutron field. The absolute activity of isotopes generated by activation or fission reactions (fission products) is then measured, using an efficiency calibrated high resolution gamma-ray spectroscopy system, and the absolute reaction rates for their production is determined. Finally the flux-spectrum is determined based upon the reaction rate values. A general method to obtain flux-spectra from the reaction rate is the unfolding method. The above process involves computations of photopeak area, reaction rate, flux perturbation corrections and flux-spectrum. The PC's are well suited for the data processing system outlined above. Using available interfaces, the PC's can be involved even in the data acquisition. Graphical facilities allow decision upon the data processing flow

  3. The magnetic recoil spectrometer for measurements of the absolute neutron spectrum at OMEGA and the NIF.

    Science.gov (United States)

    Casey, D T; Frenje, J A; Johnson, M Gatu; Séguin, F H; Li, C K; Petrasso, R D; Glebov, V Yu; Katz, J; Magoon, J; Meyerhofer, D D; Sangster, T C; Shoup, M; Ulreich, J; Ashabranner, R C; Bionta, R M; Carpenter, A C; Felker, B; Khater, H Y; LePape, S; MacKinnon, A; McKernan, M A; Moran, M; Rygg, J R; Yeoman, M F; Zacharias, R; Leeper, R J; Fletcher, K; Farrell, M; Jasion, D; Kilkenny, J; Paguio, R

    2013-04-01

    The neutron spectrum produced by deuterium-tritium (DT) inertial confinement fusion implosions contains a wealth of information about implosion performance including the DT yield, ion-temperature, and areal-density. The Magnetic Recoil Spectrometer (MRS) has been used at both the OMEGA laser facility and the National Ignition Facility (NIF) to measure the absolute neutron spectrum from 3 to 30 MeV at OMEGA and 3 to 36 MeV at the NIF. These measurements have been used to diagnose the performance of cryogenic target implosions to unprecedented accuracy. Interpretation of MRS data requires a detailed understanding of the MRS response and background. This paper describes ab initio characterization of the system involving Monte Carlo simulations of the MRS response in addition to the commission experiments for in situ calibration of the systems on OMEGA and the NIF.

  4. The magnetic recoil spectrometer for measurements of the absolute neutron spectrum at OMEGA and the NIF

    International Nuclear Information System (INIS)

    Casey, D. T.; Frenje, J. A.; Gatu Johnson, M.; Séguin, F. H.; Li, C. K.; Petrasso, R. D.; Glebov, V. Yu.; Katz, J.; Magoon, J.; Meyerhofer, D. D.; Sangster, T. C.; Shoup, M.; Ulreich, J.; Ashabranner, R. C.; Bionta, R. M.; Carpenter, A. C.; Felker, B.; Khater, H. Y.; LePape, S.; MacKinnon, A.

    2013-01-01

    The neutron spectrum produced by deuterium-tritium (DT) inertial confinement fusion implosions contains a wealth of information about implosion performance including the DT yield, ion-temperature, and areal-density. The Magnetic Recoil Spectrometer (MRS) has been used at both the OMEGA laser facility and the National Ignition Facility (NIF) to measure the absolute neutron spectrum from 3 to 30 MeV at OMEGA and 3 to 36 MeV at the NIF. These measurements have been used to diagnose the performance of cryogenic target implosions to unprecedented accuracy. Interpretation of MRS data requires a detailed understanding of the MRS response and background. This paper describes ab initio characterization of the system involving Monte Carlo simulations of the MRS response in addition to the commission experiments for in situ calibration of the systems on OMEGA and the NIF.

  5. Measuring the absolute DT neutron yield using the Magnetic Recoil Spectrometer at OMEGA and the NIF

    Energy Technology Data Exchange (ETDEWEB)

    Mackinnon, A; Casey, D; Frenje, J A; Johnson, M G; Seguin, F H; Li, C K; Petrasso, R D; Glebov, V Y; Katz, J; Knauer, J; Meyerhofer, D; Sangster, T; Bionta, R; Bleuel, D; Hachett, S P; Hartouni, E; Lepape, S; Mckernan, M; Moran, M; Yeamans, C

    2012-05-03

    A Magnetic Recoil Spectrometer (MRS) has been installed and extensively used on OMEGA and the National Ignition Facility (NIF) for measurements of the absolute neutron spectrum from inertial confinement fusion (ICF) implosions. From the neutron spectrum measured with the MRS, many critical implosion parameters are determined including the primary DT neutron yield, the ion temperature, and the down-scattered neutron yield. As the MRS detection efficiency is determined from first principles, the absolute DT neutron yield is obtained without cross-calibration to other techniques. The MRS primary DT neutron measurements at OMEGA and the NIF are shown to be in excellent agreement with previously established yield diagnostics on OMEGA, and with the newly commissioned nuclear activation diagnostics on the NIF.

  6. Activation method for measurement of neutron spectrum parameters

    International Nuclear Information System (INIS)

    Efimov, B.V.; Demidov, A.M.; Ionov, V.S.; Konjaev, S.I.; Marin, S.V.; Bryzgalov, V.I.

    2007-01-01

    Experimental researches of spectrum parameters of neutrons at nuclear installations RRC KI are submitted. The installations have different designs of the cores, reflector, parameters and types of fuel elements. Measurements were carried out with use of the technique developed in RRC KI for irradiation resonance detectors UKD. The arrangement of detectors in the cores ensured possibility of measurement of neutron spectra with distinguished values of parameters. The spectrum parameters which are introduced by parametrical representation of a neutrons spectrum in the form corresponding to formalism Westcott. On experimental data were determinate absolute values of density neutron flux (DNF) in thermal and epithermal area of a spectrum (F t , f epi ), empirical dependence of temperature of neutron gas (Tn) on parameter of a rigidity of a spectrum (z), density neutron flux in transitional energy area of the spectrum. Dependences of spectral indexes of nuclides (UDy/UX), included in UKD, from a rigidity z and-or temperatures of neutron gas Tn are obtained.B Tools of mathematical processing of results are used for activation data and estimation of parameters of a spectrum (F t , f epi , z, Tn, UDy/UX). In the paper are presented some results of researches of neutron spectrum parameters of the nuclear installations (Authors)

  7. Measuring the absolute deuterium-tritium neutron yield using the magnetic recoil spectrometer at OMEGA and the NIF.

    Science.gov (United States)

    Casey, D T; Frenje, J A; Gatu Johnson, M; Séguin, F H; Li, C K; Petrasso, R D; Glebov, V Yu; Katz, J; Knauer, J P; Meyerhofer, D D; Sangster, T C; Bionta, R M; Bleuel, D L; Döppner, T; Glenzer, S; Hartouni, E; Hatchett, S P; Le Pape, S; Ma, T; MacKinnon, A; McKernan, M A; Moran, M; Moses, E; Park, H-S; Ralph, J; Remington, B A; Smalyuk, V; Yeamans, C B; Kline, J; Kyrala, G; Chandler, G A; Leeper, R J; Ruiz, C L; Cooper, G W; Nelson, A J; Fletcher, K; Kilkenny, J; Farrell, M; Jasion, D; Paguio, R

    2012-10-01

    A magnetic recoil spectrometer (MRS) has been installed and extensively used on OMEGA and the National Ignition Facility (NIF) for measurements of the absolute neutron spectrum from inertial confinement fusion implosions. From the neutron spectrum measured with the MRS, many critical implosion parameters are determined including the primary DT neutron yield, the ion temperature, and the down-scattered neutron yield. As the MRS detection efficiency is determined from first principles, the absolute DT neutron yield is obtained without cross-calibration to other techniques. The MRS primary DT neutron measurements at OMEGA and the NIF are shown to be in excellent agreement with previously established yield diagnostics on OMEGA, and with the newly commissioned nuclear activation diagnostics on the NIF.

  8. Neutron Spectrum Parameters In Inner Irradiation Channel Of The Nigeria Research Reactor-1 (NIRR-1) For Use In Absolute And KO-NAA Methods

    International Nuclear Information System (INIS)

    Jonah, S.A; Balogun, G.I; Mayaki, M.C.

    2004-01-01

    In Nigeria, the first Nuclear Reactor achieved critically on February 03, 2004 at about 11:35 GMT and has been commissioned or training and research. It is a Miniature Neutron Source Reactor (MNSR), code-named Nigeria Research Reactor-1 (NIRR-1). NIRR-1 has a tan-in-pool structural configuration and a nominal thermal power rating of 30 Kw. With a built-in clean old core excess reactivity of 3.77 mk determined during the on-site zero and critically experimental, the reactor can operate for a n.cm-2 .s-1 in the inner irradiation channels). Under these conditions, the reactor can operate with the same fuel loading for over ten years with a burn-up of <1%. A detailed description of operating characteristics for NIRR-1, measured during the on-site zero-power and criticality experiments has been given elsewhere. In order to extend its utilization to include absolute and ko-NAA methods, the neutron spectrum parameters in the irradiation channels: power and critically experiments has been given elsewhere. In order to extend it's the irradiation channels: thermal-to-epithermal flux ration, F; and epithermal flux shape factor, a in both the inner and outer irradiation channels must be determined experimentally. In this work, we have developed and experimental procedure for monitoring the neutron spectrum parameters in an inner irradiation channel based on irradiation and gamma-ray counting of detector foils via (n,y), (n,p) and (n,a) dosimetry reactions. Results obtained indicate that a thermal neutron flux of (5.14+-0.02) x 1011 n/c m2.s determined by foil activation method in the inner irradiation channel, B2, at a power level of 15.5 kw corresponds to the flux indicators on the control console and the micro-computer control system respectively. Other parameters of the neutron spectrum determined for inner irradiation channel B2, are: a -0.0502+0.003; 18.92+-0.14; F = 3.87=0.23. The method was validated through the comparison of our result with published neutron spectrum

  9. Absolute measurement of neutron fluxes inside the reactor core

    International Nuclear Information System (INIS)

    Ajdacic, S. V.

    1964-10-01

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li 6 -semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li 6 -semiconductor spectrometer with plane geometry is given. A new type of Li 6 -semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li 6 -spectrometer made (author)

  10. Measurements of the absolute neutron fluence spectrum emitted at 00 and 900 from the Little-Boy replica

    International Nuclear Information System (INIS)

    Roberts, J.H.; Gold, R.; Preston, C.C.

    1986-01-01

    Nuclear research emulsions (NRE) have been used to characterize the neutron spectrum emitted by the Little-Boy replica. NRE were irradiated at the Little-Boy surface, as well as approximately 2 m from the center of the Little-Boy replica, using polar angles of 0 0 , 30 0 , 60 0 , and 90 0 . For the NRE exposed at 2 m, neutron background was determined using shadow shields of borated polyethylene. Emulsion scanning to date has concentrated exclusively on the 2-m, 0 0 and 2-m, 90 0 locations. Approximately 5000 proton-recoil tracks have been measured in NRE irradiated at each of these locations. At the 2-m, 90 0 location, the NRE neutron spectrum extends from 0.37 MeV up to 8.2 MeV; whereas the NRE neutron spectrum at the 2-m, 0 0 location is much softer and extends only up to 2.7 MeV. NRE neutron spectrometry results at these two locations are compared with both liquid scintillator neutron spectrometry and Monte Carlo calculations. (author)

  11. Absolute measurement of neutron fluxes inside the reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, S V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-10-15

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li{sup 6}-semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li{sup 6}-semiconductor spectrometer with plane geometry is given. A new type of Li{sup 6}-semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li{sup 6}-spectrometer made (author)

  12. Absolute instrumental neutron activation analysis at Lawrence Livermore Laboratory

    International Nuclear Information System (INIS)

    Heft, R.E.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Erdtmann, G.

    1993-08-01

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

  14. Neutron spectrum unfolding: Pt. 2

    International Nuclear Information System (INIS)

    Matiullah; Wiyaja, D.S.; Berzonis, M.A.; Bondars, H.; Lapenas, A.A.; Kudo, K.; Majeed, A.; Durrani, S.A.

    1991-01-01

    In Part I of this paper, we described the use of the computer code SAIPS in neutron spectrum unfolding. Here in Part II, we present our experimental work carried out to study the shape of the neutron spectrum in different experimental channels of a 5 MW light-water cooled and moderated research reactor. The spectral neutron flux was determined using various fission foils (placed in close contact with mica track detectors) and activation detectors. From the measured activities, the neutron spectrum was unfolded by SAIPS. (author)

  15. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Science.gov (United States)

    Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

    2016-02-01

    The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  16. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Directory of Open Access Journals (Sweden)

    Wagemans Jan

    2016-01-01

    Full Text Available The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  17. A measurement of the absolute neutron beam polarization produced by an optically pumped 3He neutron spin filter

    International Nuclear Information System (INIS)

    Rich, D.R.; Bowman, J.D.; Crawford, B.E.; Delheij, P.P.J.; Espy, M.A.; Haseyama, T.; Jones, G.; Keith, C.D.; Knudson, J.; Leuschner, M.B.; Masaike, A.; Masuda, Y.; Matsuda, Y.; Penttilae, S.I.; Pomeroy, V.R.; Smith, D.A.; Snow, W.M.; Szymanski, J.J.; Stephenson, S.L.; Thompson, A.K.; Yuan, V.

    2002-01-01

    The capability of performing accurate absolute measurements of neutron beam polarization opens a number of exciting opportunities in fundamental neutron physics and in neutron scattering. At the LANSCE pulsed neutron source we have measured the neutron beam polarization with an absolute accuracy of 0.3% in the neutron energy range from 40 meV to 10 eV using an optically pumped polarized 3 He spin filter and a relative transmission measurement technique. 3 He was polarized using the Rb spin-exchange method. We describe the measurement technique, present our results, and discuss some of the systematic effects associated with the method

  18. Absolute calibration of the neutron yield measurement on JT-60 Upgrade

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Takeuchi, Hiroshi; Barnes, C.W.

    1991-10-01

    Absolutely calibrated measurements of the neutron yield are important for the evaluation of the plasma performance such as the fusion gain Q in DD operating tokamaks. Total neutron yield is measured with 235 U and 238 U fission chambers and 3 He proportional counters in JT-60 Upgrade. The in situ calibration was performed by moving the 252 Cf neutron source toroidally through the JT-60 vacuum vessel. Detection efficiencies of three 235 U and two 3 He detectors were measured for 92 locations of the neutron point source in toroidal scans at two different major radii. The total detection efficiency for the torus neutron source was obtained by averaging the point efficiencies over the whole toroidal angle. The uncertainty of the resulting absolute plasma neutron source calibration is estimated to be ± 10%. (author)

  19. Contributions to the study of fast neutron spectrum in the 10 keV - 3 MeV range

    International Nuclear Information System (INIS)

    Garlea, I.

    1979-01-01

    The main objective of the work presented in this thesis was to create a fast neutron spectrum corresponding to the conditions required for a reference neutron field. The reference system for the fast neutron dosimetry in reactors, which the author promoted, is referred to as ΣΣ-ITN in the books. The conditions for introducing the ΣΣ systems into the thermal columns have been determined. The original contribution consists in determining the Westcott parameters of reactions 151 Eu(n,γ) 152 Eu and 176 Lu(n,γ) 17 +H7Lu used as thermal spectrum factors. The neutron description of the spectrum in cavity ΣΣ revealed that it is a Maxwell thermal spectrum displaying a temperature of 305+-7 deg C and a very small epithermal component (phisub(epi)/phisub(thermal) =4,5.10 -4 ). Better methods for determining reaction absolute rates resulted in less errors in calculating the microscopic integral sections mediated on the ΣΣ spectrum; there are under 5% errors for the fission cross sections and between 3% and 8% errors for the activating ones. The section values determined by the author have been included into the EXFOR library (IAEA); they are considered as reference measuremtns for the nuclear data improvement program. Testing the proposed method for the TRIGA on the ΣΣ-INT system proved that the multiple foil method provides correct results for both describing the spectral shape and for obtaining absolute values of the flux. Taking into account that the ΣΣ-ITN spectrum is a rapid one, the proposed method could not be tested within the low energy thermal and epithermal domain. For testing the method on an operational reactor, the core of the VVR-S IFIN reactor was employed. Due to the spectral structure of this reactor, it was possible to test the procedure within the whole energy range. In this view, the 5/10 core channel was selected which is similar to the channel required for measurements in the TRIGA-ROMANIA reactor. The absolute spectrum values are given in

  20. New fission-neutron-spectrum representation for ENDF

    International Nuclear Information System (INIS)

    Madland, D.G.

    1982-04-01

    A new representation of the prompt fission neutron spectrum is proposed for use in the Evaluated Nuclear Data File (ENDF). The proposal is made because a new theory exists by which the spectrum can be accurately predicted as a function of the fissioning nucleus and its excitation energy. Thus, prompt fission neutron spectra can be calculated for cases where no measurements exist or where measurements are not possible. The mathematical formalism necessary for application of the new theory within ENDF is presented and discussed for neutron-induced fission and spontaneous fission. In the case of neutron-induced fission, expressions are given for the first-chance, second-chance, third-chance, and fourth-chance fission components of the spectrum together with that for the total spectrum. An ENDF format is proposed for the new fission spectrum representation, and an example of the use of the format is given

  1. Intermediate neutron spectrum problems and the intermediate neutron spectrum experiment

    International Nuclear Information System (INIS)

    Jaegers, P.J.; Sanchez, R.G.

    1996-01-01

    Criticality benchmark data for intermediate energy spectrum systems does not exist. These systems are dominated by scattering and fission events induced by neutrons with energies between 1 eV and 1 MeV. Nuclear data uncertainties have been reported for such systems which can not be resolved without benchmark critical experiments. Intermediate energy spectrum systems have been proposed for the geological disposition of surplus fissile materials. Without the proper benchmarking of the nuclear data in the intermediate energy spectrum, adequate criticality safety margins can not be guaranteed. The Zeus critical experiment now under construction will provide this necessary benchmark data

  2. Development and application of a detector for absolute measurement of neutron fluence rate in MeV region

    International Nuclear Information System (INIS)

    Silva Dias, M. da.

    1988-01-01

    The development and performance of the DTS (Dual Thin Scintillator) for the absolute measurement of the neutron fluence rate between 1 and 15 MeV is decribed. The DTS detector consists of a pair of organic scintillators in a dual configuration, where the incident produces a proton-recoil which is detected in a 2Π geometry therefore avoiding the effect of the escape of protons. Thin scintillators are used resulting in small multiple scattering corrections. The theoretical caluclations of detector efficiency and proton-recoil spectrum were performed by means of a Monte Carlos code - CARLO DTS. The calculated efficiency was compared to the experimental one at two neutron energies namely 2.446 MeV and 14.04 MeV applying the Time Correlated Associated Particle technique. The theoretical and experimental efficiencies agreed within the experimental uncertainties of 1.44% and 0.77%, respectively. The performance of the DTS has been verified in an absolute 235 U(n,f) cross section measurement between 1 and 6 MeV neutron energy. The cross section results were compared to those obtained replacing the DTS detector by the NBS (National Bureau of Standards, USA) Black Neutron Detector. The agreement was excellent in the overlapping energy interval of the two experiments (between 1 and 3 MeV), within the estimated uncertainly in the range of 1,0 to 1,7%. The agreement with the most recent evaluation from the ENDF/B-VI was excellent in almost all the energy range between 1 and 6 MeV. The 235 U(n,f) cross section, average over the 252 Cf fission neutron spectrum has been evaluated. The result including the cross section values of the present work was 1220 mb, in excellent agreement with the average value among the most recent measurements, 1227 +- 12 mb, and with the value 1213 mb, using the ENDF/B-VI data. (author) [pt

  3. Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping

    International Nuclear Information System (INIS)

    Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi

    2012-01-01

    The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. - Highlights: ► The effects of silicon cross section and neutron spectrum on the radial uniformity in NTD were experimentally investigated. ► The numerical results using silicon single crystal cross section reveal good agreements. ► The radial uniformity in hard neutron spectrum was more flat than that in soft spectrum. ► The silicon single crystal cross section and hard neutron spectrum are recommended for numerical analyses and radial uniformity flattening in NTD, respectively.

  4. 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.

  5. Neutron activation analysis of certified samples by the absolute method

    Science.gov (United States)

    Kadem, F.; Belouadah, N.; Idiri, Z.

    2015-07-01

    The nuclear reactions analysis technique is mainly based on the relative method or the use of activation cross sections. In order to validate nuclear data for the calculated cross section evaluated from systematic studies, we used the neutron activation analysis technique (NAA) to determine the various constituent concentrations of certified samples for animal blood, milk and hay. In this analysis, the absolute method is used. The neutron activation technique involves irradiating the sample and subsequently performing a measurement of the activity of the sample. The fundamental equation of the activation connects several physical parameters including the cross section that is essential for the quantitative determination of the different elements composing the sample without resorting to the use of standard sample. Called the absolute method, it allows a measurement as accurate as the relative method. The results obtained by the absolute method showed that the values are as precise as the relative method requiring the use of standard sample for each element to be quantified.

  6. Study of the environmental neutron spectrum at Zacatecas city

    International Nuclear Information System (INIS)

    Vega C, H.R.

    2003-01-01

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

  7. Study of the environmental neutron spectrum at Zacatecas city

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  8. 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)

  9. Neutron spectrum measurements from a neutron guide tube facility at the ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R M.A.; El-Sayed, L A.A.; El-Kady, A S.I. [Reactor and Neutron Physics Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The present work deals with measurements of the neutron spectrum emitted from a neutron guide tube (NGT) recently installed at one of the ETRR-1 reactor horizontal channels designed to deliver thermal neutrons, free from fast neutrons and gamma ray background, to a fourier reverse-time-of-flight (RTOF) diffractometer. The measurements were performed using a {sup 6} Li glass scintillation detector combined with a multichannel analyzer set at channel width 4 M sec and installed at 3.4 m from a disc Fermi chopper. Also a theoretical model was specially developed for the neutron spectrum calculations. According to the model developed, the spectrum calculated was found to be in good agreement with the measured one. It was found, both from measurements and calculations, that the spectrum emitted from the NGT covers, after transmission through a fourier chopper, neutron wavelengths from 1-4 A adequate for neutron diffraction measurements at D values between 0.71-2.9 A respectively. 6 FIGS.

  10. Prompt neutron spectrum of the spontaneous fission of californium-252

    International Nuclear Information System (INIS)

    Zamyatnin, Yu.S.; Kroshkin, N.I.; Korostylev, V.A.; Nefedov, V.N.; Ryazanov, D.K.; Starostov, B.I.; Semenov, A.F.

    1976-01-01

    The californium-252 spontaneous fission neutron spectrum was measured in the energy range of 0.01 to 10 MeV by the time-of-flight technique using various neutron detectors. The measurements of 252 Cf neutron spectrum at energies of 0.01 to 5 MeV were performed as a function of fission fragment kinetic energy. The mean neutron spectrum energy in the range of 0.7 to 10 MeV was found from the results of measurements. The irregularity in the 252 Cf neutron spectrum in the neutron energy range of less than 0.7 MeV compared to theoretical values is discussed. The mechanism of 252 Cf neutron emission is also discussed on the basis of neutron yield angle measurements. 12 references

  11. Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping.

    Science.gov (United States)

    Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi

    2012-01-01

    The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. Copyright © 2011 Elsevier Ltd. All rights reserved.

  12. Automated absolute activation analysis with californium-252 sources

    International Nuclear Information System (INIS)

    MacMurdo, K.W.; Bowman, W.W.

    1978-09-01

    A 100-mg 252 Cf neutron activation analysis facility is used routinely at the Savannah River Laboratory for multielement analysis of many solid and liquid samples. An absolute analysis technique converts counting data directly to elemental concentration without the use of classical comparative standards and flux monitors. With the totally automated pneumatic sample transfer system, cyclic irradiation-decay-count regimes can be pre-selected for up to 40 samples, and samples can be analyzed with the facility unattended. An automatic data control system starts and stops a high-resolution gamma-ray spectrometer and/or a delayed-neutron detector; the system also stores data and controls output modes. Gamma ray data are reduced by three main programs in the IBM 360/195 computer: the 4096-channel spectrum and pertinent experimental timing, counting, and sample data are stored on magnetic tape; the spectrum is then reduced to a list of significant photopeak energies, integrated areas, and their associated statistical errors; and the third program assigns gamma ray photopeaks to the appropriate neutron activation product(s) by comparing photopeak energies to tabulated gamma ray energies. Photopeak areas are then converted to elemental concentration by using experimental timing and sample data, calculated elemental neutron capture rates, absolute detector efficiencies, and absolute spectroscopic decay data. Calculational procedures have been developed so that fissile material can be analyzed by cyclic neutron activation and delayed-neutron counting procedures. These calculations are based on a 6 half-life group model of delayed neutron emission; calculations include corrections for delayed neutron interference from 17 O. Detection sensitivities of 239 Pu were demonstrated with 15-g samples at a throughput of up to 140 per day. Over 40 elements can be detected at the sub-ppM level

  13. Spectrum and H(10) of secondary neutrons around Linacs

    International Nuclear Information System (INIS)

    Ortiz H, A.; Hernandez A, B.; Vega C, H. R.; Hernandez D, V. M.; Rivera M, T.

    2009-10-01

    Neutron spectrum and ambient dose equivalent has been measured around two 10 MV linear accelerators. Accelerators are Siemens, one is a Mevatron model while another is the Primus. Main differences between those models are the beam collimator and the vault room. Here, Bonner sphere spectrometer with a passive thermal neutron detector has been utilized to measure the neutron spectrum inside the vault. Using an active detector the neutron spectrum was measured by the vaults door of both accelerators. With a neutron area monitor the dose equivalent was measured by the doors. Neutron strength, total fluence rate and ambient dose equivalent were compared, from this was found that shielding conditions are better in the Primus model. (Author)

  14. Neutron spectrum unfolding using neural networks

    International Nuclear Information System (INIS)

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

    2004-01-01

    An artificial neural network has been designed to obtain the neutron spectra from the Bonner spheres spectrometer's count rates. The neural network was trained using a large set of neutron spectra compiled by the International Atomic Energy Agency. These include spectra from iso- topic neutron sources, reference and operational neutron spectra obtained from accelerators and nuclear reactors. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra and UTA4 matrix were used to calculate the expected count rates in Bonner spheres spectrometer. These count rates were used as input and correspondent spectrum was used as output during neural network training. The network has 7 input nodes, 56 neurons as hidden layer and 31 neurons in the output layer. After training the network was tested with the Bonner spheres count rates produced by twelve neutron spectra. The network allows unfolding the neutron spectrum from count rates measured with Bonner spheres. Good results are obtained when testing count rates belong to neutron spectra used during training, acceptable results are obtained for count rates obtained from actual neutron fields; however the network fails when count rates belong to monoenergetic neutron sources. (Author)

  15. Least-squares adjustment of a 'known' neutron spectrum: The importance of the covariance matrix of the input spectrum

    International Nuclear Information System (INIS)

    Mannhart, W.

    1986-01-01

    Based on the responses of 25 different neutron activation detectors, the neutron spectrum of Cf-252 hs been adjusted with least-squares methods. For a fixed input neutron spectrum, the covariance matrix of this spectrum has been systematically varied to investigate the influence of this matrix on the final result. The investigation showed that the adjusted neutron spectrum is rather sensitive to the structure of the covariance matrix for the input spectrum. (author)

  16. A neutron spectrum unfolding code based on iterative procedures

    International Nuclear Information System (INIS)

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

    2012-10-01

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

  17. Spectrum-averaged cross-section measurement of /sup 103/Rh(n,n)/sup 103m/Rh in the /sup 252/Cf fission neutron spectrum

    International Nuclear Information System (INIS)

    Lamaze, G.P.; Schima, F.J.; Eisenhauer, C.M.; Spiegel, V.

    1988-01-01

    Because of the similarity in energy dependence of the /sup 103/Rh(n,n') differential cross section to the kerma muscle response function for neutrons, rhodium may be useful as a neutron kerma monitor. In support of its use as a neutron monitor, the spectrum-averaged cross section σ-bar has been measured for a /sup 252/Cf fission neutron spectrum. Pairs of thin rhodium samples were irradiated on opposite sides of a thinly encapsulated /sup 252/Cf neutron source. The neutron emission rate of the /sup 252/Cf source was determined by the manganous sulfate (MnSO/sub 4/) bath technique. In this method, the californium source emission rate is determined by comparison to the known emission rate of NBS-I, a standard radium-beryllium neutron source. The neutron fluence incident on the rhodium samples is determined from the californium source strength, average sample-to-source distance, and the duration of the irradiation. Corrections are made for neutron scattering saturation of activity, and attenuation of the X rays by the sample during counting. The X rays were detected with an intrinsic germanium detector designed specifically for low-energy X-ray detection. The activity was not determined by absolute counting so that the final results depend on the value of P/sub Κx/, to total Κ X-ray emission probability. The results of five separate irradiations yield a value of σ-bar . P/sub Κx/ = 62.3 +- 1.9 mb. Using the most recently published value of P/sub Κx/ gives a value of σ-bar = 739 +- 22 mb. A discussion of systematic uncertainties is given

  18. Neutron spectrum measurement using rise-time discrimination method

    International Nuclear Information System (INIS)

    Luo Zhiping; Suzuki, C.; Kosako, T.; Ma Jizeng

    2009-01-01

    PSD method can be used to measure the fast neutron spectrum in n/γ mixed field. A set of assemblies for measuring the pulse height distribution of neutrons is built up,based on a large volume NE213 liquid scintillator and standard NIM circuits,through the rise-time discrimination method. After that,the response matrix is calculated using Monte Carlo method. The energy calibration of the pulse height distribution is accomplished using 60 Co radioisotope. The neutron spectrum of the mono-energetic accelerator neutron source is achieved by unfolding process. Suggestions for further improvement of the system are presented at last. (authors)

  19. Measured and Predicted Variations in Fast Neutron Spectrum when Penetrating Laminated Fe-D2O

    International Nuclear Information System (INIS)

    Aalto, E.; Sandlin, R.; Fraeki, R.

    1965-09-01

    Variations of the fast neutron spectrum in thin regions of alternating Fe and D O have been studied using threshold detectors (ln(n, n' ), S(n, p), Al(n, α)). The results have been compared to those calculated by two shielding codes (NRN and RASH D) of multigroup removal-diffusion type. The absolute fast spectrum calculated in our rather complicated configurations was found to agree with measurements within the same accuracy (a factor of two) as did the thermal flux. The calculated spectrum is slightly harder than the measured one, but the detailed variations (covering the range 1:5) in the form of the spectrum when penetrating Fe agree with observations to within 15-20 %. In and Al activities were found to be proportional to the integrated flux over 1 MeV throughout the whole configuration, while S showed the least proportionality

  20. A Wide Spectrum Neutron Polarizer for a Pulsed Neutron Source

    International Nuclear Information System (INIS)

    Nikitenko, Yu.V.

    1994-01-01

    A wide spectrum neutron polarizer for a pulsed neutron source is considered. The polarizer is made in a form of a set of magnetized mirrors placed on a drum. Homogeneous rotation of the polarizer is synchronized with the power pulses of the neutron source. The polarizer may be utilized in a collimated neutron beam with cross section of the order of magnitude of 100 cm 2 within a wavelength from 2 up to 20 A on sources with a pulse repetition frequency up to 50 Hz. (author). 5 refs.; 3 figs

  1. Neutron spectrum and dose-equivalent in shuttle flights during solar maximum

    Energy Technology Data Exchange (ETDEWEB)

    Keith, J E; Badhwar, G D; Lindstrom, D J [National Aeronautics and Space Administration, Houston, TX (United States). Lyndon B. Johnson Space Center

    1992-01-01

    This paper presents unambiguous measurements of the spectrum of neutrons found in spacecraft during spaceflight. The neutron spectrum was measured from thermal energies to about 10 MeV using a completely passive system of metal foils as neutron detectors. These foils were exposed to the neutron flux bare, covered by thermal neutron absorbers (Gd) and inside moderators (Bonner spheres). This set of detectors was flown on three U.S. Space Shuttle flights, STS-28, STS-36 and STS-31, during the solar maximum. We show that the measurements of the radioactivity of these foils lead to a differential neutron energy spectrum in all three flights that can be represented by a power law, J(E){approx equal}E{sup -0.765} neutrons cm{sup -2} day {sup -1} MeV{sup -1}. We also show that the measurements are even better represented by a linear combination of the terrestrial neutron albedo and a spectrum of neutrons locally produced in a aluminium by protons, computed by a previous author. We use both approximations to the neutron spectrum to produce a worst case and most probable case for the neutron spectra and the resulting dose-equivalents, computed using ICRP-51 neutron fluence-dose conversion tables. We compare these to the skin dose-equivalents due to charged particles during the same flights. (author).

  2. Neutrons from medical electron accelerators

    International Nuclear Information System (INIS)

    Swanson, W.P.; McCall, R.C.

    1979-06-01

    The significant sources of photoneutrons within a linear-accelerator treatment head are identified and absolute estimates of neutron production per treatment dose are given for typical components. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. It is found that the high-Z materials within the treatment head do not significantly alter the neutron fluence, but do substantially reduce the average energy of the transmitted spectrum. Reflected neutrons from the concrete treatment room contribute to the neutron fluence, but not substantially to the patient integral dose, because of a further reduction in average energy. Absolute depth-dose distributions for realistic neutron spectra are calculated, and a rapid falloff with depth is found

  3. Measured and Predicted Variations in Fast Neutron Spectrum when Penetrating Laminated Fe-D{sub 2}O

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, E; Sandlin, R; Fraeki, R

    1965-09-15

    Variations of the fast neutron spectrum in thin regions of alternating Fe and D{sub O} have been studied using threshold detectors (ln(n, n' ), S(n, p), Al(n, {alpha})). The results have been compared to those calculated by two shielding codes (NRN and RASH D) of multigroup removal-diffusion type. The absolute fast spectrum calculated in our rather complicated configurations was found to agree with measurements within the same accuracy (a factor of two) as did the thermal flux. The calculated spectrum is slightly harder than the measured one, but the detailed variations (covering the range 1:5) in the form of the spectrum when penetrating Fe agree with observations to within 15-20 %. In and Al activities were found to be proportional to the integrated flux over 1 MeV throughout the whole configuration, while S showed the least proportionality.

  4. Purely absolutely continuous spectrum for almost Mathieu operators

    International Nuclear Information System (INIS)

    Chulaevsky, V.; Delyon, F.

    1989-01-01

    Using a recent result of Sinai, the authors prove that the almost Mathieu operators acting on l 2 (Z), (H αλ Psi)(n) = Ψ(n + 1) + Ψ(n - 1) + λ cos(ωn + α) Ψ(n), have a purely absolutely continuous spectrum for almost all α provided that ω is a good irrational and λ is sufficiently small. Furthermore, the generalized eigenfunctions are quasiperiodic

  5. Determining of the intermediate neutron spectrum in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-01-01

    The activation method for intermediate neutron spectrum determination is given in this paper. The intermediate neutron spectrum in experimental fuel channel (EFC) at the RB reactor is determined om the basis of this method. The results of measurements are treated with PRAG code and will be treated with KRIFIT and TENET codes that are also developed. (author)

  6. Absolute measurement of the DT primary neutron yield on the National Ignition Facility

    Directory of Open Access Journals (Sweden)

    Leeper R.J.

    2013-11-01

    Full Text Available The measurement of the absolute neutron yield produced in inertial confinement fusion target experiments conducted on the National Ignition Facility (NIF is essential in benchmarking progress towards the goal of achieving ignition on this facility. This paper describes three independent diagnostic techniques that have been developed to make accurate and precise DT neutron yield measurements on the NIF.

  7. Absolute measurements of the fast neutron flux in the reactor RA

    Energy Technology Data Exchange (ETDEWEB)

    Berovic, N; Boreli, F; Dragin, R [Institute of Nuclear Sciences Boris Kidric, Department of physics, Vinca, Beograd (Serbia and Montenegro)

    1961-10-15

    The absolute neutron flux in the vertical VK-5 hole of the reactor RA was determined by using the {sup 27}Al (n, alpha) {sup 24}Na reaction, and by counting the {sup 24}Na - 2.5 MeV gamma line photopeak activity. A method for the determination of {sigma}{sub eff} as a mean value between the two large limiting cases of neutron spectra is used. The flux at the power level of 5 MW was found to be (2.5{+-}0.9){center_dot}10{sup 12}n/cm{sup 2}sec (author)

  8. Absolute measurement of the subcriticality based on the third order neutron correlation in consideration of the finite nature of neutron counts data

    International Nuclear Information System (INIS)

    Endo, Tomohiro; Kitamura, Yasunori; Yamane, Yoshihiro

    2003-01-01

    We have studied a measurement of subcriticality by using the neutron correlation method. Furuhashi proposed an absolute measurement of subcriticality by using the third order neutron correlation factor X in addition to the second order neutron correlation factor Y. In actual experiments, the number of neutron counts data is not infinity so that we take the effect of the finite nature of the neutron counts data into account. We derived new formulas in consideration of the number of data and verified them. (author)

  9. Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Vermeeren, L. [SCK/CEN, B-2400 Mol (Belgium)

    2001-07-01

    Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

  10. Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2001-01-01

    Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

  11. Absolute calibration of TFTR neutron detectors for D-T plasma operation

    International Nuclear Information System (INIS)

    Jassby, D.L.; Johnson, L.C.; Roquemore, A.L.; Strachan, J.D.; Johnson, D.W.; Medley, S.S.; Young, K.M.

    1995-03-01

    The two most sensitive TFTR fission-chamber detectors were absolutely calibrated in situ by a D-T neutron generator (∼5 x 10 7 n/s) rotated once around the torus in each direction, with data taken at about 45 positions. The combined uncertainty for determining fusion neutron rates, including the uncertainty in the total neutron generator output (±9%), counting statistics, the effect of coil coolant, detector stability, cross-calibration to the current mode or log Campbell mode and to other fission chambers, and plasma position variation, is about ±13%. The NE-451 (ZnS) scintillators and 4 He proportional counters that view the plasma in up to 10 collimated sightlines were calibrated by scanning. the neutron generator radially and toroidally in the horizontal midplane across the flight tubes of 7 cm diameter. Spatial integration of the detector responses using the calibrated signal per unit chord-integrated neutron emission gives the global neutron source strength with an overall uncertainty of ±14% for the scintillators and ±15% for the 4 He counters

  12. Experimental techniques for the consolidation of the neutron spectrum

    International Nuclear Information System (INIS)

    Chiaraviglio, N.; Bazzana, S.

    2013-01-01

    Unfolding techniques are widely known but their use is not widespread due to their complexity. Such procedure consists in the adjustment of calculated quantities to experimental results by the modification of the neutron spectrum, getting correction factors for the calculated quantities. In this work we describe the general procedure that must be executed for a neutron spectrum unfolding. (author) [es

  13. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor

    International Nuclear Information System (INIS)

    Lerner, A.M.; Madariaga, M.R.

    1998-01-01

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm 2 sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm 2 .sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm 2 .sec). ((1) According to the reaction Au 197 (n,γ)Au 198 , having a cross section of σ 0 =98.8b for thermal neutrons. (2) According to the reaction In 115 (n,n')In 115m , with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [es

  14. Determining {sup 252}Cf source strength by absolute passive neutron correlation counting

    Energy Technology Data Exchange (ETDEWEB)

    Croft, S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6166 (United States); Henzlova, D., E-mail: henzlova@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-06-21

    Physically small, lightly encapsulated, radionuclide sources containing {sup 252}Cf are widely used for a vast variety of industrial, medical, educational and research applications requiring a convenient source of neutrons. For many quantitative applications, such as detector efficiency calibrations, the absolute strength of the neutron emission is needed. In this work we show how, by using a neutron multiplicity counter the neutron emission rate can be obtained with high accuracy. This provides an independent and alternative way to create reference sources in-house for laboratories such as ours engaged in international safeguards metrology. The method makes use of the unique and well known properties of the {sup 252}Cf spontaneous fission system and applies advanced neutron correlation counting methods. We lay out the foundation of the method and demonstrate it experimentally. We show that accuracy comparable to the best methods currently used by national bodies to certify neutron source strengths is possible.

  15. 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)

  16. 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.

  17. Absolute determination of copper and silver in ancient coins using 14 MeV neutrons

    Science.gov (United States)

    Chalouhi, Ch.; Hourani, E.; Loos, R.; Melki, S.

    1982-09-01

    A method for absolute determination of copper and silver in ancient coins is described. Activation analysis by 14 MeV neutrons is performed. In the experimental procedure emphasis is placed on corrections for neutrons and gamma attenuation. In the analytical procedure, a multi linear-regression calculation is used to separate different contributions to the 511 keV gamma peak. The precision in the absolute determination of Cu and Ag is better than 2% in recent coins of definite shapes, whereas it is a somewhat lower in ancient coins of irregular shapes. The method was applied to ancient coins provided by the Museum of the American University of Beirut. Overall consistency and suitability of the method were obtained.

  18. Absolute determination of copper and silver in ancient coins using 14 MeV neutrons

    International Nuclear Information System (INIS)

    Chalouhi, C.; Hourani, E.; Loos, R.; Melki, S.

    1982-01-01

    A method for absolute determination of copper and silver in ancient coins is described. Activation analysis by 14 MeV neutrons is performed. In the experimental procedure emphasis is placed on corrections for neutrons and gamma attenuation. In the analytical procedure, a multi linear-regression calculation is used to separate different contributions to the 511 keV gamma peak. The precision in the absolute determination of Cu and Ag is better than 2% in recent coins of definite shapes, whereas it is a somewhat lower in ancient coins of irregular shapes. The method was applied to ancient coins provided by the Museum of the American University of Beirut. Overall consistency and suitability of the method were obtained. (orig.)

  19. Absolute determination of copper and silver in ancient coins using 14 MeV neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Chalouhi, C.; Hourani, E.; Loos, R.; Melki, S. (Faculty of Science, Beirut (Lebanon))

    1982-09-15

    A method for absolute determination of copper and silver in ancient coins is described. Activation analysis by 14 MeV neutrons is performed. In the experimental procedure emphasis is placed on corrections for neutrons and gamma attenuation. In the analytical procedure, a multi linear-regression calculation is used to separate different contributions to the 511 keV gamma peak. The precision in the absolute determination of Cu and Ag is better than 2% in recent coins of definite shapes, whereas it is a somewhat lower in ancient coins of irregular shapes. The method was applied to ancient coins provided by the Museum of the American University of Beirut. Overall consistency and suitability of the method were obtained.

  20. Modulation of the neutron spectrum for NCTB; Modulacion del espectro de neutrones para TCNB

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Letechipia de L, C.; Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No.10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2016-09-15

    Cancer is characterized by the uncontrolled growth of differentiated cells in a part of the organism. Currently in the world there are millions of people living with cancer. Glioblastoma multiform e is the most common and most aggressive of brain tumors and is very difficult to treat by conventional surgery, chemotherapy or radiation. The only viable alternative is its treatment through Neutron Capture Therapy in Boron (NCTB), since is a selective therapy that requires a drug with {sup 10}B (a non-radioactive isotope of boron) and a modulated neutron beam. Thermal neutrons are captured by {sup 10}B, because has a large effective section of thermal neutron absorption, in an exothermic reaction forming the nucleus composed of {sup 11}B in an excited state that induces its cleavage in two nuclei: {sup 7}Li and alpha particle ({sup 4}He). This process causes the destruction of cancer cells by direct DNA damage, without damaging normal tissue. One of the problems associated with this therapy is to have a neutron beam with adequate flow and spectrum. The neutron spectrum must be moderated and filtered from the characteristics of the source. To this end, the main sources of neutrons are nuclear research reactors and particle accelerators. The intensity of the flow should be 2 x 10 E{sup 9} n/cm{sup 2}.s, in order to treat the patient in a reasonable time; thus, is interesting to design filters for a radial beam of a TRIGA reactor, where materials such as Cd, Al, Fe and polyethylene are being implemented in the interest of having a spectrum with which the therapy can be implemented. For this design is being played with the position of the materials, to be able to see the behavior of the spectrum and thus choose some arrangement as indicated, of course taking into account the doses of both neutrons and gammas. (Author)

  1. Measured and Predicted Variations in Fast Neutron Spectrum in Massive Shields of Water and Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, E; Sandlin, R; Fraeki, R

    1965-09-15

    The absolute magnitude, and the variations in form, of the fast neutron spectrum during deep penetration (0.8 - 1.1 metre) in massive shields of water, ordinary and magnetite concrete have been studied by using threshold detectors (In (n, h'), S(n,p), Al(n, {alpha})). The results have been compared with predictions by two rigorous (NIOBE, Moments method) and two non-rigorous (multigroup removal-diffusion) shielding codes (NRN, RASH D). The absolute results predicted were in general within 50% of the measured ones, i. e. showed as good or better accuracy than thermal and epithermal flux predictions in the same small-reactor configurations. No difference in accuracy was found between the rigorous and non-rigorous methods. The changes in the relative form of the spectrum (indicated by variations in the (Al/S) and (In/S) reaction rate ratios and amounting to factors up to 3 - 4 during a one metre penetration in water) were rather accurately (within 10 - 30%) predicted by all of the methods. The photonuclear excitation of the 335 keV level used for detecting the In(n, n') reaction was found to distort completely the In results in water at penetrations > 50 cm.

  2. Computational uncertainties in silicon dioxide/plutonium intermediate neutron spectrum systems

    International Nuclear Information System (INIS)

    Jaegers, P.J.

    1997-01-01

    In the past several years, several proposals have been made for the long-term stabilization and storage of surplus fissile materials. Many of these proposed scenarios involve systems that have an intermediate neutron energy spectrum. Such intermediate-energy systems are dominated by scattering and fission events induced by neutrons ranging in energy from 1 eV to 100keV. To ensure adequate safety margins and cost effectiveness, it is necessary to have benchmark data for these intermediate-energy spectrum systems; however, a review of the nuclear criticality benchmarks indicates that no formal benchmarks are available. Nuclear data uncertainties have been reported for some types of intermediate-energy spectrum systems. Using a variety of Monte Carlo computer codes and cross-section sets, reported significant variations in the calculated k ∞ of intermediate-energy spectrum metal/ 235 U systems. We discuss the characteristics of intermediate neutron spectrum systems and some of the computational differences that can occur in calculating the k eff of these systems

  3. Pulsed White Spectrum Neutron Generator for Explosive Detection

    International Nuclear Information System (INIS)

    King, Michael J.; Miller, Gill T.; Reijonen, Jani; Ji, Qing; Andresen, Nord; Gicquel, Frederic; Kavlas, Taneli; Leung, Ka-Ngo; Kwan, Joe

    2008-01-01

    Successful explosive material detection in luggage and similar sized containers is a critical issue in securing the safety of all airline passengers. Tensor Technology Inc. has recently developed a methodology that will detect explosive compounds with pulsed fast neutron transmission spectroscopy. In this scheme, tritium beams will be used to generate neutrons with a broad energy spectrum as governed by the T(t,2n)4He fission reaction that produces 0-9 MeV neutrons. Lawrence Berkeley National Laboratory (LBNL), in collaboration with Tensor Technology Inc., has designed and fabricated a pulsed white-spectrum neutron source for this application. The specifications of the neutron source are demanding and stringent due to the requirements of high yield and fast pulsing neutron emission, and sealed tube, tritium operation. In a unique co-axial geometry, the ion source uses ten parallel rf induction antennas to externally couple power into a toroidal discharge chamber. There are 20 ion beam extraction slits and 3 concentric electrode rings to shape and accelerate the ion beam into a titanium cone target. Fast neutron pulses are created by using a set of parallel-plate deflectors switching between +-1500 volts and deflecting the ion beams across a narrow slit. The generator is expected to achieve 5 ns neutron pulses at tritium ion beam energies between 80-120 kV. First experiments demonstrated ion source operation and successful beam pulsing

  4. NEUTRON SPECTRUM MEASUREMENTS USING MULTIPLE THRESHOLD DETECTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gerken, William W.; Duffey, Dick

    1963-11-15

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

  5. 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

  6. 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)

  7. Measurements of the energy spectrum of backscattered fast neutrons

    International Nuclear Information System (INIS)

    Segal, Y.

    1976-03-01

    Experimental measurements have been made of the energy spectra of neutrons transmitted through slabs of iron, lead and perspex for incident neutron energies of 0.5, 1.0, 1.5 and 1.8 MeV. The neutron energy measurements were made using a He-3 spectrometer. The dependence of the neutrons energy spectrum as a function of scattering thickness was determined. The neutrons source used was a 3MeV Van de Graaff accelerator with a tritium target using the H 3 (p,n) He 3 reaction. The results obtained by the investigator on energy dependence of transmitted neutrons as a function of thickness of scattering material were compared, where possible, with the results obtained by other workers. The comparisons indicated good agreement. The experiment's results are compared with MORSE Monte Carlo calculated values. It is worthwhile to note that direct comparison between measured cross section values and the recommended ones are very far from satisfactory. In almost all cases the calculated spectrum is harder than the experimental one, a situation common to the penetrating and the back-scattered flux

  8. Neutron spectrum determination by activation method in fast neutron fields at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1994-01-01

    The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (author)

  9. Initial absolute calibration factors for some neutron sensitive self-powered detectors

    International Nuclear Information System (INIS)

    Kroon, J.

    1975-01-01

    Self-powered flux detectors have found extensive use as monitoring devices in PWR (Pressurized Water Reactor) cores and CANDU (Canada Deuterium Uranium) type power reactors. The detectors measure fuel power distributions and indicate trip parameters for reactor control and safety requirements. Both applications demand accurate absolute initial calibration factors. Experimental results obtained in calibrating some neutron sensitive self-powered detectors is presented. (author)

  10. Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Hee, E-mail: cyh871@snu.ac.kr [School of Energy Systems Engineering, Seoul National University, 599 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Joo, Han Gyu [School of Energy Systems Engineering, Seoul National University, 599 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

    2012-07-15

    The effect of neutron spectrum on primary damages in a structural material of a tritium breeding blanket is investigated with a newly established recoil spectrum estimation system. First, a recoil spectrum generation code is developed to obtain the energy spectrum of primary knock-on atoms (PKAs) for a given neutron spectrum utilizing the latest ENDF/B data. Secondly, a method for approximating the high energy tail of the recoil spectrum is introduced to avoid expensive molecular dynamics calculations for high energy PKAs using the concept of recoil energy of the secondary knock-on atoms originated by the INtegration of CAScades (INCAS) model. Thirdly, the modified spectrum is combined with a set of molecular dynamics calculation results to estimate the primary damage parameters such as the number of surviving point defects. Finally, the neutron spectrum is varied by changing the material of the spectral shifter and the result in primary damage parameters is examined.

  11. Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

    Science.gov (United States)

    Choi, Yong Hee; Joo, Han Gyu

    2012-07-01

    The effect of neutron spectrum on primary damages in a structural material of a tritium breeding blanket is investigated with a newly established recoil spectrum estimation system. First, a recoil spectrum generation code is developed to obtain the energy spectrum of primary knock-on atoms (PKAs) for a given neutron spectrum utilizing the latest ENDF/B data. Secondly, a method for approximating the high energy tail of the recoil spectrum is introduced to avoid expensive molecular dynamics calculations for high energy PKAs using the concept of recoil energy of the secondary knock-on atoms originated by the INtegration of CAScades (INCAS) model. Thirdly, the modified spectrum is combined with a set of molecular dynamics calculation results to estimate the primary damage parameters such as the number of surviving point defects. Finally, the neutron spectrum is varied by changing the material of the spectral shifter and the result in primary damage parameters is examined.

  12. A comparison in the reconstruction of neutron spectrums using classical iterative techniques

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Martinez B, M. R.; Vega C, H. R.; Gallego, E.

    2009-10-01

    One of the key drawbacks to the use of BUNKI code is that the process begins the reconstruction of the spectrum based on a priori knowledge as close as possible to the solution that is sought. The user has to specify the initial spectrum or do it through a subroutine called MAXIET to calculate a Maxwellian and a 1/E spectrum as initial spectrum. Because the application of iterative procedures by to resolve the reconstruction of neutron spectrum needs an initial spectrum, it is necessary to have new proposals for the election of the same. Based on the experience gained with a widely used method of reconstruction, called BUNKI, has developed a new computational tools for neutron spectrometry and dosimetry, which was first introduced, which operates by means of an iterative algorithm for the reconstruction of neutron spectra. The main feature of this tool is that unlike the existing iterative codes, the choice of the initial spectrum is performed automatically by the program, through a neutron spectra catalog. To develop the code, the algorithm was selected as the routine iterative SPUNIT be used in computing tool and response matrix UTA4 for 31 energy groups. (author)

  13. Neutron spectrum measurement by TOF

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1982-01-01

    The TOF experiments by using various facilities are described. The steady neutron spectra in light water which contains non-1/V absorbing materials were measured by the TOF method at a LINAC facility. The results were compared with the calculations based on the Koppel-Haywood model and two others. The leakage neutron spectra from a heavy-water assembly were measured and compared with model calculations. The time-dependent energy spectra in a small graphite assembly were measured. For this measurement, a chopper system was also used. The two-region calculation explains the spectrum just after the neutron burst. The time-dependent spectra in a small Be assembly and in an assembly of coolant-moderator containing hydrogen were also measured. The calculations based on various models are in progress. The TOF experiments at the reactor-chopper facility were carried out for measuring the total cross sections of crystalline moderators, the thermal neutron total cross section of high temperature beryllium, the thermal neutron total cross sections of granular lead and high temperature liquid lead, and the angle-dependent scattering spectra. A pseudo-chopper was designed and constructed. The spectra of the neutron field for medical use were measured by the chopper-TOF system. The thermal neutron total cross sections of Fe, Zr, Nb and Mg were measured, and the results were compared with the calculations by THRUSH and UNCLE-TOM codes. The random-trigger TOF experiments were made by using Cf-252. (Kato, T.)

  14. An evaluation of the spontaneous fission prompt neutron spectrum of 252Cf

    International Nuclear Information System (INIS)

    Bojkov, G.S.; Yurevich, V.I.

    1987-01-01

    An evaluation of the spontaneous fission prompt neutron spectrum of 252 Cf from 1 keV to 20 MeV is described. Variance-covariance matrices for a number of recent experimental data sets were constructed and used to evaluate the neutron spectrum following a Bayesian procedure. The evaluated spectrum is compared with various experimental and theoretical representations. (author)

  15. Comparison of Americium-Beryllium neutron spectrum obtained using activation foil detectors and NE-213 spectrometer

    International Nuclear Information System (INIS)

    Sunny, Sunil; Subbaiah, K.V.; Selvakumaran, T.S.

    1999-01-01

    Neutron spectrum of Americium - Beryllium (α,n) source is measured with two different spectrometers vis-a-vis activation foils (foil detectors) and NE-213 organic scintillator. Activity induced in the foils is measured with 4π-β-γ sodium iodide detector by integrating counts under photo peak and the saturation activity is found by correcting to elapsed time before counting. The data on calculated activity is fed into the unfolding code, SAND-II to obtain neutron spectrum. In the case of organic scintillator, the pulse height spectrum is obtained using MCA and this is processed with unfolding code DUST in order to get neutron spectrum. The Americium - Beryllium (α,n) neutron spectrum thus obtained by two different methods is compared. It is inferred that the NE-213 scintillator spectrum is in excellent agreement with the values beyond 1MeV. Neutron spectrum obtained by activation foils depends on initial guess spectrum and is found to be in reasonable agreement with NE-213 spectrum. (author)

  16. Frequency spectrum analysis of 252Cf neutron source based on LabVIEW

    International Nuclear Information System (INIS)

    Mi Deling; Li Pengcheng

    2011-01-01

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

  17. Neutron spectrum determination by activation method in fast neutron fields at the RB reactors

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.S.; Pesic, M.P.; Antic, D.P.

    1994-01-01

    The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (authors). 7 refs., 3 tabs

  18. Prompt neutron energy spectrum for the spontaneous fission of Cf-252

    International Nuclear Information System (INIS)

    Blinov, M.V.; Boykov, G.S.; Vitenko, V.A.

    1985-06-01

    The prompt neutron spectrum for the spontaneous fission of Cf-252 has been measured in 0.01-10 MeV region by the time-of-flight technique using a fast ionization chamber with U-235 layers as the neutron detector. Numerical data for the spectrum are presented, with an error file. (author)

  19. Systematic determination of the JET absolute neutron yield using the MPR spectrometer

    International Nuclear Information System (INIS)

    Kronborg-Pettersson, N.

    2003-04-01

    This thesis describes the first high-statistics systematic analysis of JET neutron yield and rate measurements obtained by using data acquired with the Magnetic Proton Recoil (MPR) neutron spectrometer. The neutron yield and rate were determined by using the count-rate from the MPR neutron spectrometer together with neutron profile information from other neutron diagnostic systems. This has previously been done manually for a few pulses. To be able to do this in a more systematic way a part of the neutron spectrum evaluation code was extracted and put into a separate custom-made program and modifications were done to extract sets of MPR data automatically. The codes have been used for analysis of a large set of pulses from the deuterium-tritium campaign at JET in 1997. Several results were obtained, the most significant of which was the clear improvement seen when neutron profile corrections were applied. Neutron yield-rates derived from MPR count-rate are shown to be in excellent agreement with other JET neutron diagnostic data

  20. Systematic determination of the JET absolute neutron yield using the MPR spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Kronborg-Pettersson, N

    2003-04-01

    This thesis describes the first high-statistics systematic analysis of JET neutron yield and rate measurements obtained by using data acquired with the Magnetic Proton Recoil (MPR) neutron spectrometer. The neutron yield and rate were determined by using the count-rate from the MPR neutron spectrometer together with neutron profile information from other neutron diagnostic systems. This has previously been done manually for a few pulses. To be able to do this in a more systematic way a part of the neutron spectrum evaluation code was extracted and put into a separate custom-made program and modifications were done to extract sets of MPR data automatically. The codes have been used for analysis of a large set of pulses from the deuterium-tritium campaign at JET in 1997. Several results were obtained, the most significant of which was the clear improvement seen when neutron profile corrections were applied. Neutron yield-rates derived from MPR count-rate are shown to be in excellent agreement with other JET neutron diagnostic data.

  1. Incident spectrum determination for time-of-flight neutron powder diffraction data analysis

    International Nuclear Information System (INIS)

    Hodges, J. P.

    1998-01-01

    Accurate characterization of the incident neutron spectrum is an important requirement for precise Rietveld analysis of time-of-flight powder neutron diffraction data. Without an accurate incident spectrum the calculated model for the measured relative intensities of individual Bragg reflections will possess systematic errors. We describe a method for obtaining an accurate numerical incident spectrum using data from a transmitted beam monitor

  2. Corrections on energy spectrum and scattering for fast neutron radiography at NECTAR facility

    International Nuclear Information System (INIS)

    Liu Shuquan; Thomas, Boucherl; Li Hang; Zou Yubin; Lu Yuanrong; Guo Zhiyu

    2013-01-01

    Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM-Ⅱ in Technische Universitaet Mounchen (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections. (authors)

  3. Corrections on energy spectrum and scatterings for fast neutron radiography at NECTAR facility

    Science.gov (United States)

    Liu, Shu-Quan; Bücherl, Thomas; Li, Hang; Zou, Yu-Bin; Lu, Yuan-Rong; Guo, Zhi-Yu

    2013-11-01

    Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM- II in Technische Universität München (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections.

  4. Absolute measurement and international intercomparison of 0.1-0.8 MeV monoenergetic neutron fluence rate

    International Nuclear Information System (INIS)

    Ma Hongchang; Lu Hanlin; Rong Chaofan

    1988-01-01

    The methods for absolute measurement of 0.1-18MeV monoenergetic neutron fluence rate are described. Which include proton recoil telescope, semicoducetor telescope, hydrogen filled proportional counter and associated particale method. A long counter used as secondary recent international intercomparison of neutron fluence rate organized by BIPM, and the results were given

  5. Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, Makoto, E-mail: makoto.fukuda@qse.tohoku.ac.jp [Tohoku University, Sendai, 980-8579 (Japan); Kiran Kumar, N.A.P.; Koyanagi, Takaaki; Garrison, Lauren M. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA, 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Hasegawa, Akira [Tohoku University, Sendai, 980-8579 (Japan)

    2016-10-15

    Neutron irradiation to single crystal pure tungsten was performed in the mixed spectrum High Flux Isotope Reactor (HFIR). To investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ∼90–∼800 °C and fast neutron fluences were 0.02–9.00 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. The hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten. - Highlights: • The microstructure and irradiation hardening of single crystal pure W irradiated in HFIR was investigated. • The neutron energy spectrum influence was evaluated by comparing the HFIR results with previous work in Joyo and JMTR. • In the dose range up to ∼1 dpa, the neutron energy spectrum influence of irradiation hardening was not clear. • In the dose range above 1 dpa, the neutron energy influence on irradiation hardening and microstructural development was clearly observed. • The irradiation induced precipitates caused significant irradiation hardening of pure W irradiated in HFIR.

  6. Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device.

    Science.gov (United States)

    Xu, Yanping; Randers-Pehrson, Gerhard; Marino, Stephen A; Garty, Guy; Harken, Andrew; Brenner, David J

    2015-09-11

    A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n) 3 He and D(d,n) 3 He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9 Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.

  7. Updated neutron spectrum characterization of SNL baseline reactor environments

    International Nuclear Information System (INIS)

    Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

    1994-04-01

    This document provides SAND-II and MANIPULATE output listings from calculations used to derive the new spectrum-averaged integral parameters that were reported in volume 1. When used in conjunction with volume 1, this document provides an audit trail for the neutron radiation field characterization and supports current quality assurance initiatives. This document provides detailed information on the neutron spectrum characteristics of the primary Sandia National Laboratories' (SNL) reactor environments. The information in this volume is not intended for the casual user of the SNL reactor facilities. This detailed characterization of the neutron and gamma environments at the Sandia Pulsed Reactor (SPR) and the Annular Core Research Reactor (ACRR) is provided to aid the users who wish to convert the information given in the Radiation Metrology Laboratory (RML) dosimetry reports into other (non-silicon) measures of neutron damage. The spectra provided in these appendices can be used as a source term for Monte Carlo radiation transport calculations to study the impact of experimenter's test package on the neutron environment

  8. Spectrum of neutrons leaking from an iron sphere with a central 14 MeV neutron source

    International Nuclear Information System (INIS)

    Borisov, A.A.; Zagryadskij, V.A.; Chuvilin, D.Yu.; Kralik, M.; Pulpan, J.; Tichy, M.

    1991-01-01

    Following a review of the present state of nuclear data requisite for the calculation of the transport of 14 MeV neutrons through iron of natural isotopic composition, the results are given of the calculation of the energy spectrum of such neutrons after their passage through an iron sphere 240 mm o.d. and 90 mm i.d., the neutron source being accommodated in the centre of the sphere. The calculations were made using the one-dimensional code BLANK working with the nuclear data libraries ENDL-75, ENDL-83, ENDL/B-IV, JENDL-2 and BROND, and using the three-dimensional code BRAND with the library ENDL-78. The calculated spectra were compared with the experimental spectrum measured at a distance of 3 m from the sphere by means of an NE-213 scintillator, which records reflected protons. The reflected proton spectrum was processed by the matrix method (program FORIST), and the result was normalized to one neutron emitted by the source, as were the calculated spectra. The comparison demonstrates that the experiment is best fitted by the spectrum calculated by using the library JENDL-2, where the integrals of the observed and calculated spectra over the 1-15 MeV range differ as little as approximately 10%. (author). 3 figs., 5 tabs., 16 refs

  9. Approximation for the adjoint neutron spectrum

    International Nuclear Information System (INIS)

    Suster, Luis Carlos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2002-01-01

    The proposal of this work is the determination of an analytical approximation which is capable to reproduce the adjoint neutron flux for the energy range of the narrow resonances (NR). In a previous work we developed a method for the calculation of the adjoint spectrum which was calculated from the adjoint neutron balance equations, that were obtained by the collision probabilities method, this method involved a considerable quantity of numerical calculation. In the analytical method some approximations were done, like the multiplication of the escape probability in the fuel by the adjoint flux in the moderator, and after these approximations, taking into account the case of the narrow resonances, were substituted in the adjoint neutron balance equation for the fuel, resulting in an analytical approximation for the adjoint flux. The results obtained in this work were compared to the results generated with the reference method, which demonstrated a good and precise results for the adjoint neutron flux for the narrow resonances. (author)

  10. Simultaneous neutron and gamma spectrum adjustment

    International Nuclear Information System (INIS)

    Remec, I.

    1996-01-01

    The spectrum adjustment procedure was extended to simultaneous neutron and gamma spectrum adjustment, and the feasibility of this technique is demonstrated in the analysis of HFIR dosimetry experiments. Conditions in which gamma rays may contribute considerably to radiation damage in steels are discussed. Beryllium helium accumulation fluence monitors (HAFMs) were found to be good monitors in gamma fields of intensities high enough to contribute to steel embrittlement. Use of 237 Np, 238 U, and 9 Be HAFM as gamma dosimeters is proposed for high-dose irradiations in high-energy, high-intensity gamma fields

  11. A neutron spectrum unfolding computer code based on artificial neural networks

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  12. Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

    International Nuclear Information System (INIS)

    Shahbunder, Hesham; Pyeon, Cheol Ho; Misawa, Tsuyoshi; Lim, Jae-Yong; Shiroya, Seiji

    2010-01-01

    The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k s , external source efficiency φ*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton-tungsten source in hard and soft neutron spectra cores and 14 MeV D-T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6-13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.

  13. Little Boy neutron spectrum below 3 MeV

    International Nuclear Information System (INIS)

    Evans, A.E.; Bennett, E.F.; Yule, T.J.

    1984-01-01

    The leakage neutron spectrum from the Little Boy replica has been measured from 12 keV to 3 MeV using a high-resolution 3 He ionization chamber, and from 1 keV to 3 MeV using proton-recoil proportional counters. The 3 He-spectrometer measurements were made at distances of 0.75 and 2.0 m from the active center and at angles of 0 0 , 45 0 , and 90 0 with respect to the axis of the assembly. Proton-recoil measurments were made at 90 0 to the assembly axis at distances of 0.75 and 2.0 m, with a shielded measurement made at 2.0 m to estimate background due to scattering. The 3 He spectrometer was calibrated at Los Alamos using monoenergetic 7 Li(p,n) 7 Be neutrons to generate a family of response functions. The proton-recoil counters were calibrated at Argonne by studying the capture of thermal neutrons by nitrogen in the counters, by observation of the 24-keV neutron resonance in iron, and by relating to the known hydrogen content of the counters. The neutron spectrum from Little Boy was found to be highly structured, with peaks corresponding to minima in the iron total neutron cross section. In particular, influence of the 24-keV iron window was evident in both sets of spectra. The measurements provide information for dosimetry calculations and also a valuable intercomparison of neutron spectrometry using the two different detector types. Spectra measured with both detectors are in essential agreement. 8 references, 7 figures, 2 tables

  14. A novel neutron energy spectrum unfolding code using particle swarm optimization

    International Nuclear Information System (INIS)

    Shahabinejad, H.; Sohrabpour, M.

    2017-01-01

    A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to match well with those of the TGASU code for both under determined and over-determined problems. In addition the SDPSO has been shown to be nearly two times faster than the TGASU code. - Highlights: • Introducing a novel method for neutron spectrum unfolding. • Implementation of a particle swarm optimization code for neutron unfolding. • Comparing results of the PSO code with those of recently published TGASU code. • Match results of the PSO code with those of TGASU code. • Greater convergence rate of implemented PSO code than TGASU code.

  15. Measurement of Neutron Energy Spectrum Emitted by Cf-252 Source Using Time-of-Flight Method

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Cheol Ho; Son, Jaebum; Kim, Tae Hoon; Lee, Sangmin; Kim, Yong-Kyun [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    The techniques proposed to detect the neutrons usually require the detection of a secondary recoiling nucleus in a scintillator (or other type of detector) to indicate the rare collision of a neutron with a nucleus. This is the same basic technique, in this case detection of a recoil proton that was used by Chadwick in the 1930 s to discover and identify the neutron and determine its mass. It is primary technique still used today for detection of fast neutron, which typically involves the use of a hydrogen based organic plastic or liquid scintillator coupled to a photo-multiplier tube. The light output from such scintillators is a function of the cross section and nuclear kinematics of the n + nucleus collision. With the exception of deuterated scintillators, the scintillator signal does not necessarily produce a distinct peak in the scintillator spectrum directly related to the incident neutron energy. Instead neutron time-of-flight (TOF) often must be utilized to determine the neutron energy, which requires generation of a prompt start signal from the nuclear source emitting the neutrons. This method takes advantage of the high number of prompt gamma rays. The Time-of-Flight method was used to measure neutron energy spectrum emitted by the Cf-252 neutron source. Plastic scintillator that has a superior discrimination ability of neutron and gamma-ray was used as a stop signal detector and liquid scintillator was used as a stat signal detector. In experiment, neutron and gamma-ray spectrum was firstly measured and discriminated using the TOF method. Secondly, neutron energy spectrum was obtained through spectrum analysis. Equation of neutron energy spectrum that was emitted by Cf-252 source using the Gaussian fitting was obtained.

  16. Refinements in the Los Alamos model of the prompt fission neutron spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Madland, D.G., E-mail: dgm@lanl.gov; Kahler, A.C.

    2017-01-15

    This paper presents a number of refinements to the original Los Alamos model of the prompt fission neutron spectrum and average prompt neutron multiplicity as derived in 1982. The four refinements are due to new measurements of the spectrum and related fission observables many of which were not available in 1982. They are also due to a number of detailed studies and comparisons of the model with previous and present experimental results including not only the differential spectrum, but also integral cross sections measured in the field of the differential spectrum. The four refinements are (a) separate neutron contributions in binary fission, (b) departure from statistical equilibrium at scission, (c) fission-fragment nuclear level-density models, and (d) center-of-mass anisotropy. With these refinements, for the first time, good agreement has been obtained for both differential and integral measurements using the same Los Alamos model spectrum.

  17. Continuous energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the neutron energy spectrum by target nuclei tagging

    Energy Technology Data Exchange (ETDEWEB)

    Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Grupo de Estudos Nucleares; Leite, Sergio Q. Bogado, E-mail: sbogado@ibest.com.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In this work a Monte Carlo simulator with continuous energy is used. This simulator distinguishes itself by using the sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum (for high energy neutrons) and the Maxwell-Boltzmann distribution (for thermal neutrons). The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. It is common practice in neutron transport calculations, e.g. multi-group transport, to consider that the neutrons only lose energy with each scattering reaction and then to use a thermal group with a Maxwellian distribution. Such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies, i.e. in the thermal energy region, in which it can be regarded as a Maxwell-Boltzmann distribution for thermal equilibrium. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution. It is then shown how this procedure can emulate the up-scattering effect by the increase in the neutron population kinetic energy. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process. This work contains some preliminary results obtained from a Monte Carlo simulator for neutron transport that is being developed at Federal University of Rio Grande do Sul. (author)

  18. Determination of the absolute configuration of (+)-neopentyl-1-d alcohol by neutron and x-ray diffraction analysis.

    Science.gov (United States)

    Yuan, H S; Stevens, R C; Bau, R; Mosher, H S; Koetzle, T F

    1994-12-20

    The absolute configuration of (+)-neopentyl-1-d alcohol, prepared by the reduction of 2,2-dimethylpropanal-1-d by actively fermenting yeast, has been determined to be S by neutron diffraction. The neutron study was carried out on the phthalate half ester of neopentyl-1-d alcohol, crystallized as its strychnine salt. The absolute configuration of the (-)-strychninium cation was first determined by an x-ray anomalous dispersion study of its iodide salt. The chiral skeleton of strychnine then served as a reference from which the absolute configuration of the -O-CHD-C(CH3)3 group of neopentyl phthalate was determined. Difference Fourier maps calculated from the neutron data showed unambiguously that the -O-CHD-C(CH3)3 groups of both independent molecules in the unit cell had the S configuration. This work proves conclusively that the yeast system reduces aldehydes by delivering hydrogen to the re face of the carbonyl group. Crystallographic details: (-)-strychninium (+)-neopentyl-1-d phthalate, space group P2(1) (monoclinic), a = 18.564(6) A, b = 7.713(2) A, c = 23.361(8) A, beta = 94.18(4) degrees, V = 3336.0(5) A3, Z = 2 (T = 100 K). Final agreement factors are R(F) = 0.073 for 2768 reflections collected at room temperature (x-ray analysis) and R(F) = 0.144 for 960 reflections collected at 100 K (neutron analysis).

  19. Spectrum shaping of accelerator-based neutron beams for BNCT

    CERN Document Server

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

    2002-01-01

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

  20. Cadmium depletion impacts on hardening neutron spectrum for advanced fuel testing in ATR

    International Nuclear Information System (INIS)

    Chang, Gray S.

    2011-01-01

    For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim-Effect in the test region. (author)

  1. Comparison of neutron spectrum unfolding codes

    International Nuclear Information System (INIS)

    Zijp, W.

    1979-02-01

    This final report contains a set of four ECN-reports. The first is dealing with the comparison of the neutron spectrum unfolding codes CRYSTAL BALL, RFSP-JUL, SAND II and STAY'SL. The other three present the results of calculations about the influence of statistical weights in CRYSTAL BALL, SAND II and RFSP-JUL

  2. Experimental measurement of neutron spectrum in the reflector of a light water reactor

    International Nuclear Information System (INIS)

    Brethe, P.

    1963-09-01

    1. Thermal neutrons: The temperature of the thermal neutron spectrum was calculated using Au-Lu foils. This temperature varies from 300 deg. K (temperature of the moderator) at 30 cm of the core to 350 deg. K in a hole of the core. 2. Slowing down of neutron: Four resonance detectors have been used (Au, In, Co, Mn). We can write a 1/E form of the spectrum. The linking up energy E M between thermal neutron spectrum and slowing down spectrum is about 0.23 eV and is free from the Maxwell spectrum temperature. The decrease of slowing down flux regarding thermal flux, farther from the core, has been showed. 3. Fast neutrons: We used 3 threshold detectors (Ni, Al, Mg). We supposed a E 1/2 e -βE from of the spectrum above 3 MeV. The values of β are in a range from 0.775, at the centre of the core and in a loop-hole, to 0,64 at about 30 cm of the core. 4. Continuous shape of the spectrum: The following interpolations give useful informations between the field where measurements have been made: between 340 eV and 10 keV: 1/E form between 10 keV and 330 keV: 1/(E σ S (E)) form (σ S (E) elastic scattering section on hydrogen) between 330 keV and 3 MeV: calculated form by the moments method (ref. BSR). (author) [fr

  3. The Neutron Spectrum in a Uranium Tube

    International Nuclear Information System (INIS)

    Johansson, E.; Jonsson, E.; Lindberg, M.; Mednis, J.

    1963-10-01

    A series of experimental and theoretical investigations on neutron spectra in lattice cells has been started at the reactor R1. This report gives the results from the first one of these cells - one with a tube of natural -uranium surrounded by heavy water. In the measurements the cell was placed in the central, vertical channel of the reactor. The neutron spectrum from a lead scatterer in the uranium tube - outer diameter 49.2 mm, inner diameter 28.3 mm - was measured with a fast chopper in the energy region 0.01 to 100 eV. Subsidiary measurements indicated that the spectrum in the beam from the lead piece corresponds to the spectrum of the angular flux integrated over all angles. This correspondence is important for the interpretation of the experimental data. The thermal part of the spectrum was found to deviate significantly from a Maxwellian. However, the deviation is not very large, and one could use a Maxwellian, at least to give a rough idea of the hardness of the spectrum. For the present tube the temperature of this Maxwellian was estimated as 90 to 100 deg C above the moderator temperature (33 deg C). In the joining region the rise of the spectrum towards the thermal part is slower than for the cell boundary spectrum, measured earlier. In the epithermal region the limited resolution of the chopper has affected the measurements at the energies of the uranium resonances. However, the shape of the spectrum on the flanks of the first resonance in 238 U (6.68 eV) has been obtained accurately. In the theoretical treatment the THERMOS code with a free gas scattering model has been used. The energy region was 3.06 - 0.00025 eV. The agreement with the measurements is good for the thermal part - possibly the theoretical spectrum is a little softer than the experimental one. In the joining region the results from THERMOS are comparatively high - probably due to the scattering model used

  4. The Neutron Spectrum in a Uranium Tube

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Jonsson, E; Lindberg, M; Mednis, J

    1963-10-15

    A series of experimental and theoretical investigations on neutron spectra in lattice cells has been started at the reactor R1. This report gives the results from the first one of these cells - one with a tube of natural -uranium surrounded by heavy water. In the measurements the cell was placed in the central, vertical channel of the reactor. The neutron spectrum from a lead scatterer in the uranium tube - outer diameter 49.2 mm, inner diameter 28.3 mm - was measured with a fast chopper in the energy region 0.01 to 100 eV. Subsidiary measurements indicated that the spectrum in the beam from the lead piece corresponds to the spectrum of the angular flux integrated over all angles. This correspondence is important for the interpretation of the experimental data. The thermal part of the spectrum was found to deviate significantly from a Maxwellian. However, the deviation is not very large, and one could use a Maxwellian, at least to give a rough idea of the hardness of the spectrum. For the present tube the temperature of this Maxwellian was estimated as 90 to 100 deg C above the moderator temperature (33 deg C). In the joining region the rise of the spectrum towards the thermal part is slower than for the cell boundary spectrum, measured earlier. In the epithermal region the limited resolution of the chopper has affected the measurements at the energies of the uranium resonances. However, the shape of the spectrum on the flanks of the first resonance in {sup 238}U (6.68 eV) has been obtained accurately. In the theoretical treatment the THERMOS code with a free gas scattering model has been used. The energy region was 3.06 - 0.00025 eV. The agreement with the measurements is good for the thermal part - possibly the theoretical spectrum is a little softer than the experimental one. In the joining region the results from THERMOS are comparatively high - probably due to the scattering model used.

  5. The measurement of prompt neutron spectrum in spontaneous fission of {sup 244}Cm

    Energy Technology Data Exchange (ETDEWEB)

    Batenkov, O.I.; Boykov, G.S.; Drapchinsky, L.V.; Majorov, M.Ju.; Trenkin, V.A. [V.G. Khlopin Radium Inst., Saint Petersburg (Russian Federation)

    1997-03-01

    Under the Program of Measurements of Prompt Fission Neutron Spectra of Minor Actinides for Transmutation Purposes the integral neutron spectrum in spontaneous fission of {sup 244}Cm has been measured by the time-of-flight method in the energy range of 0.1-15 MeV relative to the standard neutron spectrum in {sup 252}Cf spontaneous fission. Essential attention was paid to revealing of possible systematic errors. It is shown, that the {sup 244}Cm spectrum shape may be well described by using Mannhart evaluation with appropriate parameter of Maxwell temperature T{sub M} = 1.37 MeV. (author)

  6. The fast neutron emission spectrum of 252-Cf

    International Nuclear Information System (INIS)

    Bensch, F.

    1979-07-01

    The aim of this work was a new measurement of the neutron emission spectrum of 252-Cf neutron standard sources as the IAEA is offering to users. The main feature was the application of gas-filled proton-recoil spectrometers and no TOF technique. The special interest of this document was in the temperature parameter of the Maxwellian distribution and in its relative deviations. In this connection, special measurements with high energy resolution were carried out in a search for fine structure neutron groups, which have been observed in some TOF measurements, but could not be reproduced during this measurement

  7. Neutron spectrum perturbations due to scattering materials and their effect on the average neutron energy, the spectral index, and the hardness parameter

    International Nuclear Information System (INIS)

    Wright, H.L.; Meason, J.L.; Wolf, M.; Harvey, J.T.

    1976-01-01

    Measurements have been performed on the perturbing effect of a number of scattering materials by the 'free-field' neutron leakage spectrum from a Godiva Type Critical Assembly (White Sands Missile Range Fast Burst Reactor). The results of these measurements are interpreted in relation to some of the general parameters characterizing a neutron environment, namely, the average neutron energy >10 KeV, the spectral index and the hardness parameter. Three neutron spectrum measurements have been performed, each under different experimental configurations of scattering materials. Results from these measurements show the following with relation to the spectral index: (1) The neutron environment on the core surface and at 12-inches from the core surface (free-field) yield a spectral index of 6.8, (2) The neutron environment behind a 4.75-inch Plexiglas plate yield 4.6 for the spectral index and (3) The neutron environment behind a 2-inch aluminum plate yield 6.7 for the spectral index. It is concluded that the core surface and the 12-inch from core surface neutron environment are identical with the 'free-field' neutron environment at 20-inches when considering only those neutrons with energy >10 KeV. On the other hand, it appears that the 4.75 inches of Plexiglas severely perturbs the 'free-field' neutron environment, i.e., a much harder neutron spectrum >10 KeV. In the situation where 2-inches of aluminum is used as the perturbing medium, essentially no change in the neutron spectrum >10 KeV is noted

  8. The Real-time Frequency Spectrum Analysis of Neutron Pulse Signal Series

    International Nuclear Information System (INIS)

    Tang Yuelin; Ren Yong; Wei Biao; Feng Peng; Mi Deling; Pan Yingjun; Li Jiansheng; Ye Cenming

    2009-01-01

    The frequency spectrum analysis of neutron pulse signal is a very important method in nuclear stochastic signal processing Focused on the special '0' and '1' of neutron pulse signal series, this paper proposes new rotation-table and realizes a real-time frequency spectrum algorithm under 1G Hz sample rate based on PC with add, address and SSE. The numerical experimental results show that under the count rate of 3X10 6 s -1 , this algorithm is superior to FFTW in time-consumption and can meet the real-time requirement of frequency spectrum analysis. (authors)

  9. Neutron metrology file NMF-90. An integrated database for performing neutron spectrum adjustment calculations

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1996-01-01

    The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariances files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs

  10. Determination of the absolute configuration of (+)-neopentyl-1-d alcohol by neutron and x-ray diffraction analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, H.S.H.; Stevens, R.C.; Bau, R. (Univ. of Southern California, Los Angeles, CA (United States)); Mosher, H.S. (Stanford Univ., CA (United States)); Koetzle, T.F. (Brookhaven National Lab., Upton, NY (United States))

    1994-12-20

    The absolute configuration of (+)-neopentyl-1-d alcohol, prepared by the reduction of 2,2-dimethylpropanol-1-d by actively fermenting yeast, has been determined to be S by neutron diffraction. The neutron study was carried out on the phthalate half ester of neopentyl-1-d alcohol, crystallized as its strychnine salt. The absolute configuration of the (-)-strychninium cation was first determined by an x-ray anomalous dispersion study of its iodide salt. The chiral skeleton of strychnine then served as a reference from which the absolute configuration of the -O-CHD-C(CH[sub 3])[sub 3] group of neopentyl phthalate was determined. Difference Fourier maps calculated from the neutron data showed unambiguously that the -O-CHD-C(CH[sub 3])[sub 3] groups of both independent molecules in the unit cell had the S configuration. This work proves conclusively that the yeast system reduces aldehydes by delivering hydrogen to the re face of the carbonyl group. Crystallographic details: (-)-strychninium (+)-neopentyl-1-d phthalate, space group P2[sub 1] (monoclinic), a = 18.564(6) [angstrom], b = 7.713(2) [angstrom], c = 23.361(8) [angstrom], [beta] = 94.18(4)[degrees], V = 3336.0(5) [angstrom][sup 3], Z = 2 (T = 100 K). Final agreement factors are R(F) = 0.073 for 2768 reflections collected at room temperature (x-ray analysis) and R(F) = 0.144 for 960 reflections collected at 100 K (neutron analysis). 49 refs., 7 figs., 2 tabs.

  11. Thermal neutron spectrum distribution in TRIGA fuels

    International Nuclear Information System (INIS)

    Gui Ah Auu; Harasawa, Susumu; An, Shigehiro

    1989-01-01

    The dependence of thermal neutron spectrum in TRIGA fuel cell on fuel temperature and TRIGA fuel types were studied using LIBP and THERMOS codes. Some characteristics of the TRIGA fuel including its prompt negative temperature coefficient of reactivity were explained using the results of the study. (author)

  12. Absolute measurement of thermal neutron fluence and its application for fission track dating

    International Nuclear Information System (INIS)

    Ganzawa, Yoshihiro; Honda, Teruyuki; Nozaki, Tetsuya.

    1988-01-01

    The absolute measurements of thermal neutron fluence for fission track dating have been developed after the proceeding results of Honda et al. (1987). The 2,200 m/sec activation cross section of 197 Au (98.8 barn) is corrected to 87.4 barn (σa) by the three factors of the neutron temperature, Maxwellian distribution of thermal neutrons and non 1/v correction factor for the above absolute measurement. The calibrated factor (B th ) of standard glasses (SRM613, SRM962a, CN-1 and CN-2) and zeta-a (ζa) values for fission track dating are determined on the basis of these experimental results. The values of B th , (7.47 ± 0.29) x 10 9 for SRM613, (7.43 ± 0.34) x 10 9 for SRM962a, (2.50 ± 0.06) x 10 9 for CN-1 and (2.74 ± 0.06) x 10 9 for CN-2 closely agree with those reported previously by Honda et al. (1987). Further, the ζa values of 392.3 ± 16.5 for SRM962a and SRM613, 131.4 ± 3.1 for CN-1 and 144.1 ± 3.3 for CN-2 calculated from B th , effective thermal neutron fission cross-section σf (497.4 barn), isotopic abundance ratio 235 U/ 239 U, I (7.2527 x 10 -3 ) and spontaneous fission decay constant of 238 U, λ f (6.85 x 10 -17 a -7 ) show close agreement with ζ b values (392.5 ± 10.0, 131.6 ± 3.3, 140.1 ± 3.5) derived from the absolute age of Fish Canyon Tuff (27.9 ± 0.7 Ma) respectively. The fission track dating of zircons separated from Oligocene-Miocene tuff distributed in Eastern Hokkaido have been carried out by the external detector method using ζ a . The obtained ages are 28.6 ± 0.7 Ma (1 - 2) and 23.3 ± 0.7 Ma (3 - 2). These results agree well with the geologic age supported from Ashoro Fossil Fauna, K-Ar ages of volcanic rocks and stratigraphy in this area. (author)

  13. Neutron energy spectrum flux profile of Ghana's miniature neutron source reactor core

    International Nuclear Information System (INIS)

    Sogbadji, R.B.M.; Abrefah, R.G.; Ampomah-Amoako, E.; Agbemava, S.E.; Nyarko, B.J.B.

    2011-01-01

    Highlights: → The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was studied. → Using 20,484 energy grids, the thermal, slowing down and fast neutron energy regions were studied. - Abstract: The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) x 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) x 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) x 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939-3.7880) x 10 -08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) x 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 x 10 -01 MeV and 8.2680 x 10 -01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) x 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

  14. DIFFERENTIAL SPECTRUM OF NEUTRONS AT CHACALTAYA-BOLIVIA

    International Nuclear Information System (INIS)

    Mayta, R.; Zanini, A.; Ticona, R.; Velarde, A.

    2009-01-01

    We describe the Neutron Spectrometer Experiment installed at Chacaltaya Cosmic Rays Observatory (68 deg. O, 16.2 deg. S), located in Bolivia, at 5230 m.a.s.l. This experimental system is constituted by passive detectors which register the flux of neutrons, in an energy range of 10 KeV-20 MeV. Using the unfolding code BUNTO a peak around 1 MeV of the characteristic spectrum of neutrons was obtained. Experimental values, observed during April of 2008, are compared with similar ones carried out in 1997 at the same place, in order to look for eventual changes due to local atmosphere. A similar experiment was also carried up at the Laboratory of Testa Grigia-Italy (45.56 deg. N, 7.42 deg. E,. 3480 m.a.l.s). Data of both stations allow us to compare the spectra in order to explain the difference of neutron flux of these two stations.

  15. Thick-foils activation technique for neutron spectrum unfolding with the MINUIT routine-Comparison with GEANT4 simulations

    Science.gov (United States)

    Vagena, E.; Theodorou, K.; Stoulos, S.

    2018-04-01

    Neutron activation technique has been applied using a proposed set of twelve thick metal foils (Au, As, Cd, In, Ir, Er, Mn, Ni, Se, Sm, W, Zn) for off-site measurements to obtain the neutron spectrum over a wide energy range (from thermal up to a few MeV) in intense neutron-gamma mixed fields such as around medical Linacs. The unfolding procedure takes into account the activation rates measured using thirteen (n , γ) and two (n , p) reactions without imposing a guess solution-spectrum. The MINUIT minimization routine unfolds a neutron spectrum that is dominated by fast neutrons (70%) peaking at 0.3 MeV, while the thermal peak corresponds to the 15% of the total neutron fluence equal to the epithermal-resonances area. The comparison of the unfolded neutron spectrum against the simulated one with the GEANT4 Monte-Carlo code shows a reasonable agreement within the measurement uncertainties. Therefore, the proposed set of activation thick-foils could be a useful tool in order to determine low flux neutrons spectrum in intense mixed field.

  16. Study of U235 neutron fission spectrum by the knowledge of cross sections average over that spectrum

    International Nuclear Information System (INIS)

    Suarez, P.M.

    1997-01-01

    A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the 45 Sc(n,2n) 44g Sc and 45 Sc(n,2n) 44m Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, 235 U. The results of the calculations mentioned above lead us to propose an analytical form for the 235 U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole energy range. These two new

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  18. Fast neutron spectrum measurement in the JMTR

    International Nuclear Information System (INIS)

    Sakurai, K.; Mizuho, M.

    1980-01-01

    Fast neutron spectrum measurement at positions of K-10 (fuel region), J-11 (first beryllium reflector region) and I-12 (second beryllium reflector region) in the JMTRC has been performed with the threshold detectors such as 107 Ag(n,n')sup(107m)Ag, 103 Rh(n,n')sup(103m)Rh, 115 In(n,n')sup(115m)In and 238 U(n,f)F.P. above 0.1 MeV energy region. The activity data have been analyzed by the unfolding code SAND-II using ANISN spectrum for initial spectrum. An effective cross section of 54 Fe(n,p) 54 Mn is calculated with Fabry's cross section data and SAND-II spectrum for K-10, J-11 and I-12. They are 76.7 mb, 69.0 mb and 68.2 mb for K-10, J-11 and I-12 respectively. These values agree with the effective cross sections (calculated by Fabry's cross section data and ANISN spectrum) within +-6%

  19. A compact proton spectrometer for measurement of the absolute DD proton spectrum from which yield and ρR are determined in thin-shell inertial-confinement-fusion implosions.

    Science.gov (United States)

    Rosenberg, M J; Zylstra, A B; Frenje, J A; Rinderknecht, H G; Johnson, M Gatu; Waugh, C J; Séguin, F H; Sio, H; Sinenian, N; Li, C K; Petrasso, R D; Glebov, V Yu; Hohenberger, M; Stoeckl, C; Sangster, T C; Yeamans, C B; LePape, S; Mackinnon, A J; Bionta, R M; Talison, B; Casey, D T; Landen, O L; Moran, M J; Zacharias, R A; Kilkenny, J D; Nikroo, A

    2014-10-01

    A compact, step range filter proton spectrometer has been developed for the measurement of the absolute DD proton spectrum, from which yield and areal density (ρR) are inferred for deuterium-filled thin-shell inertial confinement fusion implosions. This spectrometer, which is based on tantalum step-range filters, is sensitive to protons in the energy range 1-9 MeV and can be used to measure proton spectra at mean energies of ∼1-3 MeV. It has been developed and implemented using a linear accelerator and applied to experiments at the OMEGA laser facility and the National Ignition Facility (NIF). Modeling of the proton slowing in the filters is necessary to construct the spectrum, and the yield and energy uncertainties are ±DD-neutron yield diagnostics at the NIF.

  20. Research on determine the absolute neutron output of distributed pulse generators

    International Nuclear Information System (INIS)

    Li Bojun; Tang Zhangkui; Wang Dong; Yang Gaozhao; Peng Taiping

    2009-01-01

    In order to determine the absolute neutron output of distributed pulse generators, we deduced equivalent length to deal with experimental data, according to the different layout and weighting of multiple pulse generators. The deposited energy in scintillation crystal and the integral flux which drilling through crystal interface was simulated by MCNP code. The result shows the simulated proportion of different distributed pulse generators is approximately agreed with experimental data. The validity of the equivalent length model was proved by the consistent results between calculation and experimental data. (authors)

  1. Analysis of the fast-neutron spectrum inside the experimental cavity of the NRU Mk4 FN rod

    International Nuclear Information System (INIS)

    Leung, T.C.

    1995-01-01

    The fast-neutron (FN) rods in the NRU reactor provide a facility to study the effects of irradiation on CANDU reactor materials. The Mark 4 (Mk4) FN rods use natural uranium and supply fast-neutrons for experiments on irradiation creep and growth, and corrosion, for pressure- and calandria-tube materials. The neutron fluxes above 1 MeV are up to 2.7x10 17 n.m -2 .s -1 . This paper describes a calculation of the fast-neutron spectrum inside the NRU Mk4 FN rod cavity. The calculation was performed using the WIMS-AECL code, which is a multi-group transport code with two dimensional capabilities using the collision-probability method. Results for the fast-neutron spectrum above 1 MeV are presented in nine groups. The analysis confirms that the spectrum in the fast-neutron irradiation facility in NRU is representative of the actual irradiation spectrum for fast-neutron damage in a CANDU reactor. The effects of changes in specimen holder size, temperature, coolant density and fuel burnup on the fast neutron spectrum are also presented. (author). 9 refs., 3 tabs., 4 figs

  2. Sequential measurements of spectrum and dose for cosmic-ray neutrons on the ground

    International Nuclear Information System (INIS)

    Hirabayashi, N.; Nunomiya, T.; Suzuki, H.; Nakamura, T.

    2002-01-01

    The earth is continually bathed in high-energy particles that come from outside the solar system, known as galactic cosmic rays. When these particles penetrate the magnetic fields of the solar system and the Earth and reach the Earth's atmosphere, they collide with atomic nuclei in air and secondary cosmic rays of every kind. On the other hand, levels of accumulation of the semiconductor increase recently, and the soft error that the cosmic-ray neutrons cause has been regarded as questionable. There have been long-term measurements of cosmic-ray neutron fluence at several places in the world, but no systematic study on cosmic-ray neutron spectrum measurements. This study aimed to measure the cosmic-ray neutron spectrum and dose on the ground during the solar maximum period of 2000 to 2002. Measurements have been continuing in a cabin of Tohoku University Kawauchi campus, by using five multi-moderator spectrometers (Bonner sphere), 12.7 cm diam by 12.7 cm long NE213 scintillator, and rem counter. The Bonner sphere uses a 5.08 cm diam spherical 3 He gas proportional counter and the rem counter uses a 12.7 cm diam 3 He gas counter. The neutron spectra were obtained by unfolding from the count rates measured with the Bonner sphere using the SAND code and the pulse height spectra measured with the NE213 scintillator using the FORIST code . The cosmic- ray neutron spectrum and ambient dose rates have been measured sequentially from April 2001. Furthermore, the correlation between ambient dose rate and the atmospheric pressure was investigated with a barometer. We are also very much interested in the variation of neutron spectrum following big solar flares. From the sequential measurements, we found that the cosmic-ray neutron spectrum has two peaks at around 1 MeV and at around 100 MeV, and the higher energy peak increases with a big solar flare

  3. Approaches for the generation of a covariance matrix for the Cf-252 fission-neutron spectrum

    International Nuclear Information System (INIS)

    Mannhart, W.

    1983-01-01

    After a brief retrospective glance is cast at the situation, the evaluation of the Cf-252 neutron spectrum with a complete covariance matrix based on the results of integral experiments is proposed. The different steps already taken in such an evaluation and work in progress are reviewed. It is shown that special attention should be given to the normalization of the neutron spectrum which must be reflected in the covariance matrix. The result of the least-squares adjustment procedure applied can easily be combined with the results of direct spectrum measurements and should be regarded as the first step in a new evaluation of the Cf-252 fission-neutron spectrum. (author)

  4. Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-03-01

    The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

  5. Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2006-01-01

    The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

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

    CERN Document Server

    Mukherjee, B

    1999-01-01

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

  7. Neutron fluence rate and energy spectrum in SPRR-300 reactor thermal column

    International Nuclear Information System (INIS)

    Dou Haifeng; Dai Junlong

    2006-01-01

    In order to modify the simple one-dimension model, the neutron fluence rate distribution calculated with ANISN code ws checked with that calculated with MCNP code. To modify the error caused by ignoring the neutron landscape orientation leaking, the reflector that can't be modeled in a simple one-dimension model was dealt by extending landscape orientation scale. On this condition the neutron fluence rate distribution and the energy spectrum in the thermal column of SPRR-300 reactor were calculated with one-dimensional code ANISN, and the results of Cd ratio are well accorded with the experimental results. The deviation between them is less than 5% and it isn't above 10% in one or two special positions. It indicates that neutron fluence rate distribution and energy spectrum in the thermal column can be well calculated with one-dimensional code ANISN. (authors)

  8. Modeling the Complete Gravitational Wave Spectrum of Neutron Star Mergers.

    Science.gov (United States)

    Bernuzzi, Sebastiano; Dietrich, Tim; Nagar, Alessandro

    2015-08-28

    In the context of neutron star mergers, we study the gravitational wave spectrum of the merger remnant using numerical relativity simulations. Postmerger spectra are characterized by a main peak frequency f2 related to the particular structure and dynamics of the remnant hot hypermassive neutron star. We show that f(2) is correlated with the tidal coupling constant κ(2)^T that characterizes the binary tidal interactions during the late-inspiral merger. The relation f(2)(κ(2)^T) depends very weakly on the binary total mass, mass ratio, equation of state, and thermal effects. This observation opens up the possibility of developing a model of the gravitational spectrum of every merger unifying the late-inspiral and postmerger descriptions.

  9. Study on neutron spectrum for effective transmutation of minor actinides in thermal reactors

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yokoyama, Kenji

    1997-01-01

    The transmutation of minor actinides (MAs) has been investigated in thermal reactor cells using mixed oxide fuel with MAs. The effect of neutron spectra on transmutation is studied by changing the neutron spectra. Five transmutation rates are compared: direct fission incineration rate, capture transmutation rate, consumption rate, overall fission incineration rate and inventory difference transmutation rate. The relations between these transmutation rates and their dependence on the neutron spectrum were investigated. To effectively incinerate MAs it is necessary to maximize the overall fission incineration rate and the inventory difference transmutation rate. These transmutation rates become maximum when the fraction of neutrons below 1 eV is about 8% for the case where the MA addition is 1-3%. When the MA addition is over 5%, the transmutation rates become maximum for very hard neutron spectrum. (Author)

  10. Techniques of in vivo neutron activation analysis

    International Nuclear Information System (INIS)

    Chettle, D.R.; Fremlin, J.H.

    1984-01-01

    This review is dealt with under the following headings, intended to reflect the different factors affecting the measurement sensitivity, starting with the choice of neutron source and proceeding, through the reaction characteristics, to the detection system, the questions of dosimetry and ethical constraints being also discussed: 1) neutron sources, slowing down and interaction processes, energy spectrum and flux uniformity, timing 2) neutron reactions used for in vivo analyses 3) detectors, choice, geometrical considerations and detector shielding 4) data collection and processing 5) interpretation, major elements, absolute or sequential measurements, relationship to other parameters 6) dosimetry, framework for dose levels, biological effects of neutron interactions, neutron doses in practice 7) implications for measurement of calcium, nitrogen and cadmium. (U.K.)

  11. Neutron spectrum in small iron pile surrounded by lead reflector

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Hayashi, S.A.; Kobayashi, Katsuhei; Matsumura, Tetsuo; Nishihara, Hiroshi.

    1978-01-01

    In order to save the quantity of sample material, a possibility to assess group constants of a reactor material through measurement and analysis of neutron spectrum in a small sample pile surrounded by a reflector of heavy moderator, was investigated. As the sample and the reflector, we chose iron and lead, respectively. Although the time dispersion in moderation of neutrons was considerably prolonged by the lead reflector, this hardly interferes with the assessment of group constants. Theoretical calculation revealed that both the neutron flux spectrum and the sensitivity coefficient of group constants in an iron sphere, 35 cm in diameter surrounded by the lead reflector, 25 cm thick, were close to those of the bare iron sphere, 108 cm in diameter. The neutron spectra in a small iron pile surrounded by a lead reflector were experimentally obtained by the time-of-flight method with an electron linear accelerator and the result was compared with the predicted values. It could be confirmed that a small sample pile surrounded by a reflector, such as lead, was as useful as a much larger bulk pile for the assessment of group constants of a reactor material. (auth.)

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

    Science.gov (United States)

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

    2017-11-01

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

  13. 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

  14. Experimental Validation of Ex-Vessel Neutron Spectrum by Means of Dosimeter Materials Activation Method

    Directory of Open Access Journals (Sweden)

    S.A. Santa

    2017-06-01

    Full Text Available Neutron spectrum information in reactor core and around of ex-vessel reactor needs to be known with a certain degree of accuracy to support the development of fuels, materials, and other components. The most common method to determine neutron spectra is by utilizing the radioactivation of dosimeter materials. This report presents the evaluation of neutron flux incident on M3dosimeter sets which were irradiated outside the reactor vessel,as well as the validation of  neutron spectrum calculation. Al capsules containing both dosimeter set covered withCd and dosimeter set without Cd cover have been irradiated during the 35th operational cycle in the M3 ex-vessel irradiation hole position207 cmfrom core centerline at the space between the reactor vessel and the safety vessel. The capsules were positioned at Z=0.0 cm of core midplane. Each dosimeter set consists of Co-Al, Sc, Fe, Np, Nb, Ni, B, and Ta. The gamma-ray spectra of irradiated dosimeter materials were measured by 63 cc HPGe solid-state detector and photo-peak spectra were analyzed using BOB75 code. The reaction rates of each dosimeter materials and its uncertainty were analyzed based on 59Co (n,g 60Co, 237Np (n,f 95Zr-103Ru,  45Sc (n,g 46Sc, 58Fe (n,g 59Fe, 181Ta (n,g 182Ta, and 58Ni (n,p58Co reactions. The measured Cd ratios indicate that neutron spectrum at the irradiated dosimeter sets was dominated by low energy neutron. The experimental result shows that the calculated neutron spectra by DORT code at the ex-vessel positions need correction, especially in the fast neutron energy region, so as to obtain reasonable unfolding result consistent with the reaction rate measurement without any exception. Using biased DORT initial spectrum, the neutron spectrum and its integral quantity were unfolded by NEUPAC code. The result shows that total neutron flux, flux above 1.0 MeV, flux above 0.1 MeV, and the displacement rate of the dosimeter set not covered with Cd were 1.75× 1012 n cm2 s-1, 1

  15. Approximation for the adjoint neutron spectrum; Aproximacao para o espectro adjunto de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Suster, Luis Carlos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    The proposal of this work is the determination of an analytical approximation which is capable to reproduce the adjoint neutron flux for the energy range of the narrow resonances (NR). In a previous work we developed a method for the calculation of the adjoint spectrum which was calculated from the adjoint neutron balance equations, that were obtained by the collision probabilities method, this method involved a considerable quantity of numerical calculation. In the analytical method some approximations were done, like the multiplication of the escape probability in the fuel by the adjoint flux in the moderator, and after these approximations, taking into account the case of the narrow resonances, were substituted in the adjoint neutron balance equation for the fuel, resulting in an analytical approximation for the adjoint flux. The results obtained in this work were compared to the results generated with the reference method, which demonstrated a good and precise results for the adjoint neutron flux for the narrow resonances. (author)

  16. Bayesian statistics applied to neutron activation data for reactor flux spectrum analysis

    International Nuclear Information System (INIS)

    Chiesa, Davide; Previtali, Ezio; Sisti, Monica

    2014-01-01

    Highlights: • Bayesian statistics to analyze the neutron flux spectrum from activation data. • Rigorous statistical approach for accurate evaluation of the neutron flux groups. • Cross section and activation data uncertainties included for the problem solution. • Flexible methodology applied to analyze different nuclear reactor flux spectra. • The results are in good agreement with the MCNP simulations of neutron fluxes. - Abstract: In this paper, we present a statistical method, based on Bayesian statistics, to analyze the neutron flux spectrum from the activation data of different isotopes. The experimental data were acquired during a neutron activation experiment performed at the TRIGA Mark II reactor of Pavia University (Italy) in four irradiation positions characterized by different neutron spectra. In order to evaluate the neutron flux spectrum, subdivided in energy groups, a system of linear equations, containing the group effective cross sections and the activation rate data, has to be solved. However, since the system’s coefficients are experimental data affected by uncertainties, a rigorous statistical approach is fundamental for an accurate evaluation of the neutron flux groups. For this purpose, we applied the Bayesian statistical analysis, that allows to include the uncertainties of the coefficients and the a priori information about the neutron flux. A program for the analysis of Bayesian hierarchical models, based on Markov Chain Monte Carlo (MCMC) simulations, was used to define the problem statistical model and solve it. The first analysis involved the determination of the thermal, resonance-intermediate and fast flux components and the dependence of the results on the Prior distribution choice was investigated to confirm the reliability of the Bayesian analysis. After that, the main resonances of the activation cross sections were analyzed to implement multi-group models with finer energy subdivisions that would allow to determine the

  17. Neutron flux measurements in C-9 capsule pressure tube

    International Nuclear Information System (INIS)

    Barbos, D.; Roth, C. S.; Gugiu, D.; Preda, M.

    2001-01-01

    C-9 capsule is a fuel testing facility in which the testing consists of a daily cycle ranging between the limits 100% power to 50% power. C-9 in-pile section with sample holder an instrumentation are introduced in G-9 and G-10 experimental channels. The experimental fuel channel has a maximum value when the in-pile section (pressure tube) is in G-9 channel and minimum value in G-10 channel. In this paper the main goals are determination or measurements of: - axial thermal neutron flux distribution in C-9 pressure tube both in G-9 and G-10 channel; - ratio of maximum neutron flux value in G-9 and the same value in G-9 channel and the same value in G-10 channel; - neutron flux-spectrum. On the basis of axial neutron flux distribution measurements, the experimental fuel element in sample holder position in set. Both axial neutron flux distribution of thermal neutrons and neutron flux-spectrum were performed using multi- foil activation technique. Activation rates were obtained by absolute measurements of the induced activity using gamma spectroscopy methods. To determine the axial thermal neutron flux distribution in G-9 and G-10, Cu 100% wire was irradiated at the reactor power of 2 MW. Ratio between the two maximum values, in G-9 and G-10 channels, is 2.55. Multi-foil activation method was used for neutron flux spectrum measurements. The neutron spectra and flux were obtained from reaction rate measurements by means of SAND 2 code. To obtain gamma-ray spectra, a HPGe detector connected to a multichannel analyzer was used. The spectrometer is absolute efficiency calibrated. The foils were irradiated at 2 MW reactor power in previously determined maximum flux position resulted from wire measurements. This reaction rates were normalized for 10 MW reactor power. Neutron self shielding corrections for the activation foils were applied. The self-shielding corrections are computed using Monte Carlo simulation methods. The measured integral flux is 1.1·10 14 n/cm 2 s

  18. Absolute calibration of neutron detectors on the C-2U advanced beam-driven FRC

    Energy Technology Data Exchange (ETDEWEB)

    Magee, R. M., E-mail: rmagee@trialphaenergy.com; Clary, R.; Korepanov, S.; Jauregui, F.; Allfrey, I.; Garate, E.; Valentine, T.; Smirnov, A. [Tri Alpha Energy, Inc., Rancho Santa Margarita, California 92688 (United States)

    2016-11-15

    In the C-2U fusion energy experiment, high power neutral beam injection creates a large fast ion population that sustains a field-reversed configuration (FRC) plasma. The diagnosis of the fast ion pressure in these high-performance plasmas is therefore critical, and the measurement of the flux of neutrons from the deuterium-deuterium (D-D) fusion reaction is well suited to the task. Here we describe the absolute, in situ calibration of scintillation neutron detectors via two independent methods: firing deuterium beams into a high density gas target and calibration with a 2 × 10{sup 7} n/s AmBe source. The practical issues of each method are discussed and the resulting calibration factors are shown to be in good agreement. Finally, the calibration factor is applied to C-2U experimental data where the measured neutron rate is found to exceed the classical expectation.

  19. Experimental study of modification of neutron spectrum using filters

    International Nuclear Information System (INIS)

    Kobayashi, H.; Matsubayashi, M.; Brenizer, J.S. Jr.; Lindsay, J.T.

    1996-01-01

    Filter effects for continuum thermal neutron beams were experimentally studied by means of an effective energy. Be, Bi and Pb were used as filter materials to examine the energy shift of the spectrum. It was found that the effective energy of a thermal neutron beam is easily lowered into a sub-thermal region (down to 10 meV) by use of a filter system without any cooling system. The effectiveness and its applicability will be discussed in this study. (orig.)

  20. Neutron sources and their characteristics

    International Nuclear Information System (INIS)

    McCall, R.C.; Swanson, W.P.

    1979-03-01

    The significant sources of photoneutrons within a linear-accelerator treatment head are identified and absolute estimates of neutron production per treatment dose are given for typical components. It is found that the high-Z materials within the treatment head do not significantly alter the neutron fluence but do substantially reduce the average energy of the transmitted spectrum. Reflection of neutrons from the concrete treatment room contribute to the neutron fluence, but not substantially to the patient integral dose, because of a further reduction in average energy. The ratio of maximum fluence to the treatment dose at the same distance is given as a function of electron energy. This ratio rises with energy to an almost constant value of 2.1 x 10 5 neutrons cm -2 rad -1 at electron energies above about 25 MeV. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. Reasons for apparent deviations are suggested. Absolute depth-dose and depth-dose-equivalent distributions for realistic neutron spectra that occur at therapy installations are calculated, and a rapid falloff with depth is found. The ratio of neutron integral absorbed dose to leakage photon absorbed dose is estimated to be 0.04 and 0.2 for 14 to 25 MeV incident electron energy, respectively. Possible reasons are given for lesser neutron production from betatrons than from linear accelerators. Possible ways in which neutron production can be reduced are discussed

  1. 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.

  2. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  3. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  4. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in

  5. Evaluating the 239Pu Prompt Fission Neutron Spectrum Induced by Thermal to 30 MeV Neutrons

    Directory of Open Access Journals (Sweden)

    Neudecker D.

    2016-01-01

    Full Text Available We present a new evaluation of the 239Pu prompt fission neutron spectrum (PFNS induced by thermal to 30 MeV neutrons. Compared to the ENDF/B-VII.1 evaluation, this one includes recently published experimental data as well as an improved and extended model description to predict PFNS. For instance, the pre-equilibrium neutron emission component to the PFNS is considered and the incident energy dependence of model parameters is parametrized more realistically. Experimental and model parameter uncertainties and covariances are estimated in detail. Also, evaluated covariances are provided between all PFNS at different incident neutron energies. Selected evaluation results and first benchmark calculations using this evaluation are briefly discussed.

  6. New Measurements and Calculations to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)

    2011-07-01

    Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)

  7. A neutron spectrum unfolding computer code based on artificial neural networks

    Science.gov (United States)

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

    2014-02-01

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

  8. Neutron spectrum unfolding using computer code SAIPS

    International Nuclear Information System (INIS)

    Karim, S.

    1999-01-01

    The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results. (author)

  9. Monte Carlo simulation of fission yields, kinetic energy, fission neutron spectrum and decay γ-ray spectrum for 232Th(n,f) reaction induced by 3H(d,n) 4He neutron source

    International Nuclear Information System (INIS)

    Zheng Wei; Zeen Yao; Changlin Lan; Yan Yan; Yunjian Shi; Siqi Yan; Jie Wang; Junrun Wang; Jingen Chen; Chinese Academy of Sciences, Shanghai

    2015-01-01

    Monte Carlo transport code Geant4 has been successfully utilised to study of neutron-induced fission reaction for 232 Th in the transport neutrons generated from 3 H(d,n) 4 He neutron source. The purpose of this work is to examine the applicability of Monte Carlo simulations for the computation of fission reaction process. For this, Monte Carlo simulates and calculates the characteristics of fission reaction process of 232 Th(n,f), such as the fission yields distribution, kinetic energy distribution, fission neutron spectrum and decay γ-ray spectrum. This is the first time to simulate the process of neutron-induced fission reaction using Geant4 code. Typical computational results of neutron-induced fission reaction of 232 Th(n,f) reaction are presented. The computational results are compared with the previous experimental data and evaluated nuclear data to confirm the certain physical process model in Geant4 of scientific rationality. (author)

  10. 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

  11. An investigation of TRU recycling with various neutron spectrums

    International Nuclear Information System (INIS)

    Yong-Nam, Kim; Hong-Chul, Kim; Chi-Young, Han; Jong-Kyung, Kim; Won-Seok Park

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single batch fuel loading, the burn-up calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analysed in terms of burn-up reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behaviour between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction. (author)

  12. Characterization of a deuterium-deuterium plasma fusion neutron generator

    Science.gov (United States)

    Lang, R. F.; Pienaar, J.; Hogenbirk, E.; Masson, D.; Nolte, R.; Zimbal, A.; Röttger, S.; Benabderrahmane, M. L.; Bruno, G.

    2018-01-01

    We characterize the neutron output of a deuterium-deuterium plasma fusion neutron generator, model 35-DD-W-S, manufactured by NSD/Gradel-Fusion. The measured energy spectrum is found to be dominated by neutron peaks at 2.2 MeV and 2.7 MeV. A detailed GEANT4 simulation accurately reproduces the measured energy spectrum and confirms our understanding of the fusion process in this generator. Additionally, a contribution of 14 . 1 MeV neutrons from deuterium-tritium fusion is found at a level of 3 . 5%, from tritium produced in previous deuterium-deuterium reactions. We have measured both the absolute neutron flux as well as its relative variation on the operational parameters of the generator. We find the flux to be proportional to voltage V 3 . 32 ± 0 . 14 and current I 0 . 97 ± 0 . 01. Further, we have measured the angular dependence of the neutron emission with respect to the polar angle. We conclude that it is well described by isotropic production of neutrons within the cathode field cage.

  13. A compact proton spectrometer for measurement of the absolute DD proton spectrum from which yield and ρR are determined in thin-shell inertial-confinement-fusion implosions

    Energy Technology Data Exchange (ETDEWEB)

    Rosenberg, M. J., E-mail: mrosenbe@mit.edu; Zylstra, A. B.; Frenje, J. A.; Rinderknecht, H. G.; Gatu Johnson, M.; Waugh, C. J.; Séguin, F. H.; Sio, H.; Sinenian, N.; Li, C. K.; Petrasso, R. D. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Glebov, V. Yu.; Hohenberger, M.; Stoeckl, C.; Sangster, T. C. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States); Yeamans, C. B.; LePape, S.; Mackinnon, A. J.; Bionta, R. M.; Talison, B. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

    2014-10-01

    A compact, step range filter proton spectrometer has been developed for the measurement of the absolute DD proton spectrum, from which yield and areal density (ρR) are inferred for deuterium-filled thin-shell inertial confinement fusion implosions. This spectrometer, which is based on tantalum step-range filters, is sensitive to protons in the energy range 1-9 MeV and can be used to measure proton spectra at mean energies of ~1-3 MeV. It has been developed and implemented using a linear accelerator and applied to experiments at the OMEGA laser facility and the National Ignition Facility (NIF). Modeling of the proton slowing in the filters is necessary to construct the spectrum, and the yield and energy uncertainties are ±<10% in yield and ±120 keV, respectively. This spectrometer can be used for in situ calibration of DD-neutron yield diagnostics at the NIF.

  14. Prompt neutron fission spectrum mean energies for the fissile nuclides and 252Cf

    International Nuclear Information System (INIS)

    Holden, N.E.

    1985-01-01

    The international standard for a neutron spectrum is that produced from the spontaneous fission of 252 Cf, while the thermal neutron induced fission neutron spectra for the four fissile nuclides, 233 U, 235 U, 239 Pu, and 241 Pu are of interest from the standpoint of nuclear reactors. The average neutron energies of these spectra are tabulated. The individual measurements are recorded with the neutron energy range measured, the method of detection as well as the average neutron energy for each author. Also tabulated are the measurements of the ratio of mean energies for pairs of fission neutron spectra. 75 refs., 9 tabs

  15. Neutron energy spectrum in graphite blankets of fusion reactors

    International Nuclear Information System (INIS)

    Tsechanski, A.

    1981-09-01

    Neutron flux measurements were performed in a graphite stack and compared with calculations made with a two dimensional transport computer code. In the present work it is observed that the calculated spectrum in the elastic and inelastic scattering ranges (the first collision range in both cases), is sensitive to details of the angular distribution of these neutrons. Regarding the discrepancies in the elastic scattering range it is concluded that the microscopic cross section library ENDF/B-IV overestimates the large angle scattering (back scattering) as can be seen from comparison of measured and calculated spectra. The two most important conclusions of the present work are: 1. Inelastic scattering interaction of D-T neutrons in graphite cannot be calculated without a proper account of energy-angle correlation. 2. An experimental setup supplying monoenergetic collimated D-T neutrons constitutes a sensitive although indirect means for measuring angular distributions in inelastic and elastic scattering

  16. Calculations of neutron spectra after neutron-neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-09-01

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

  17. Neutron spectrum measurement in D + Be reaction

    CERN Document Server

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

    2002-01-01

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

  18. Fission neutron spectrum averaged cross sections for threshold reactions on arsenic

    International Nuclear Information System (INIS)

    Dorval, E.L.; Arribere, M.A.; Kestelman, A.J.; Comision Nacional de Energia Atomica, Cuyo Nacional Univ., Bariloche; Ribeiro Guevara, S.; Cohen, I.M.; Ohaco, R.A.; Segovia, M.S.; Yunes, A.N.; Arrondo, M.; Comision Nacional de Energia Atomica, Buenos Aires

    2006-01-01

    We have measured the cross sections, averaged over a 235 U fission neutron spectrum, for the two high threshold reactions: 75 As(n,p) 75 mGe and 75 As(n,2n) 74 As. The measured averaged cross sections are 0.292±0.022 mb, referred to the 3.95±0.20 mb standard for the 27 Al(n,p) 27 Mg averaged cross section, and 0.371±0.032 mb referred to the 111±3 mb standard for the 58 Ni(n,p) 58m+g Co averaged cross section, respectively. The measured averaged cross sections were also evaluated semi-empirically by numerically integrating experimental differential cross section data extracted for both reactions from the current literature. The calculations were performed for four different representations of the thermal-neutron-induced 235 U fission neutron spectrum. The calculated cross sections, though depending on analytical representation of the flux, agree with the measured values within the estimated uncertainties. (author)

  19. The gravitational wave spectrum of non-axisymmetric, freely precessing neutron stars

    International Nuclear Information System (INIS)

    Broeck, Chris van den

    2005-01-01

    Evidence for free precession has been observed in the radio signature of several pulsars. Freely precessing pulsars radiate gravitationally at frequencies near the rotation rate and twice the rotation rate, which for rotation frequencies greater than ∼10 Hz is in the LIGO band. In older work, the gravitational wave spectrum of a precessing neutron star has been evaluated to first order in a small precession angle. Here, we calculate the contributions to second order in the wobble angle, and we find that a new spectral line emerges. We show that for reasonable wobble angles, the second-order line may well be observable with the proposed advanced LIGO detectors for precessing neutron stars as far away as the galactic centre. Observation of the full second-order spectrum permits a direct measurement of the star's wobble angle, oblateness and deviation from axisymmetry, with the potential to significantly increase our understanding of neutron star structure

  20. Neutron spectrum survey around the cyclotron of IEN/Brazilian CNEN: calibration of neutron personnel dosemeter

    International Nuclear Information System (INIS)

    Fajardo, P.W.

    1991-01-01

    The albedo neutron dosimeter is calibrated directly at the work place due to its high energy dependence. This thesis deals with the study, analysis and application of neutron measurement techniques in order to obtain information about the neutron spectrum and neutron dose equivalent at several representative working places of the cyclotron laboratory of the Nuclear Engineering Institute (IEN). These data are employed mainly in the calibration of the brazilian albedo neutron dosimeter. Bonner spheres and foil activation were used in neutron spectra measurements and the neutron dose equivalents were measured with the single sphere albedo technique. BF 3 and 3 He proportional detectors and 6 LiI scintillation detector were also used in these measurements. The single sphere technique turned out to be more appropriate for neutron dosimetry for calibrating the albedo dosimeter in the varying fields of the cyclotron. Calibration the albedo dosimeter in the varying fields of the cyclotron. Calibration factors were found for routine applications, when the workers are protected by shielding and for radiological accident applications, in the case that a worker is exposed inside the cyclotron room. In all situations the performance of the brazilian albedo dosimeter is compared with that of the german albedo dosimeters. (author)

  1. Pure Absolutely Continuous Spectrum for Random Operators on $l^2(Z^d)$ at Low Disorder

    CERN Document Server

    Grinshpun, V

    2006-01-01

    Absence of singular continuous component, with probability one, in the spectra of random perturbations of multidimensional finite-difference Hamiltonians, is for the first time rigorously established under certain conditions ensuring either absence of point component, or absence of absolutely continuous component in the corresponding regions of spectra. The main technical tool involved is the rank-one perturbation theory of singular spectra. The respective new result (the non-mixing property) is applied to establish existence and bounds of the (non-empty) pure absolutely continuous component in the spectrum of the Anderson model with bounded random potential in dimension d=2 at low disorder (similar proof holds for d>4). The new result implies, via the trace-class perturbation analysis, Anderson model with the unbounded random potential having only pure point spectrum (complete system of localized wave-functions) with probability one in arbitrary dimension. The basic idea is to establish absence of the mixed,...

  2. Use of new threshold detector 199Hg(n,n')/sup 199m/Hg for neutron spectrum unfolding

    International Nuclear Information System (INIS)

    Sakurai, K.

    1982-01-01

    The nuclear data for the 199 Hg(n,n')/sup 199m/Hg reaction are reviewed and the data are used for neutron spectrum unfolding. The neutron spectrum of the YAYOI glory-hole is unfolded by SAND II with 10 nuclear reactions including the 199 Hg(n,n')/sup 199m/Hg reaction. The ratio of the measured reaction rate to the calculated reaction rate is about 1:1.1 for the guess spectrum. The 199 Hg(n,n')/sup 199m/Hg, 115 In(n,n')/sup 115m/In, 103 Rh(n,n')/sup 103m/Rh reactions should be useful threshold detectors for the neutron dosimetry with low level fast neutron flux

  3. Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor

    International Nuclear Information System (INIS)

    Zhou Xuemei; Liu Guimin; Wang Xiaohe; Li Da; Meng Lingjie

    2014-01-01

    Neutron spectrum should be measured before test samples are irradiated. Neutron spectrum in an irradiation chamber of a research reactor was measured by using activation method when the reactor is in normal operation under 2 MW. Sixteen kinds of non-fission foils (19 reaction channels) were selected, of which 10 were sensitive to thermal and intermediate energy regions, while the others were of different threshold energy and sensitive to fast energy regions. By measuring the foil radioactivity, the neutron spectrum was unfolded with the iterative methods SAND-II and MSIT. Finally, shielding corrections of group cross-section and main factors affecting the calculation accuracy were studied and the uncertainty of solution was analyzed using the Monte Carlo method in the process of SAND-II. (authors)

  4. Neutron energy spectrum determination near the surface on the JET vacuum vessel using the multifoil activation technique

    Energy Technology Data Exchange (ETDEWEB)

    Pillon, M.; Jarvis, O.N.; Conroy, S. (Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy) JET Joint Undertaking, Abingdonm Oxon (U.K.) Imperial College of Science, Technology and Medicine, London (U.K.))

    1990-03-01

    The activation of foils of zinc, indium, aluminium, copper and magnesium has been used as a means of examining the energy spectrum of neutrons produced by discharges in the Joint European Torus (JET). Several threshold reactions have been used together with a least-squares unfolding code to determine the 2.5 and 14 MeV neutron yields produced by the JET plasma. The analysis shows that the energy spectrum produced by downscattered neutrons is satisfactorily calculated with the MCNP neutron transport code.

  5. Calibration of time of flight detectors using laser-driven neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Mirfayzi, S. R.; Kar, S., E-mail: s.kar@qub.ac.uk; Ahmed, H.; Green, A.; Alejo, A.; Jung, D. [Centre for Plasma Physics, School of Mathematics and Physics, Queen’s University Belfast, Belfast BT7 1NN (United Kingdom); Krygier, A. G.; Freeman, R. R. [Department of Physics, The Ohio State University, Columbus, Ohio 43210 (United States); Clarke, R. [Central Laser Facility, Rutherford Appleton Laboratory, Didcot, Oxfordshire OX11 0QX (United Kingdom); Fuchs, J.; Vassura, L. [LULI, Ecole Polytechnique, CNRS, Route de Saclay, 91128 Palaiseau Cedex (France); Kleinschmidt, A.; Roth, M. [Institut für Kernphysik, Technische Universität Darmstadt, Schloßgartenstrasse 9, D-64289 Darmstadt,Germany (Germany); Morrison, J. T. [Propulsion Systems Directorate, Air Force Research Lab, Wright Patterson Air Force Base, Ohio 45433 (United States); Najmudin, Z.; Nakamura, H. [Blackett Laboratory, Department of Physics, Imperial College, London SW7 2AZ (United Kingdom); Norreys, P. [Central Laser Facility, Rutherford Appleton Laboratory, Didcot, Oxfordshire OX11 0QX (United Kingdom); Department of Physics, University of Oxford, Oxford OX1 3PU (United Kingdom); Oliver, M. [Department of Physics, University of Oxford, Oxford OX1 3PU (United Kingdom); Zepf, M. [Centre for Plasma Physics, School of Mathematics and Physics, Queen’s University Belfast, Belfast BT7 1NN (United Kingdom); Helmholtz Institut Jena, D-07743 Jena (Germany); Borghesi, M. [Centre for Plasma Physics, School of Mathematics and Physics, Queen’s University Belfast, Belfast BT7 1NN (United Kingdom); Institute of Physics of the ASCR, ELI-Beamlines Project, Na Slovance 2, 18221 Prague (Czech Republic)

    2015-07-15

    Calibration of three scintillators (EJ232Q, BC422Q, and EJ410) in a time-of-flight arrangement using a laser drive-neutron source is presented. The three plastic scintillator detectors were calibrated with gamma insensitive bubble detector spectrometers, which were absolutely calibrated over a wide range of neutron energies ranging from sub-MeV to 20 MeV. A typical set of data obtained simultaneously by the detectors is shown, measuring the neutron spectrum emitted from a petawatt laser irradiated thin foil.

  6. Calibration of time of flight detectors using laser-driven neutron source

    Science.gov (United States)

    Mirfayzi, S. R.; Kar, S.; Ahmed, H.; Krygier, A. G.; Green, A.; Alejo, A.; Clarke, R.; Freeman, R. R.; Fuchs, J.; Jung, D.; Kleinschmidt, A.; Morrison, J. T.; Najmudin, Z.; Nakamura, H.; Norreys, P.; Oliver, M.; Roth, M.; Vassura, L.; Zepf, M.; Borghesi, M.

    2015-07-01

    Calibration of three scintillators (EJ232Q, BC422Q, and EJ410) in a time-of-flight arrangement using a laser drive-neutron source is presented. The three plastic scintillator detectors were calibrated with gamma insensitive bubble detector spectrometers, which were absolutely calibrated over a wide range of neutron energies ranging from sub-MeV to 20 MeV. A typical set of data obtained simultaneously by the detectors is shown, measuring the neutron spectrum emitted from a petawatt laser irradiated thin foil.

  7. Calibration of time of flight detectors using laser-driven neutron source

    International Nuclear Information System (INIS)

    Mirfayzi, S. R.; Kar, S.; Ahmed, H.; Green, A.; Alejo, A.; Jung, D.; Krygier, A. G.; Freeman, R. R.; Clarke, R.; Fuchs, J.; Vassura, L.; Kleinschmidt, A.; Roth, M.; Morrison, J. T.; Najmudin, Z.; Nakamura, H.; Norreys, P.; Oliver, M.; Zepf, M.; Borghesi, M.

    2015-01-01

    Calibration of three scintillators (EJ232Q, BC422Q, and EJ410) in a time-of-flight arrangement using a laser drive-neutron source is presented. The three plastic scintillator detectors were calibrated with gamma insensitive bubble detector spectrometers, which were absolutely calibrated over a wide range of neutron energies ranging from sub-MeV to 20 MeV. A typical set of data obtained simultaneously by the detectors is shown, measuring the neutron spectrum emitted from a petawatt laser irradiated thin foil

  8. “Influence Method” applied to measure a moderated neutron flux

    International Nuclear Information System (INIS)

    Rios, I.J.; Mayer, R.E.

    2016-01-01

    The “Influence Method” is conceived for the absolute determination of a nuclear particle flux in the absence of known detector efficiency. This method exploits the influence of the presence of one detector, in the count rate of another detector when they are placed one behind the other and define statistical estimators for the absolute number of incident particles and for the efficiency. The method and its detailed mathematical description were recently published (Rios and Mayer, 2015 [1]). In this article we apply it to the measurement of the moderated neutron flux produced by an "2"4"1AmBe neutron source surrounded by a light water sphere, employing a pair of "3He detectors. For this purpose, the method is extended for its application where particles arriving at the detector obey a Poisson distribution and also, for the case when efficiency is not constant over the energy spectrum of interest. Experimental distributions and derived parameters are compared with theoretical predictions of the method and implications concerning the potential application to the absolute calibration of neutron sources are considered. - Highlights: • “Influence Method” applied to measure a moderated neutron flux. • Effective efficiency defined independently of calibration sources. • Neutron sources calibration discussion.

  9. Iterative code for the reconstruction of the neutrons spectrum using the Bonner spheres

    International Nuclear Information System (INIS)

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

    2012-10-01

    The neutrons are the particles more difficult of detecting for their intrinsic nature. The absence of the neutrons charge makes that an interaction exists with the matter in a different way. The term radiation spectrometry can use to describe the measurement of the intensity of a radiation field with regard to the energy. The intensity distribution with relationship to the energy is commonly known as spectrum. A method to know the neutrons spectrum in the radiation fields to those that people are exposed is the use of the known system as spectrometry system of Bonner spheres, being the more used for the purposes of the radiological protection. The current interest in the electrons spectrometry has stimulated the development of several procedures to carry out the reconstruction of the spectra. During the last decades new codes have been developed such as BUNKIUT, Bums, Fruit, UMG, etc., however, these methods still present several inconveniences as the complexity in their use, the necessity of an expert user and a very near initial spectrum to the spectrum that is wanted to obtain. To solve the mentioned problems it was development the program NSDUAZ (Neutron Spectrometry and Dosimetry from Autonomous University of Zacatecas). The objective of the present work is to prove and to validate the code before mentioned making an analysis of likeness and differences and of advantages and disadvantages with relationship to the codes used at the present time. (Author)

  10. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    International Nuclear Information System (INIS)

    Sugimoto, Masayoshi

    2001-01-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  11. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  12. Effect of normalization on the neutron spectrum adjustment procedure

    International Nuclear Information System (INIS)

    Zsolnay, E.M.; Zijp, W.L.; Nolthenius, H.J.

    1983-10-01

    Various computer programs currently applied for neutron spectrum adjustment based on multifoil activation data, use different ways to determine the normalization factor to be applied to an unnormalized input spectrum. The influence is shown of the various definitions of the normalization factor on the adjusted results for the case of the ORR and YAYOI spectra considered in the international REAL-80 exercise. The actual expression for defining the normalization factor is more important than previously assumed. The theory of the generalized least squares principle provides an optimal definition for the normalization factor

  13. Neutron spectrum adjustment. The role of covariances

    International Nuclear Information System (INIS)

    Remec, I.

    1992-01-01

    Neutron spectrum adjustment method is shortly reviewed. Practical example dealing with power reactor pressure vessel exposure rates determination is analysed. Adjusted exposure rates are found only slightly affected by the covariances of measured reaction rates and activation cross sections, while the multigroup spectra covariances were found important. Approximate spectra covariance matrices, as suggested in Astm E944-89, were found useful but care is advised if they are applied in adjustments of spectra at locations without dosimetry. (author) [sl

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Absolute measurement of {beta} activities and application to the determination of neutronic densities; Mesure absolue d'activites {beta} et application a la determination des densites neutronique

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, R [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1951-01-15

    M. Berthelot, to my entrance to the ''Commissariat a l 'Energie Atomique'', proposed me to study the absolute measurement of neutron densities. Very quickly the problem of the absolute activity of {beta} sources became the central object of this work. In a first part, we will develop the methods of absolute determination for {beta} activities. The use of a 4{pi} counter permits to get the absolute activity of all beta radioactive source, susceptible to be put as thin leaf and of period superior than some minutes. The method is independent of the spectra of the measured radioelement. we will describe in the second part some applications which use neutron densities measurement, neutron sources intensities and ratio of cross sections of capture of thermal neutrons. (M.B.) [French] M. Berthelot, a mon entree au ''Commissariat a l 'Energie Atomique'', m'a propose d'etudier la mesure absolue des densites neutroniques. Tres rapidement le probleme de l'activite absolue des sources beta est devenu l'objet central de ce travail. Dans une premiere partie, on abordera les methodes de determination absolue des activites beta. L'utilisation d'un compteur 4{pi} permet d 'obtenir l'activite absolue de toute source radioactive beta, susceptible d'etre mise sous forme de feuille mince et de periode superieure a quelques minutes. La methode est independante du spectre du radioelement mesure. On decrira dans la seconde partie quelques applications a des mesures de densites neutroniques, d'intensites de sources de neutrons et de rapport de sections efficaces de capture de neutrons thermiques. (M.B.)

  16. Absolute calibration of TFTR helium proportional counters

    International Nuclear Information System (INIS)

    Strachan, J.D.; Diesso, M.; Jassby, D.; Johnson, L.; McCauley, S.; Munsat, T.; Roquemore, A.L.; Loughlin, M.

    1995-06-01

    The TFTR helium proportional counters are located in the central five (5) channels of the TFTR multichannel neutron collimator. These detectors were absolutely calibrated using a 14 MeV neutron generator positioned at the horizontal midplane of the TFTR vacuum vessel. The neutron generator position was scanned in centimeter steps to determine the collimator aperture width to 14 MeV neutrons and the absolute sensitivity of each channel. Neutron profiles were measured for TFTR plasmas with time resolution between 5 msec and 50 msec depending upon count rates. The He detectors were used to measure the burnup of 1 MeV tritons in deuterium plasmas, the transport of tritium in trace tritium experiments, and the residual tritium levels in plasmas following 50:50 DT experiments

  17. Neutron flux uncertainty and covariances for spectrum adjustment and estimation of WWER-1000 pressure vessel fluences

    International Nuclear Information System (INIS)

    Boehmer, Bertram

    2000-01-01

    Results of estimation of the covariance matrix of the neutron spectrum in the WWER-1000 reactor cavity and pressure vessel positions are presented. Two-dimensional calculations with the discrete ordinates transport code DORT in r-theta and r-z-geometry used to determine the neutron group spectrum covariances including gross-correlations between interesting positions. The new Russian ABBN-93 data set and CONSYST code used to supply all transport calculations with group neutron data. All possible sources of uncertainties namely caused by the neutron gross sections, fission sources, geometrical dimensions and material densities considered, whereas the uncertainty of the calculation method was considered negligible in view of the available precision of Monte Carlo simulation used for more precise evaluation of the neutron fluence. (Authors)

  18. Absolute nuclear material assay

    Science.gov (United States)

    Prasad, Manoj K [Pleasanton, CA; Snyderman, Neal J [Berkeley, CA; Rowland, Mark S [Alamo, CA

    2010-07-13

    A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

  19. Reconstruction of the neutron spectrum using an artificial neural network in CPU and GPU

    International Nuclear Information System (INIS)

    Hernandez D, V. M.; Moreno M, A.; Ortiz L, M. A.; Vega C, H. R.; Alonso M, O. E.

    2016-10-01

    The increase in computing power in personal computers has been increasing, computers now have several processors in the CPU and in addition multiple CUDA cores in the graphics processing unit (GPU); both systems can be used individually or combined to perform scientific computation without resorting to processor or supercomputing arrangements. The Bonner sphere spectrometer is the most commonly used multi-element system for neutron detection purposes and its associated spectrum. Each sphere-detector combination gives a particular response that depends on the energy of the neutrons, and the total set of these responses is known like the responses matrix Rφ(E). Thus, the counting rates obtained with each sphere and the neutron spectrum is related to the Fredholm equation in its discrete version. For the reconstruction of the spectrum has a system of poorly conditioned equations with an infinite number of solutions and to find the appropriate solution, it has been proposed the use of artificial intelligence through neural networks with different platforms CPU and GPU. (Author)

  20. Procedure to measure the neutrons spectrum around a lineal accelerator for radiotherapy

    International Nuclear Information System (INIS)

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

    2013-10-01

    An experimental procedure was developed, by means of Bonner spheres, to measure the neutrons spectrum around Linacs of medical use that only requires of a single shot of the accelerator; to this procedure we denominate Planetary or Isocentric method. One of the problems associated to the neutrons spectrum measurement in a radiotherapy room with lineal accelerator is because inside the room a mixed, intense and pulsed radiation field takes place affecting the detection systems based on active detector; this situation is solved using a passive detector. In the case of the Bonner spheres spectrometer the active detector has been substituted by activation detectors, trace detectors or thermoluminescent dosimeters. This spectrometer uses several spheres that are situated one at a time in the measurement point, this way to have the complete measurements group the accelerator should be operated, under the same conditions, so many times like spheres have the spectrometer, this activity can consume a long time and in occasions due to the work load of Linac to complicate the measurement process too. The procedure developed in this work consisted on to situate all the spectrometer spheres at the same time and to make the reading by means of a single shot, to be able to apply this procedure, is necessary that before the measurements two characteristics are evaluated: the cross-talking of the spheres and the symmetry conditions of the neutron field. This method has been applied to determine the photo-neutrons spectrum produced by a lineal accelerator of medical use Varian ix of 15 MV to 100 cm of the isocenter located to 5 cm of depth of a solid water mannequin of 30 x 30 x 15 cm. The spectrum was used to determine the total flow and the environmental dose equivalent. (Author)

  1. Measurement of the energy spectrum of the neutrons inside the neutron flux trap assembled in the center of the reactor core IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto Credidio; Santos, Diogo Feliciano dos; Jerez, Rogerio; Mura, Luis Felipe Liamos, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    This paper presents the neutron energy spectrum in the central position of a neutron flux trap assembled in the core center of the research nuclear reactor IPEN/MB-01 obtained by an unfolding method. To this end, have been used several different types of activation foils (Au, Sc, Ti, Ni, and plates) which have been irradiated in the central position of the reactor core (setting number 203) at a reactor power level of 64.57 ±2.91 watts . The activation foils were counted by solid-state detector HPGe (gamma spectrometry). The experimental data of nuclear reaction rates (saturated activity per target nucleus) and a neutron spectrum estimated by a reactor physics computer code are the main input data to get the most suitable neutron spectrum in the irradiation position obtained through SANDBP code: a neutron spectra unfolding code that use an iterative adjustment method. The adjustment resulted in 3.85 ± 0.14 10{sup 9} n cm{sup -2} s{sup -1} for the integral neutron flux, 2.41 ± 0.01 10{sup 9} n cm{sup -2} s{sup -1} for the thermal neutron flux, 1.09 ± 0.02 10{sup 9} n cm{sup -2} s{sup -1} for intermediate neutron flux and 3.41± 0.02 10{sup 8} n cm{sup -2} s{sup -1} for the fast neutrons flux. These results can be used to verify and validate the nuclear reactor codes and its associated nuclear data libraries, besides show how much is effective the use of a neutron flux trap in the nuclear reactor core to increase the thermal neutron flux without increase the operation reactor power level. The thermal neutral flux increased 4.04 ± 0.21 times compared with the standard configuration of the reactor core. (author)

  2. Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

    International Nuclear Information System (INIS)

    Brown, N.R.; Powers, J.J.; Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G.; Todosow, M.; Worrall, A.; Gehin, J.C.; Kim, T.K.; Taiwo, T.A.

    2015-01-01

    Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems

  3. Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.R., E-mail: nbrown@bnl.gov [Brookhaven National Laboratory, Upton, NY (United States); Powers, J.J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G. [Argonne National Laboratory, Argonne, IL (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States); Worrall, A.; Gehin, J.C. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Kim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States)

    2015-08-15

    Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10{sup 5} eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

  4. Calculating the energy spectrum of neutrons from tritium target of the NG-150 type generator

    International Nuclear Information System (INIS)

    Bortash, A.I.; Kuznetsov, V.S.

    1987-01-01

    Calculation procedure of neutron spectra yielding from the NG-150 generator target chamber with regard to deutron moderation is suggested. Using the suggested procedure, neutron spectra for different escape angles formed in the tritium target are calculated. The spectrum of neutrons scattered in cooling water is calculated. The mean energy of neutrons escaping at the angle of 0 deg equalling 14.5 MeV is obtained

  5. Measurements of integral cross sections in the californium-252 fission neutron spectrum

    International Nuclear Information System (INIS)

    Alberts, W.G.; Guenther, E.; Matzke, M.; Rassl, G.

    1977-01-01

    In a low-scattering arrangement cross sections averaged over the californium-252 spontaneous fission neutron spectrum were measured. The reactions 27 Al(n,α) 46 Ti, 47 Ti, 48 Ti(n,p), 54 Fe, 56 Fe(n,p), 58 Ni(n,p), 64 Zn(n,p), 115 In(n,n') were studied in order to obtain a consistent set of threshold detectors used in fast neutron flux density measurements. Overall uncertainties between 2 and 2.5% could be achieved; corrections due to neutron scattering in source and samples are discussed

  6. Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

    Science.gov (United States)

    Hosseini, Seyed Abolfazl; Esmaili Paeen Afrakoti, Iman

    2018-01-17

    The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were

  7. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  8. Planetary method to measure the neutrons spectrum in lineal accelerators of medical use

    International Nuclear Information System (INIS)

    Vega C, H. R.; Benites R, J. L.

    2014-08-01

    A novel procedure to measure the neutrons spectrum originated in a lineal accelerator of medical use has been developed. The method uses a passive spectrometer of Bonner spheres. The main advantage of the method is that only requires of a single shot of the accelerator. When this is used around a lineal accelerator is necessary to operate it under the same conditions so many times like the spheres that contain the spectrometer, activity that consumes enough time. The developed procedure consists on situating all the spheres of the spectrometer at the same time and to realize the reading making a single shot. With this method the photo neutrons spectrum produced by a lineal accelerator Varian ix of 15 MV to 100 cm of the isocenter was determined, with the spectrum is determined the total flow and the ambient dose equivalent. (Author)

  9. Neutron metrology in the L.F.R. Neutron flux density spectrum in the inner graphite reflector of the L.F.R

    International Nuclear Information System (INIS)

    Zsolnay, E.M.

    1979-01-01

    The neutron spectrum in the vertical central plug of the Low Flux Reactor has been determined experimentally. Sets of activation and fission detectors have been irradiated, and the neutron spectrum has been unfolded with aid of 3 special computer programs SAND-II, RFSP-JUEL and CRYSTAL BALL. Using these 3 programs calculations are made on the improvement ratio, which is defined as the ratio of the variance of the input flux density to that of the output flux density. A Monte Carlo error analysis is made to examine the quality of the 3 solution spectra. The results obtained with the different computer codes were compared, and showed a general agreement. The experiment confirmed that the shape of the spectrum in the intermediate energy region is near the 1/E pattern. (author)

  10. Absolute measurement of {beta} activities and application to the determination of neutronic densities; Mesure absolue d'activites {beta} et application a la determination des densites neutronique

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, R. [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1951-01-15

    M. Berthelot, to my entrance to the ''Commissariat a l 'Energie Atomique'', proposed me to study the absolute measurement of neutron densities. Very quickly the problem of the absolute activity of {beta} sources became the central object of this work. In a first part, we will develop the methods of absolute determination for {beta} activities. The use of a 4{pi} counter permits to get the absolute activity of all beta radioactive source, susceptible to be put as thin leaf and of period superior than some minutes. The method is independent of the spectra of the measured radioelement. we will describe in the second part some applications which use neutron densities measurement, neutron sources intensities and ratio of cross sections of capture of thermal neutrons. (M.B.) [French] M. Berthelot, a mon entree au ''Commissariat a l 'Energie Atomique'', m'a propose d'etudier la mesure absolue des densites neutroniques. Tres rapidement le probleme de l'activite absolue des sources beta est devenu l'objet central de ce travail. Dans une premiere partie, on abordera les methodes de determination absolue des activites beta. L'utilisation d'un compteur 4{pi} permet d 'obtenir l'activite absolue de toute source radioactive beta, susceptible d'etre mise sous forme de feuille mince et de periode superieure a quelques minutes. La methode est independante du spectre du radioelement mesure. On decrira dans la seconde partie quelques applications a des mesures de densites neutroniques, d'intensites de sources de neutrons et de rapport de sections efficaces de capture de neutrons thermiques. (M.B.)

  11. Consolidation of the neutron spectrum in the RA-6 reactor

    International Nuclear Information System (INIS)

    Bazzana, S.; Chiaraviglio, N.

    2013-01-01

    Unfolding procedures can be used to determine the neutron or gamma spectrum in a multigroup structure from experimental and calculation results. In this way, it is possible to adjust with high reliability magnitudes that cannot be directly measured. For neutron unfolding it is necessary the use of a set of detectors with different energetic response. In this work we describe two unfolding experiences in different positions of the RA-6 reactor of the Bariloche Atomic Centre. One of them consisted in the unfolding in an incore position and the other one in the BNCT facility beam.Experimental techniques and neutron detectors for each experience are described along with the correction factors that must be taken into account for each experience. In both cases there is good agreement between measured and adjusted quantities. (author) [es

  12. A neutron detector for measurement of total neutron production cross sections

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  13. First measurement of the VESUVIO neutron spectrum in the 30-80 MeV energy range using a Proton Recoil Telescope technique

    Science.gov (United States)

    Cazzaniga, C.; Tardocchi, M.; Croci, G.; Frost, C.; Giacomelli, L.; Grosso, G.; Hjalmarsson, A.; Rebai, M.; Rhodes, N. J.; Schooneveld, E. M.; Gorini, G.

    2013-11-01

    Measurements of the fast neutron energy spectrum at the ISIS spallation source are reported. The measurements were performed with a Proton Recoil Telescope consisting of a thin plastic foil placed in the neutron beam and two scintillator detectors. Results in the neutron energy range 30 MeV < En < 80 MeV are in good agreement with Monte Carlo simulations of the neutron spectrum.

  14. Study of muon-induced neutron production using accelerator muon beam at CERN

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Y.; Lin, C. J.; Ochoa-Ricoux, J. P. [Lawrence Berkeley National Laboratory, Berkeley, California (United States); Draeger, E.; White, C. G. [Illinois Institute of Technology, Chicago, Illinois (United States); Luk, K. B.; Steiner, H. [Lawrence Berkeley National Laboratory, Berkeley, California (United States); Department of Physics, University of California, Berkeley, California (United States)

    2015-08-17

    Cosmogenic muon-induced neutrons are one of the most problematic backgrounds for various underground experiments for rare event searches. In order to accurately understand such backgrounds, experimental data with high-statistics and well-controlled systematics is essential. We performed a test experiment to measure muon-induced neutron production yield and energy spectrum using a high-energy accelerator muon beam at CERN. We successfully observed neutrons from 160 GeV/c muon interaction on lead, and measured kinetic energy distributions for various production angles. Works towards evaluation of absolute neutron production yield is underway. This work also demonstrates that the setup is feasible for a future large-scale experiment for more comprehensive study of muon-induced neutron production.

  15. Neutron energy spectrum from 120 GeV protons on a thick copper target

    Energy Technology Data Exchange (ETDEWEB)

    Shigyo, Nobuhiro; /Kyushu U.; Sanami, Toshiya; /KEK, Tsukuba; Kajimoto, Tsuyoshi; /Kyushu U.; Iwamoto, Yosuke; /JAEA, Ibaraki; Hagiwara, Masayuki; Saito, Kiwamu; /KEK, Tsukuba; Ishibashi, Kenji; /Kyushu U.; Nakashima, Hiroshi; Sakamoto, Yukio; /JAEA, Ibaraki; Lee, Hee-Seock; /Pohang Accelerator Lab.; Ramberg, Erik; /Fermilab

    2010-08-01

    Neutron energy spectrum from 120 GeV protons on a thick copper target was measured at the Meson Test Beam Facility (MTBF) at Fermi National Accelerator Laboratory. The data allows for evaluation of neutron production process implemented in theoretical simulation codes. It also helps exploring the reasons for some disagreement between calculation results and shielding benchmark data taken at high energy accelerator facilities, since it is evaluated separately from neutron transport. The experiment was carried out using a 120 GeV proton beam of 3E5 protons/spill. Since the spill duration was 4 seconds, protoninduced events were counted pulse by pulse. The intensity was maintained using diffusers and collimators installed in the beam line to MTBF. The protons hit a copper block target the size of which is 5cm x 5cm x 60 cm long. The neutrons produced in the target were measured using NE213 liquid scintillator detectors, placed about 5.5 m away from the target at 30{sup o} and 5 m 90{sup o} with respect to the proton beam axis. The neutron energy was determined by time-of-flight technique using timing difference between the NE213 and a plastic scintillator located just before the target. Neutron detection efficiency of NE213 was determined on basis of experimental data from the high energy neutron beam line at Los Alamos National Laboratory. The neutron spectrum was compared with the results of multiparticle transport codes to validate the implemented theoretical models. The apparatus would be applied to future measurements to obtain a systematic data set for secondary particle production on various target materials.

  16. 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

  17. 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

  18. Fusion spectrum neutron source computation in "6LiD convertor for HFETR

    International Nuclear Information System (INIS)

    Sun Shouhua; Hu Yifei; Ye Bin

    2014-01-01

    A computation model of 14 MeV neutron from the "6LiD convertor has been established, the 14 MeV neutron sources and flux in the irradiation samples from the "6LiD convertor and the core have been computed separately, the neutron spectrum in the irradiation samples have been computed, too. The results show that the neutron sources that over 13 MeV account for 1 MeV above in the "6LiD convertor is 25.7%, 24.6% respectively, 14 MeV neutron sources get 4.31 × 10"1"3 n_T·s"-"1, 3.34 × 10"1"3 n_T·s"-"1, 14 MeV neutron flux get 2.66 × 10"1"0 n_T·cm"-"2·s"-"1, 3.53 × 10"1"0 n_T·cm"-"2·s"-"1, as He and H_2O charged in the irradiation capsule. (authors)

  19. Correction Factor Analysis Of Foil Activation And The Effect Of Neglecting The Correction On Neutron Flux And Spectrum Measurement; ANALISIS FAKTOR KOREKSI KEPING AKTIVASI DAN PENGARUH PENGABAIANNYA PADA PENGUKURAN FLUKS DAN SPEKTRUM NEUTRON

    Energy Technology Data Exchange (ETDEWEB)

    Radiyanti, Ita Budi; Hamzah, Amir; Pinem, Surian [Multipurpose Reactor Centre Indonesia, Serpong, (Indonesia)

    1996-04-15

    Foil activation method is commonly used in flux and neutron spectrum measurement in nuclear reactor and other research. The effect of the thickness, type of foil material and neutron spectrum shape on the self shielding correction and activities correction on the edges of the foil have been analyzed. Also the effect of neglecting those correction factors on neutron flux and spectrum measurement were analyzed. The calculation of the correction factor has been done by using the program which had been verified for several foils. The foils used are Au, In. Cu, Co and Dy of 0.00254 cm -0.127 cm thickness and 1.27 cm diameter. The result showed that the correction factor foils were not similar due to the variation of activation cross section and neutron spectrum shape. For the neutron spectrum in RS-2 multi purpose reactor GAS using foils of 0.00254 cm thick. The effect of neglecting correction factor on thermal flux measurement for Au, In, Co and Cu were less than -6%, for Dy was about -25%. On epithermal flux measurement for Au and In were about -60%, Co and Dy was -12% and -6%, for Cu less than -2%. The effect of neglecting correction factor on spectrum measurement was the change on the neutron flux density values along neutron energy region.

  20. Absolutely Continuous Spectrum for Random Schrödinger Operators on the Fibonacci and Similar Tree-strips

    Energy Technology Data Exchange (ETDEWEB)

    Sadel, Christian, E-mail: Christian.Sadel@ist.ac.at [University of British Columbia, Mathematics Department (Canada)

    2014-12-15

    We consider cross products of finite graphs with a class of trees that have arbitrarily but finitely long line segments, such as the Fibonacci tree. Such cross products are called tree-strips. We prove that for small disorder random Schrödinger operators on such tree-strips have purely absolutely continuous spectrum in a certain set.

  1. A neutron detector for measurement of total neutron production cross sections

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

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

  2. First measurement of the VESUVIO neutron spectrum in the 30–80 MeV energy range using a Proton Recoil Telescope technique

    International Nuclear Information System (INIS)

    Cazzaniga, C; Tardocchi, M; Croci, G; Grosso, G; Rebai, M; Gorini, G; Frost, C; Rhodes, N J; Schooneveld, E M; Giacomelli, L; Hjalmarsson, A

    2013-01-01

    Measurements of the fast neutron energy spectrum at the ISIS spallation source are reported. The measurements were performed with a Proton Recoil Telescope consisting of a thin plastic foil placed in the neutron beam and two scintillator detectors. Results in the neutron energy range 30 MeV n < 80 MeV are in good agreement with Monte Carlo simulations of the neutron spectrum

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  4. Transformation of a wave energy spectrum from encounter to absolute domain when observing from an advancing ship

    DEFF Research Database (Denmark)

    Nielsen, Ulrik Dam

    2017-01-01

    directly in the encounter domain. The encounter domain is that observed from a ship when it advances in a seaway, whereas the absolute domain is that corresponding to making observations from a fixed point in the inertial frame. Spectrum transformation can be uniquely carried out if the ship sails ”against...

  5. Feasibility study of a neutron activation system for EU test blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Tian, Kuo, E-mail: kuo.tian@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Calderoni, Pattrick [Fusion for Energy(F4E), Barcelona (Spain); Ghidersa, Bradut-Eugen; Klix, Axel [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany)

    2016-11-01

    Highlights: • This paper summarizes the technical baseline and preliminary design of EU TBM Neutron Activation System, briefly describes the key components, and outlines the major integration challenges. - Abstract: The Neutron Activation System (NAS) for the EU Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Systems (TBSs) is an instrument that is proposed to determine the absolute neutron fluence and absolute neutron flux with information on the neutron spectrum in selected positions of the corresponding Test Blanket Modules (TBMs). In the NAS activation probes are exposed to the ITER neutron flux for periods ranging from several tens of seconds up to a full plasma pulse length, and the induced gamma activities are subsequently measured. The NAS is composed of a pneumatic transfer system and a counting station. The pneumatic transfer system includes irradiation ends in TBMs, transfer pipes, return gas pipes, a transfer station with a distributor (carousel), and a pressurized gas driving system, while the counting station consists of gamma ray detectors, signal processing electronic devices, and data analyzing software for neutron source strength evaluation. In this paper, a brief description on the proposed TBM NAS as well as the key components is presented, and the integration challenges of TBM NAS are outlined.

  6. Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor

    International Nuclear Information System (INIS)

    Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.

    2008-01-01

    Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)

  7. Neutron spectrum at 900 from 800 MeV (p,n) reactions on a Ta target

    International Nuclear Information System (INIS)

    Howe, S.D.; Lisowski, P.W.; King, N.S.P.; Russell, G.J.; Donnert, H.J.

    1979-01-01

    The neutron time-of-flight spectrum produced by a thick tantalum target bombarded by 800-MeV protons was measured at an angle of 90 0 . The data were taken at the Weapons Neutron Research facility by use of a cylindrical Ta target with a radius of 1.27 cm and a length of 15 cm. An NE-213 liquid scintillator was used to detect the neutrons over an energy range of 0.5 to 350 MeV. The neutron yield is presented and compared to a intranuclear-cascade/evaporation model prediction. 3 figures

  8. Evaluation of the transmutation of transuranic using neutrons spectrum from the spallation reaction

    Energy Technology Data Exchange (ETDEWEB)

    Gilberti, Mauricio; Pereira, Claubia, E-mail: mgilber@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Angra dos Reis, RJ (Brazil); Veloso, Maria A. Fortini, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizante, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    The transmutation of transuranic was analyzed by simulating the neutron flux from different spallation sources across arrays of fissile material with isotopic composition PWR reprocessing. A simplified model of Accelerator-Driven Systems (ADS) containing target, moderator graphite, lead-bismuth coolant or sodium coolant, is used. The simulation was made using the particles transport code MCNPX 2.6.0 which allowed to evaluate the rate of transmutation of actinides (Np, Pu, Am, and Cm) at different locations in the system. The objective of the study is to evaluate which the behavior and influences the spectrum of the spallation in the transmutation without the contribution or interference of multiplier, medium subcritical, which would add the contribution of fission neutrons generated, thus interfering in the analysis. The arrangement enable to infer the influence of hardened neutron flux from the spallation reaction in the transmutation, the results show that this is independent of the target material chosen, and the spectrum of spallation has a negligible importance compared to the influence of moderation and scattering generated by the coolant or moderator used. (author)

  9. Evaluation of the transmutation of transuranic using neutrons spectrum from the spallation reaction

    International Nuclear Information System (INIS)

    Gilberti, Mauricio; Pereira, Claubia; Veloso, Maria A. Fortini

    2013-01-01

    The transmutation of transuranic was analyzed by simulating the neutron flux from different spallation sources across arrays of fissile material with isotopic composition PWR reprocessing. A simplified model of Accelerator-Driven Systems (ADS) containing target, moderator graphite, lead-bismuth coolant or sodium coolant, is used. The simulation was made using the particles transport code MCNPX 2.6.0 which allowed to evaluate the rate of transmutation of actinides (Np, Pu, Am, and Cm) at different locations in the system. The objective of the study is to evaluate which the behavior and influences the spectrum of the spallation in the transmutation without the contribution or interference of multiplier, medium subcritical, which would add the contribution of fission neutrons generated, thus interfering in the analysis. The arrangement enable to infer the influence of hardened neutron flux from the spallation reaction in the transmutation, the results show that this is independent of the target material chosen, and the spectrum of spallation has a negligible importance compared to the influence of moderation and scattering generated by the coolant or moderator used. (author)

  10. Differential and integral comparisons of three representations of the prompt neutron spectrum for the spontaneous fission of 252Cf

    International Nuclear Information System (INIS)

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

    1984-01-01

    Because of their importance as neutron standards, we present comparisons of measured and calculated prompt fission neutron spectra N(E) and average prompt neutron multiplicities anti nu/sub p/ for the spontaneous fission of 252 Cf. In particular, we test three representations of N(E) against recent experimental measurements of the differential spectrum and threshold integral cross sections. These representations are the Maxwellian spectrum, the NBS spectrum, and the Los Alamos spectrum of Madland and Nix. For the Maxwellian spectrum, we obtain the value of the Maxwellian temperature T/sub M/ by a least-squares adjustment to the experimental differential spectrum of Poenitz and Tamura. For the Los Alamos spectrum, a similar least-squares adjustment determines the nuclear level-density parameter a, which is the single unknown parameter that appears. The NBS spectrum has been previously constructed by adjustments to eight differential spectra measured during the period 1965 to 1974. Among these three representations, we find that the Los Alamos spectrum best reproduces both the differential and integral measurements, assuming ENDF/B-V cross sections in the calculation of the latter. Although the NBS spectrum reproduces the integral measurements fairly well, it fails to satisfactorily reproduce the new differential measurement, and the Maxwellian spectrum fails to satisfactorily reproduce the integral measurements. Additionally, we calculate a value of anti nu/sub p/ from the Los Alamos theory that is within approximately 1% of experiment. 25 references

  11. Investigation of the influence of the neutron spectrum in determinations of integral cross-section ratios

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.

    1987-11-01

    Ratio measurements are routinely employed in studies of neutron interaction processes in order to generate new differential cross-section data or to test existing differential cross-section information through examination of the corresponding response in integral neutron spectra. Interpretation of such data requires that careful attention be given to details of the neutron spectra involved in these measurements. Two specific tasks are undertaken in the present investigation: (1) Using perturbation theory, a formula is derived which permits one to relate the ratio measured in a realistic quasimonoenergetic spectrum to the desired pure monoenergetic ratio. This expression involves only the lowest-order moments of the neutron energy distribution and corresponding parameters which serve to characterize the energy dependence of the differential cross sections, quantities which can generally be estimated with reasonable precision from the uncorrected data or from auxiliary information. (2) Using covariance methods, a general formalism is developed for calculating the uncertainty of a measured integral cross-section ratio which involves an arbitrary neutron spectrum. This formalism is employed to further examine the conditions which influence the sensitivity of such measured ratios to details of the neutron spectra and to their uncertainties. Several numerical examples are presented in this report in order to illustrate these principles, and some general conclusion are drawn concerning the development and testing of neutron cross-section data by means of ratio experiments. 16 refs., 1 fig., 4 tabs.

  12. Investigation of the influence of the neutron spectrum in determinations of integral cross-section ratios

    International Nuclear Information System (INIS)

    Smith, D.L.

    1987-11-01

    Ratio measurements are routinely employed in studies of neutron interaction processes in order to generate new differential cross-section data or to test existing differential cross-section information through examination of the corresponding response in integral neutron spectra. Interpretation of such data requires that careful attention be given to details of the neutron spectra involved in these measurements. Two specific tasks are undertaken in the present investigation: (1) Using perturbation theory, a formula is derived which permits one to relate the ratio measured in a realistic quasimonoenergetic spectrum to the desired pure monoenergetic ratio. This expression involves only the lowest-order moments of the neutron energy distribution and corresponding parameters which serve to characterize the energy dependence of the differential cross sections, quantities which can generally be estimated with reasonable precision from the uncorrected data or from auxiliary information. (2) Using covariance methods, a general formalism is developed for calculating the uncertainty of a measured integral cross-section ratio which involves an arbitrary neutron spectrum. This formalism is employed to further examine the conditions which influence the sensitivity of such measured ratios to details of the neutron spectra and to their uncertainties. Several numerical examples are presented in this report in order to illustrate these principles, and some general conclusion are drawn concerning the development and testing of neutron cross-section data by means of ratio experiments. 16 refs., 1 fig., 4 tabs

  13. Measurements of the neutron spectrum in transit to Mars on the Mars Science Laboratory.

    Science.gov (United States)

    Köhler, J; Ehresmann, B; Zeitlin, C; Wimmer-Schweingruber, R F; Hassler, D M; Reitz, G; Brinza, D E; Appel, J; Böttcher, S; Böhm, E; Burmeister, S; Guo, J; Lohf, H; Martin, C; Posner, A; Rafkin, S

    2015-04-01

    The Mars Science Laboratory spacecraft, containing the Curiosity rover, was launched to Mars on 26 November 2011. Although designed for measuring the radiation on the surface of Mars, the Radiation Assessment Detector (RAD) measured the radiation environment inside the spacecraft during most of the 253-day, 560-million-kilometer cruise to Mars. An important factor for determining the biological impact of the radiation environment inside the spacecraft is the specific contribution of neutrons with their high biological effectiveness. We apply an inversion method (based on a maximum-likelihood estimation) to calculate the neutron and gamma spectra from the RAD neutral particle measurements. The measured neutron spectrum (12-436 MeV) translates into a radiation dose rate of 3.8±1.2 μGy/day and a dose equivalent of 19±5 μSv/day. Extrapolating the measured spectrum (0.1-1000 MeV), we find that the total neutron-induced dose rate is 6±2 μGy/day and the dose equivalent rate is 30±10 μSv/day. For a 360 day round-trip from Earth to Mars with comparable shielding, this translates into a neutron induced dose equivalent of about 11±4 mSv. Copyright © 2015 The Committee on Space Research (COSPAR). Published by Elsevier Ltd. All rights reserved.

  14. Teratogenic and embryolethal effects in mice of fission-spectrum neutrons and γ-rays

    International Nuclear Information System (INIS)

    Cairnie, A.B.; Grahn, D.; Rayburn, H.B.; Williamson, F.S.; Brown, R.J.

    1974-01-01

    Fission-spectrum neutrons from the Janus reactor at Argonne National Laboratory were compared with γ-rays in terms of their relative biological effectiveness (RBE) for embryolethal and teratogenic effects in mice. No evidence was found of any processes that were abnormally sensitive to neutrons. The RBE for killing embryos and producing abnormal embryos or specific abnormalities was between 2 and 3. This is close to the values found in other systems for processes involving cell killing. (U.S.)

  15. A genetic algorithm based method for neutron spectrum unfolding

    International Nuclear Information System (INIS)

    Suman, Vitisha; Sarkar, P.K.

    2013-03-01

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

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

    Science.gov (United States)

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

    2004-01-01

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

  17. The 235U prompt fission neutron spectrum measured by the Chi-Nu project at LANSCE

    Directory of Open Access Journals (Sweden)

    Gomez J.A.

    2017-01-01

    Full Text Available The Chi-Nu experiment aims to accurately measure the prompt fission neutron spectrum for the major actinides. At the Los Alamos Neutron Science Center (LANSCE, fission can be induced with neutrons ranging from 0.7 MeV and above. Using a two arm time-of-flight (TOF technique, the fission neutrons are measured in one of two arrays: a 22-6Li glass array for lower energies, or a 54-liquid scintillator array for outgoing energies of 0.5 MeV and greater. Presented here are the collaboration's preliminary efforts at measuring the 235U PFNS.

  18. The Prompt Fission Neutron Spectrum of 235U for Einc 0.7-5.0 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, Jaime A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Devlin, Matthew James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haight, Robert Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); O' Donnell, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lee, Hye Young [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taddeucci, Terry Nicholas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelly, Keegan John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fotiadis, Nikolaos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Neudecker, Denise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Talou, Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Solomon, Clell Jeffrey Jr. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wu, Ching-Yen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bucher, Brian Michael [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Buckner, Matthew Quinn [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Henderson, Roger Alan [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-03-23

    The Chi-Nu experiment aims to accurately measure the prompt fission neutron spectrum (PFNS) for the major actinides. At the Los Alamos Neutron Science Center (LANSCE), fission can be induced using the white neutron source. Using a two arm time of flight (T.O.F) technique; Chi-Nu presents a preliminary result of the low energy component of the 235U PFNS measured using an array of 22-Lithium glass scintillators.

  19. Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki, E-mail: koyanagit@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kumar, N.A.P. Kiran [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hwang, Taehyun [Tohoku University, Sendai, 980-8579 (Japan); Garrison, Lauren M.; Hu, Xunxiang [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2017-07-15

    Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

  20. Tritium solid targets for intense D-T neutron production and its related problems

    International Nuclear Information System (INIS)

    Sumita, Kenji

    1988-01-01

    This review paper is divided into three parts. Firstly, to attain an intense neutron production rate, the construction of a design with a higher tritium-containing surface and an effective cooling system like a rotating target device are discussed. The maximum attainable intensity based on tritium solid targets shall be estimated regarding planning for future D-T sources. Secondly, on the way to carry out some experiments, an absolute intensity calibration and an angular dependent neutron energy spectrum of the neutron source are essential parameters to analyse the results of the experiments. Sometimes the space dependent neutron spectrum is required as well as the space dependent neutron flux near the targets and irradiation samples. The measurement methods and their examples are reviewed for tritium solid targets. The third part is devoted to discuss the protection to tritium contamination problems due to unavoidable release of tritium gas from targets. Performance and effectiveness of tritium collection systems for intense D-T neutron sources shall be discussed in some examples. Tritium contamination incidents due to the faulted film powder of target surface are also reported in some real incident cases. (author). Abstract only

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-11

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  4. Experimental measurement of neutron spectrum in the reflector of a light water reactor; Determination experimentale du spectre des neutrons dans le reflecteur d'une pile a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Brethe, P [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1963-09-15

    1. Thermal neutrons: The temperature of the thermal neutron spectrum was calculated using Au-Lu foils. This temperature varies from 300 deg. K (temperature of the moderator) at 30 cm of the core to 350 deg. K in a hole of the core. 2. Slowing down of neutron: Four resonance detectors have been used (Au, In, Co, Mn). We can write a 1/E form of the spectrum. The linking up energy E{sub M} between thermal neutron spectrum and slowing down spectrum is about 0.23 eV and is free from the Maxwell spectrum temperature. The decrease of slowing down flux regarding thermal flux, farther from the core, has been showed. 3. Fast neutrons: We used 3 threshold detectors (Ni, Al, Mg). We supposed a E{sup 1/2} e{sup -{beta}}{sup E} from of the spectrum above 3 MeV. The values of {beta} are in a range from 0.775, at the centre of the core and in a loop-hole, to 0,64 at about 30 cm of the core. 4. Continuous shape of the spectrum: The following interpolations give useful informations between the field where measurements have been made: between 340 eV and 10 keV: 1/E form between 10 keV and 330 keV: 1/(E {sigma}{sub S}(E)) form ({sigma}{sub S}(E) elastic scattering section on hydrogen) between 330 keV and 3 MeV: calculated form by the moments method (ref. BSR). (author) [French] 1. Neutrons thermiques: La temperature du spectre des neutrons thermiques a ete determinee par la methode (or-lutecium). Cette temperature varie de 300 deg. K (temperature du moderateur) a 30 cm du coeur, a 350 deg. K dans une encoche du coeur. 2. Neutrons en ralentissement: 4 detecteurs resonnants ont ete employes (Au, In, Co, Mn). Le spectre peut etre mis sous la forme 1/E quelle que soit la distance a la limite coeur-reflecteur. L'energie de raccordement E{sub M} entre spectre des neutrons thermiques et spectre en ralentissement est environ 0,23 eV et independante de la temperature du spectre de Maxwell. La diminution relative du flux en ralentissement par rapport au flux thermique quand la distance au coeur

  5. Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

  6. Fast neutron spectrum in the exposure room of the TRIGA Mark II reactor in Ljubljana

    International Nuclear Information System (INIS)

    Kristof, E.S.

    2003-01-01

    In this paper a description of the high energy neutrons at a usual position in the dry cell of our reactor is given. Neutrons emerging from the graphite reflector enter the exposure room through the horizontal shaft. At the irradiation position samples of detection materials were irradiated. After irradiation γ-ray spectra were measured and from the saturation activities the spectrum was calculated. (author)

  7. Sequential measurements of cosmic-ray neutron spectrum and dose rate at sea level in Sendai, Japan

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Nunomiya, Tomoya; Abe, Shigeru; Terunuma, Kazutaka; Suzuki, Hiroyuki

    2005-01-01

    The cosmic-ray neutron energy spectrum and dose rate were measured sequentially for two years from April 2001 up to March 2003 by using three neutron detectors, a 3 He-loaded multi-moderator detector (Bonner ball), 12.7 cm diameter by 12.7 cm long NE213 organic liquid scintillator, and high-sensitivity rem (dose equivalent) counter at the Kawauchi campus of Tohoku University in Sendai, Japan of geomagnetic latitude, 29degN, and cutoff rigidity, 10.43 GV. The neutron spectrum has three major peaks, thermal energy peak, evaporation peak around 1 MeV and cascade peak around 100 MeV. The ambient neutron dose equivalent rates measured by the rem counter, and the Bonner ball keep almost constant values of 4.0 and 6.5 (nSv/h), respectively, throughout this time period, after atmospheric pressure correction, and it often decreased about 30% after a large Solar Flare, that is called as the Forbush decrease. The total neutron flux was also obtained by the Bonner ball measurements to be 7.5x10 -3 (ncm -2 ·s -1 ) in average. The altitude variation of neutron flux and dose was also investigated by comparing the measured results with other results measured at Mt. Fuji area and aboard an airplane, where the cutoff rigidities are similar. (author)

  8. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.; Luis L, M. A.; Raya A, R.; Cruz G, H. S.

    2013-10-01

    This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10 -10 to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of 197 Au, 58 Ni, 115 In, 24 Mg, 27 Al, 58 Fe, 59 Co and 63 Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

  9. Measurements of europium-152 depth profile of stone embankments exposed the Nagasaki atomic bomb for neutron spectrum analysis

    International Nuclear Information System (INIS)

    Tatsumi-Miyajima, Junko; Shimasaki, Tatsuya; Okajima, Shunzo; Takada, Jitsuya; Yoshida, Masahiro; Takao, Hideaki; Okumura, Yutaka; Nakazawa, Masaharu.

    1990-01-01

    Quantitative measurement of neutron-induced radionuclide of 152 Eu in rocks near the hypocenter (ground center of the atomic bomb explosion) in Nagasaki was performed to obtain the depth profiles and calculate the neutron energy spectrum. Core samples were drilled and taken from the stone embankments on both sides of river within a radius of 500 m from the hypocenter. After cutting each core into about 27 mm-thick sections, each section was measured its gamma-ray spectrum with a pure germanium semiconductor detector and analyzed a content of natural europium by the activation method. The highest value 8.0 x 10 -2 Bq/μg of 152 Eu at the time of the blast was obtained from the surface plates of rock cores collected near the hypocenter. The surface activity of cores was reduced with increasing the slant distances from the hypocenter. The slopes of the depth profiles were similar among samples taken from the same location. In order to analyze the depth profile of 152 Eu activity in rock andesite, experiments using a fast neutron reactor and thermal neutron reactor were carried out. Comparing the measurements on the A-bomb exposure rock with the simulated results at the reactors, among the experiments, the depth profile using the neutron moderator of 10 mm polyethylene was closed to that obtained from the A-bomb exposed samples. The experiment of thermal neutron incidence only could not reproduce the profiles from the A-bomb exposed samples. This fact indicates that the depth profiles of 152 Eu in rock exposed to the A-bomb include valuable information concerning the neutron spectrum and intensity. (author)

  10. Group Representation of the Prompt Fission Neutron Spectrum of {sup 252}Cf

    Energy Technology Data Exchange (ETDEWEB)

    Croft, S.; Miller, K. A. [Safeguards Science and Technology Group (N-1), Nuclear Nonproliferation Division, Los Alamos National Laboratory, Los Alamos(United States)

    2011-12-15

    We review the spectral representation used for the prompt fission neutron spectrum of 252Cf in the International Organization for Standardization document ISO 8529-1. We find corrections to Table A.2, the discrete group structure form, of this report are needed. We describe the approach to generating replacement values and provide a new tabulation.

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

    Science.gov (United States)

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

    2016-12-01

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

  12. Gamma spectrum following neutron capture in {sup 167}Er

    Energy Technology Data Exchange (ETDEWEB)

    Visser, D.; Khoo, T.L.; Lister, C.J. [and others

    1995-08-01

    Statistical decay from a highly excited state samples all the lower-lying states and, hence, provides a sensitive measure of the level density. Pairing has a major impact on the level density, e.g. creating a pair gap between the 0- and 2-quasiparticle configurations. Hence the shape of the statistical spectrum contains information on pairing, and can be used to provide information on the reduction of pairing with thermal excitation energy. For this reason, we measured the complete spectrum of {gamma}rays following thermal neutron capture in {sup 167}Er. The experiment was performed at the Brookhaven reactor using Compton-suppressed Ge detectors from TESSA. The spectrum, which was corrected for detector response and efficiency, reveals primary (first-step, high-energy) transitions up to nearly 8 MeV, secondary (last-step, lower-energy) transitions, as we as a continuous statistical component. Effort was expanded to identify all lines from contaminant sources and an upper limit of 5% was tentatively set for their contributions. The spectral shape of the statistical spectrum will be compared with theoretical spectra obtained from a calculation of pairing which accounts for a stepwise reduction of the pair correlations as the number of quasiparticles increases. The primary lines which decay directly to the near-yrast states will also be used to deduce the level densities.

  13. Kinetic energy spectrum and polarization of neutrons from the reaction 12C(p,n)X at 590 MeV

    International Nuclear Information System (INIS)

    Arnold, J.

    1998-01-01

    The kinetic energy spectrum and the polarization of the PSI neutron beam produced in the reaction 12 C(p,n)X at 0 with 590 MeV polarized protons were investigated. A strong energy dependence of the neutron beam polarization is observed which was not expected at the time the neutron beam was built. (orig.)

  14. Spallation Neutron Spectrum on a Massive Lead/Paraffin Target Irradiated with 1 GeV Protons

    CERN Document Server

    Adam, J; Barashenkov, V S; Brandt, R; Golovatiouk, V M; Kalinnikov, V G; Katovsky, K; Krivopustov, M I; Kumar, V; Kumawat, H; Odoj, R; Pronskikh, V S; Solnyshkin, A A; Stegailov, V I; Tsoupko-Sitnikov, V M; Westmeier, W

    2004-01-01

    The spectra of gamma-ray emitted by decaying residual nuclei, produced by spallation neutrons with (n, xn), (n,xnyp), (n,p), (n,gamma) reactions in activation threshold detectors - namely, ^{209}Bi, ^{197}Au, ^{59}Co, ^{115}In, ^{232}Th, were measured in the Laboratory of Nuclear Problems (LNP), JINR, Dubna, Russia. Spallation neutrons were generated by bombarding a 20 cm long cylindrical lead target, 8 cm in diameter, surrounded by a 6 cm thick layer of paraffin moderator, with a 1 GeV proton beam from the Nuclotron accelerator. Reaction rates and spallation neutron spectrum were measured and compared with CASCADE code calculations.

  15. Determination of the absolute efficiency of an organic scintillator for neutrons with energies between 0.5 and 800 MeV

    International Nuclear Information System (INIS)

    Howe, S.D.; Lisowski, P.W.; Russell, G.J.; King, N.S.P.; Donnert, H.J.

    1984-01-01

    We have determined the absolute efficiency of an NE-213 scintillator for neutrons with energies from 0.5 to 800 MeV. The detector was 5.1 cm in diameter and 2.5 cm deep. The efficiencies were obtained for detector thresholds of 0.011, 0.48, 1.12, and 4.48 MeVee. Our results are compared to predictions of the STANTON computer code. (orig.)

  16. Electric excitations in liquid He4 and their role in neutron scattering spectrum

    International Nuclear Information System (INIS)

    Poluehktov, Yu.M.; Karnatsevich, L.V.

    2001-01-01

    Data of experiments on excitation spectrum in liquid He 4 by inelastic neutron scattering method are discussed. Exact solution of particle scattering in ideal Bose-gas is given. Influence of inter-particle interactions on the structure of many-particle Bose system is analysed qualitatively. 55 refs., 1 figs

  17. Applications of a lead pile coupled with fast reactor core of Yayoi as an intermediate energy neutron standard field

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakazawa, Masaharu; Sekiguchi, Akira; Wakabayashi, Hiroaki.

    1976-10-01

    Intermediate neutron column of YAYOI reactor is here evaluated as an intermediate energy neutron standard field which provides a base of the measurements of various reaction rates in that energy region, including detector calibration and Doppler coefficient determination. The experiments were performed using YAYOI's core as a fast neutron source by coupling with the large lead pile, which is a 160 ton's octagon of 2.5 m high and with a thickness of about 2.5 m face to face distance. Spatial variation of the neutron flux in the lead pile was estimated by gold activation foils, and the neutron spectrum by sandwich foils, a helium-3 proportional counter and a proton recoil counter. The calculated results were obtained using one and two- dimensional discrete ordinate code, ANISN and TWOTRAN II. Through comparison of experiment with calculation, it became clear that the neutron field at the central block has simple energy spectrum and stable spatial distribution of the neutron flux, the absolute of which was 5.0 x 10 4 (n/cm 2 /sec/Watt) at the representative energy of 1 KeV. The energy spectrum of the position and the spatial dependent neutron flux in the lead pile are both represented by the semiempirical formula, which must be useful both for evaluation of experimental data and for future applications. (auth.)

  18. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.

    2014-01-01

    This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au 197 , Ni 58 , In 115 , Mg 24 , Al 27 , Fe 58 , Co 59 and Cu 63 , they were selected to cover the full range the energies (10 -10 to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or research projects. (Author)

  19. Absolute instabilities of travelling wave solutions in a Keller-Segel model

    Science.gov (United States)

    Davis, P. N.; van Heijster, P.; Marangell, R.

    2017-11-01

    We investigate the spectral stability of travelling wave solutions in a Keller-Segel model of bacterial chemotaxis with a logarithmic chemosensitivity function and a constant, sublinear, and linear consumption rate. Linearising around the travelling wave solutions, we locate the essential and absolute spectrum of the associated linear operators and find that all travelling wave solutions have parts of the essential spectrum in the right half plane. However, we show that in the case of constant or sublinear consumption there exists a range of parameters such that the absolute spectrum is contained in the open left half plane and the essential spectrum can thus be weighted into the open left half plane. For the constant and sublinear consumption rate models we also determine critical parameter values for which the absolute spectrum crosses into the right half plane, indicating the onset of an absolute instability of the travelling wave solution. We observe that this crossing always occurs off of the real axis.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-21

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  3. Absolute limit on rotation of gravitationally bound stars

    Science.gov (United States)

    Glendenning, N. K.

    1994-03-01

    The authors seek an absolute limit on the rotational period for a neutron star as a function of its mass, based on the minimal constraints imposed by Einstein's theory of relativity, Le Chatelier's principle, causality, and a low-density equation of state, uncertainties which can be evaluated as to their effect on the result. This establishes a limiting curve in the mass-period plane below which no pulsar that is a neutron star can lie. For example, the minimum possible Kepler period, which is an absolute limit on rotation below which mass-shedding would occur, is 0.33 ms for a M = 1.442 solar mass neutron star (the mass of PSR1913+16). If the limit were found to be broken by any pulsar, it would signal that the confined hadronic phase of ordinary nucleons and nuclei is only metastable.

  4. A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

    Energy Technology Data Exchange (ETDEWEB)

    Waugh, C. J., E-mail: cjwaugh@mit.edu; Zylstra, A. B.; Frenje, J. A.; Séguin, F. H.; Petrasso, R. D. [Plasma Science and Fusion Center, MIT, Cambridge, Massachusetts 02139 (United States); Rosenberg, M. J.; Glebov, V. Yu.; Sangster, T. C.; Stoeckl, C. [Laboratory for Laser Energetics, Rochester, New York 14623 (United States)

    2015-05-15

    Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition, comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.

  5. A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors.

    Science.gov (United States)

    Waugh, C J; Rosenberg, M J; Zylstra, A B; Frenje, J A; Séguin, F H; Petrasso, R D; Glebov, V Yu; Sangster, T C; Stoeckl, C

    2015-05-01

    Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition, comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.

  6. A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

    International Nuclear Information System (INIS)

    Waugh, C. J.; Zylstra, A. B.; Frenje, J. A.; Séguin, F. H.; Petrasso, R. D.; Rosenberg, M. J.; Glebov, V. Yu.; Sangster, T. C.; Stoeckl, C.

    2015-01-01

    Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition, comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule

  7. Reconstruction of the neutron spectrum using an artificial neural network in CPU and GPU; Reconstruccion del espectro de neutrones usando una red neuronal artificial (RNA) en CPU y GPU

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez D, V. M.; Moreno M, A.; Ortiz L, M. A. [Universidad de Cordoba, 14002 Cordoba (Spain); Vega C, H. R.; Alonso M, O. E., E-mail: vic.mc68010@gmail.com [Universidad Autonoma de Zacatecas, 98000 Zacatecas, Zac. (Mexico)

    2016-10-15

    The increase in computing power in personal computers has been increasing, computers now have several processors in the CPU and in addition multiple CUDA cores in the graphics processing unit (GPU); both systems can be used individually or combined to perform scientific computation without resorting to processor or supercomputing arrangements. The Bonner sphere spectrometer is the most commonly used multi-element system for neutron detection purposes and its associated spectrum. Each sphere-detector combination gives a particular response that depends on the energy of the neutrons, and the total set of these responses is known like the responses matrix Rφ(E). Thus, the counting rates obtained with each sphere and the neutron spectrum is related to the Fredholm equation in its discrete version. For the reconstruction of the spectrum has a system of poorly conditioned equations with an infinite number of solutions and to find the appropriate solution, it has been proposed the use of artificial intelligence through neural networks with different platforms CPU and GPU. (Author)

  8. Neutron spectrum for neutron capture therapy in boron

    International Nuclear Information System (INIS)

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

    2016-10-01

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

  9. Study of U{sup 235} neutron fission spectrum by the knowledge of cross sections average over that spectrum; Estudio del espectro de neutrones de fision del {sup 235}U a traves del conocimiento de secciones eficaces promediadas sobre dicho espectro

    Energy Technology Data Exchange (ETDEWEB)

    Suarez, P M [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

    1998-12-31

    A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the {sup 45}Sc(n,2n) {sup 44g}Sc and {sup 45}Sc(n,2n) {sup 44m}Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, {sup 235}U. The results of the calculations mentioned above lead us to propose an analytical form for the {sup 235}U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole

  10. Calculation analysis of Wims/D4-Batan-2DIFF neutron spectrum on RSG-GAS with cadmium ratio

    International Nuclear Information System (INIS)

    Radianti, I.B.; Zuhair; Hamzah, A.

    1998-01-01

    The calculation analysis of WIMS/D4-BATAN-2DIFF neutron spectrum was performed by comparison the calculation result of cadmium ratio with the experiment result on CIP, IP2, IP3 and IP4 irradiation positions of RSG GAS tenth core. The foils of Au, Mn and Co were used for determination of the measured and calculated cadmium ratios. Spectrum calculation was done in 69 energy group with 541 energy group (till 10 MeV) cross section of foil absorption reaction. The difference values between cadmium ratio calculation and experiment result for all cases were in interval of 11.0%-26.3% which are out of measurement deviation range. From these result, it concluded that the use of WIM /D4 in generating group constant is not sufficient to obtain the neutron spectrum, especially for non-fuel region

  11. Assessing neutron generator output using neutron activation of silicon

    International Nuclear Information System (INIS)

    Kehayias, Pauli M.; Kehayias, Joseph J.

    2007-01-01

    D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the 28 Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10 3 n/s/cm 2 ± 5%, which is consistent with the manufacturer's specifications

  12. Assessing neutron generator output using neutron activation of silicon

    Energy Technology Data Exchange (ETDEWEB)

    Kehayias, Pauli M. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States); Kehayias, Joseph J. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States)]. E-mail: joseph.kehayias@tufts.edu

    2007-08-15

    D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the {sup 28}Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10{sup 3} n/s/cm{sup 2} {+-} 5%, which is consistent with the manufacturer's specifications.

  13. Measuring Neutron Spectrum at MIT Research Reactor Utilizing He-3 Bonner Cylinder Approach with an Unfolding Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Leder, A. [MIT; Anderson, A. J. [Chicago U., KICP; Billard, J. [Lyon, IPN; Figueroa-Feliciano, E. [Northwestern U.; Formaggio, J. A. [MIT; Hasselkus, C. [Wisconsin U., Madison; Newman, E. [MIT; Palladino, K. [Wisconsin U., Madison; Phuthi, M. [MIT; Winslow, L. [MIT; Zhang, L. [MIT

    2017-10-02

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2e18 neutrinos/second at the core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped a Bonner cylinder around a He-3 thermal neutron detector, whose data was then unfolded to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet in the future at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.

  14. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    International Nuclear Information System (INIS)

    Cartier, J.; Casoli, P.; Chappert, F.

    2013-01-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  15. SYSTEMATIC UNCERTAINTIES IN THE SPECTROSCOPIC MEASUREMENTS OF NEUTRON STAR MASSES AND RADII FROM THERMONUCLEAR X-RAY BURSTS. III. ABSOLUTE FLUX CALIBRATION

    Energy Technology Data Exchange (ETDEWEB)

    Güver, Tolga [Istanbul University, Science Faculty, Department of Astronomy and Space Sciences, Beyazıt, 34119, Istanbul (Turkey); Özel, Feryal; Psaltis, Dimitrios [Department of Astronomy, University of Arizona, 933 N. Cherry Avenue, Tucson, AZ 85721 (United States); Marshall, Herman [Center for Space Research, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Guainazzi, Matteo [European Space Astronomy Centre of ESA, P.O. Box 78, Villanueva de la Cañada, E-28691 Madrid (Spain); Díaz-Trigo, Maria [ESO, Karl-Schwarzschild-Strasse 2, D-85748 Garching bei München (Germany)

    2016-09-20

    Many techniques for measuring neutron star radii rely on absolute flux measurements in the X-rays. As a result, one of the fundamental uncertainties in these spectroscopic measurements arises from the absolute flux calibrations of the detectors being used. Using the stable X-ray burster, GS 1826–238, and its simultaneous observations by Chandra HETG/ACIS-S and RXTE /PCA as well as by XMM-Newton EPIC-pn and RXTE /PCA, we quantify the degree of uncertainty in the flux calibration by assessing the differences between the measured fluxes during bursts. We find that the RXTE /PCA and the Chandra gratings measurements agree with each other within their formal uncertainties, increasing our confidence in these flux measurements. In contrast, XMM-Newton EPIC-pn measures 14.0 ± 0.3% less flux than the RXTE /PCA. This is consistent with the previously reported discrepancy with the flux measurements of EPIC-pn, compared with EPIC MOS1, MOS2, and ACIS-S detectors. We also show that any intrinsic time-dependent systematic uncertainty that may exist in the calibration of the satellites has already been implicity taken into account in the neutron star radius measurements.

  16. Nuclear data and measurements series: Ratio of the prompt-fission-neutron spectrum of plutonium 239 to that of uranium 235

    International Nuclear Information System (INIS)

    Sugimoto, M.; Smith, A.B.; Guenther, P.T.

    1986-09-01

    The prompt-fission-neutron spectrum resulting from 239 Pu fission induced by 0.55 MeV incident neutrons is measured from 1.0 to 10.0 MeV relative to that of 235 U fission induced by the same incident-energy neutrons. The measurements employ the time-of-flight technique. Energy-dependent ratios of the two spectra are deduced from the measured values over the energy range 1.0 to 10.0 MeV. The experimentally-derived ratio results are compared with those calculated from ENDF/B-V, revision-2, and with results of recent microscopic measurements. Using the ENDF/B-V 235 U Watt parameters for the 235 U spectrum, the experimental measurements imply a ratio of average fission-spectrum energies of 239 Pu/ 235 U = 1.045 +- 0.003, compared to the value 1.046 calculated from ENDF/B-V, revision 2. 12 refs., 2 figs., 2 tabs

  17. International intercomparison on the neutron flux density spectrum just before the REAL-80 project

    International Nuclear Information System (INIS)

    Ertek, C.

    1981-06-01

    This work briefly presents the results of the international intercomparison on the neutron flux density spectrum just before the REAL-80 intercomparison project. Some of the results of this intercomparison with a smaller number of laboratories will be also reflected in the REAL-80 project, therefore, it has some significant issues. This work is performed within the IAEA programme on standardization of reactor radiation measurements, one of the important objectives of which is the assistance of laboratories in Member States to implement or intercompare the multiple foil activation techniques for different neutron field measurements

  18. The effect of electromagnetic interactions on the proton spectrum in free neutron β-decay

    International Nuclear Information System (INIS)

    Bunatyan, G.G.

    2000-01-01

    In the β decay of an unpolarized free neutron, the effect of electromagnetic interactions on the proton recoil spectrum is studied in the light of the experiments which are carried out and planned for now. The corrections to the energy distribution of protons prove to amount to the value of a few per cent. Nowadays, this is substantial for obtaining with a high accuracy, of ∼ 1% or better, the characteristics of weak interactions by processing the data of the experiments on the proton distribution in the free neutron β-decay

  19. The GEANT4 simulation study of the characteristic γ-ray spectrum of TNT under soil induced by DT neutron

    International Nuclear Information System (INIS)

    Qin Xue; Han Jifeng; Yang Chaowen

    2014-01-01

    The characteristic γ-ray spectrum of TNT under soil induced by DT neutron is measured based on the PFTNA demining system. GEANT4 Monte Carlo simulation toolkit is used to simulate the whole experimental procedure. The simulative spectrum is compared with the experimental spectrum. The result shows that they are mainly consistent. It is for the first time to analyze the spectrum by Monte Carlo simulation, the share of the background sources such as neutron, gamma are obtained, the contribution that the experimental apparatus such as shielding, detector sleeve, moderator make to the background is analysed. The study found that the effective gamma signal (from soil and TNT) is only 29% of the full-spectrum signal, and the background signal is more than 68% of the full-spectrum signal, which is mainly produced in the shielding and the detector sleeve. The simulation result shows that by gradually improving the shielding and the cadmium of the detector sleeve, the share of the effective gamma signal can increase to 36% and the background signal can fell 7% eventually. (authors)

  20. Study on the energy response to neutrons for a new scintillating-fiber-array neutron detector

    CERN Document Server

    Zhang Qi; Wang Qun; Xie Zhong Shen

    2003-01-01

    The energy response of a new scintillating-fiber-array neutron detector to neutrons in the energy range 0.01 MeV<=E sub n<=14 MeV was modeled by combining a simplified Monte Carlo model and the MCNP 4b code. In order to test the model and get the absolute sensitivity of the detector to neutrons, one experiment was carried out for 2.5 and 14 MeV neutrons from T(p,n) sup 3 He and T(d,n) sup 4 He reactions at the Neutron Generator Laboratory at the Institute of Modern Physics, the Chinese Academy of Science. The absolute neutron fluence was obtained with a relative standard uncertainty 4.5% or 2.0% by monitoring the associated protons or sup 4 He particles, respectively. Another experiment was carried out for 0.5, 1.0, 1.5, 2.0, 2.5 MeV neutrons from T(p,n) sup 3 He reaction, and for 3.28, 3.50, 4.83, 5.74 MeV neutrons from D(d,n) sup 3 He reaction on the Model 5SDH-2 accelerator at China Institute of Atomic Energy. The absolute neutron fluence was obtained with a relative standard uncertainty 5.0% by usin...

  1. Singular Spectrum Near a Singular Point of Friedrichs Model Operators of Absolute Type

    International Nuclear Information System (INIS)

    Iakovlev, Serguei I.

    2006-01-01

    In L 2 (R) we consider a family of self adjoint operators of the Friedrichs model: A m =|t| m +V. Here |t| m is the operator of multiplication by the corresponding function of the independent variable t element of R, and (perturbation) is a trace-class integral operator with a continuous Hermitian kernel ν(t,x) satisfying some smoothness condition. These absolute type operators have one singular point of order m>0. Conditions on the kernel ν(t,x) are found guaranteeing the absence of the point spectrum and the singular continuous one of such operators near the origin. These conditions are actually necessary and sufficient. They depend on the finiteness of the rank of a perturbation operator and on the order of singularity. The sharpness of these conditions is confirmed by counterexamples

  2. Planetary method to measure the neutrons spectrum in lineal accelerators of medical use; Metodo planetario para medir el espectro de neutrones en aceleradores lineales de uso medico

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Benites R, J. L., E-mail: fermineutron@yahoo.com [Centro Estatal de Cancerologia de Nayarit, Servicio de Seguridad Radiologica, Calzada de la Cruz 118 Sur, 63000 Tepic, Nayarit (Mexico)

    2014-08-15

    A novel procedure to measure the neutrons spectrum originated in a lineal accelerator of medical use has been developed. The method uses a passive spectrometer of Bonner spheres. The main advantage of the method is that only requires of a single shot of the accelerator. When this is used around a lineal accelerator is necessary to operate it under the same conditions so many times like the spheres that contain the spectrometer, activity that consumes enough time. The developed procedure consists on situating all the spheres of the spectrometer at the same time and to realize the reading making a single shot. With this method the photo neutrons spectrum produced by a lineal accelerator Varian ix of 15 MV to 100 cm of the isocenter was determined, with the spectrum is determined the total flow and the ambient dose equivalent. (Author)

  3. The effective delayed neutron fraction for bare-metal criticals

    International Nuclear Information System (INIS)

    Pearlstein, S.

    1999-01-01

    Given sufficient material, a large number of actinides could be used to form bare-metal criticals. The effective delayed neutron fraction for a bare critical comprised of a fissile material is comparable with the absolute delayed neutron fraction. The effective delayed neutron fraction for a bare critical composed of a fissionable material is reduced by factors of 2 to 10 when compared with the absolute delayed neutron fraction. When the effective delayed neutron fraction is small, the difference between delayed and prompt criticality is small, and extreme caution must be used in critical assemblies of these materials. This study uses an approximate but realistic model to survey the actinide region to compare effective delayed neutron fractions with absolute delayed neutron fractions

  4. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cartier, J. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France); Casoli, P. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, Valduc, F-21120 Is sur Tille (France); Chappert, F. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France)

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  5. Methods of neutron spectrum calculation from measured reaction rates in saips. Part 1. Review of mathematical methods

    International Nuclear Information System (INIS)

    Bondars, Kh.Ya.; Lapenas, A.A.

    1981-01-01

    We adapted or used on ES EhVM, operating under the control of OS ES, the currently most common algorithms for calculating neutron spectra from measured reaction rates. These programs, together with the neutron cross-section and spectrum libraries, are part of the computerized information system SAIPS. The present article descibes the basic mathematical concepts used in the algorithms of the SAIPS calculation programs

  6. {sup 252}Cf spontaneous prompt fission neutron spectrum measured at 0 degree and 180 degree relative to the fragment motion

    Energy Technology Data Exchange (ETDEWEB)

    Shanglian, Bao; Jinquan, Liu [Beijing Univ., BJ (China); Batenkov, O I; Blinov, M V; Smirnov, S N [V.G. Khlopin Radium Institute, ST. Petersburg (Russian Federation)

    1994-09-01

    The {sup 252}Cf spontaneous prompt fission neutron spectrum at 0 degree and 180 degree relative to the motion direction of corresponding fission fragments was measured. High angular resolution for fragment measurements and high energy resolution for neutron measurements were obtained using multi-parameter TOF spectrometer. The results showed that there is a symmetric distribution of `forward` and `backward` for low energy in C.M.S. neutrons, which was an evidence of nonequilibrium neutrons existed in fission process.

  7. Determination of the absolute K through O conversion coefficients of the 80-keV M4 transition in /sup 193/Ir/sup m/

    International Nuclear Information System (INIS)

    Lindner, M.; Gunnink, R.; Nagle, R.J.

    1987-01-01

    We produced carrier-free, nearly isotopically pure /sup 193/Ir/sup m/ from thermal-neutron irradiation of /sup 192/Os. From studies of the conversion-electron spectrum, the photon spectrum, and the absolute counting rates, we determined the absolute K, L1, L2, L3, M1, M3, M5, N, and O+P conversion coefficients for the first time. With possibly one exception, these values agree closely with theoretical calculations. The value for the energy of the unconverted gamma ray was found to be 80.22 +- 0.02 keV. The half-life for /sup 193/Ir/sup m/ determined by absolute electron counting in a proportional counter of 100% counting efficiency (4πβ) was found to be 10.53 +- 0.04 d. We have determined the L 1 subshell fluorescence yield, ω 1 , to be 0.120 +- 0.003 for iridium. Based upon our measurements of initial and final L subshell vacancies together with a best-fit literature value for ω 2 , we have found a plausible value for the Coster-Kronig coefficient f/sub 12/ to be 0.091 +- 0.011

  8. Neutron source multiplication method

    International Nuclear Information System (INIS)

    Clayton, E.D.

    1985-01-01

    Extensive use has been made of neutron source multiplication in thousands of measurements of critical masses and configurations and in subcritical neutron-multiplication measurements in situ that provide data for criticality prevention and control in nuclear materials operations. There is continuing interest in developing reliable methods for monitoring the reactivity, or k/sub eff/, of plant operations, but the required measurements are difficult to carry out and interpret on the far subcritical configurations usually encountered. The relationship between neutron multiplication and reactivity is briefly discussed and data presented to illustrate problems associated with the absolute measurement of neutron multiplication and reactivity in subcritical systems. A number of curves of inverse multiplication have been selected from a variety of experiments showing variations observed in multiplication during the course of critical and subcritical experiments where different methods of reactivity addition were used, with different neutron source detector position locations. Concern is raised regarding the meaning and interpretation of k/sub eff/ as might be measured in a far subcritical system because of the modal effects and spectrum differences that exist between the subcritical and critical systems. Because of this, the calculation of k/sub eff/ identical with unity for the critical assembly, although necessary, may not be sufficient to assure safety margins in calculations pertaining to far subcritical systems. Further study is needed on the interpretation and meaning of k/sub eff/ in the far subcritical system

  9. Absolute nuclear material assay using count distribution (LAMBDA) space

    Science.gov (United States)

    Prasad, Manoj K [Pleasanton, CA; Snyderman, Neal J [Berkeley, CA; Rowland, Mark S [Alamo, CA

    2012-06-05

    A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

  10. On the proton spectrum in free neutron β-decay

    International Nuclear Information System (INIS)

    Bunatyan, G.G.

    2000-01-01

    We consider the calculations which are appropriate to acquire with a high precision, of ∼ 1% or better, the general characteristics of weak interactions from the experiments on the free neutron β-decay; the principle emphasis is placed on the phenomena associated with the recoil of protons. The part played by electromagnetic interactions in β-decay is visualized, with special attention drawn to the influence of the γ-radiation on the momentum distribution of the particles in the final state. The effect of electromagnetic interactions on the proton recoil spectrum is studied, in the light of the experiments which are carried out and planned for now. The results of the calculations, which are to be confronted with the experimental data, are presented upright in terms of the effective Lagrangian underlying the inquiry. Owing to electromagnetic interactions, the corrections to the energy distribution of protons prove to amount to the value of a few per cent. Nowadays, this is substantial to obtain with a high accuracy the characteristics of weak interactions by processing the data of the experiments on the proton distribution in the free neutron β-decay

  11. Possible error-prone repair of neoplastic transformation induced by fission-spectrum neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Hill, C K; Han, A; Elkind, M M

    1948-01-01

    An examination was made of the effect of fission-spectrum neutrons from the JANUS reactor at Argonne National Laboratory, delivered either as acute or protracted irradiation, on the incidence of neoplastic transformation in the C3H 10T1/2 mouse embryo cell line. Acute exposures were delivered at 10-38 cGy min/sup -1/, protracted exposures at 0.086 or 0.43 cGy min/sup -1/. The total doses for both ranged from 2.4 to 350 cGy. In the low dose region (2.4-80 cGy), there was a large enhancement in transformation frequency when the neutrons were delivered at the low dose rates compared with the high dose rates, but the survival of the cells was not significantly different between the two exposures conditions. Analysis of the intial parts of the curves shows that the regression line for protracted doses is about 9 times steeper than that for single acute exposures. Finally, the possibility is discussed that an ''error-prone'' repair process may be causing the enhanced transformation frequency by protracted neutron exposures.

  12. Average cross section measurements in U-235 fission neutron spectrum for some threshold reactions

    International Nuclear Information System (INIS)

    Maidana, N.L.

    1993-01-01

    The average cross section in the 235 U fission spectrum has been measured by the activation technique, for the following thresholds reactions: 115 In(n,n') 115m In, 232 Th(n,f) P.F., 46 , 47 , 48 Ti(n,p) 46,47 , 48 Sc, 55 Mn(n,2 n) 54 Mn, 51 V(n,α) 48 Sc, 90 Zr(n,2 n) 89 Zr, 93 Nb(n,2 n) 92m Nb, 58 Ni(n,2 n) 57 Ni, 24 Mg(n,p) 24 Na, 56 Fe(n,p) 56 Mn, 59 Co(n,α) 56 Mn and 63 Cu(n,α) 60 Co. The activation foils were irradiated close (∼ 4 mm) to the core of the IEA-R1 research reactor in the IPEN-CNEN/SP. The reactor was operated at 2 MW yielding a fast neutron flux around 5 x 10 12 n.cm -2 . s -1 . The neutron flux density was monitored by activation reactions with well known averaged cross sections and with effective thresholds above 1 MeV. The foil activities were measured in a calibrated HPGe spectrometer. The neutron spectrum has been calculated using the SAIPS unfolding system applied to the activation data. A detailed error analysis was performed using the covariance matrix methodology. The results were compared with those from other authors. (author)

  13. Study of a neutronic potential of a modular fast spectrum ADS for radiotoxic waste transmutation

    International Nuclear Information System (INIS)

    Slessarev, I.; Arkhipov, V.

    1999-01-01

    transmutation are the principal criteria for ADS development in nuclear power (NP). It is known that fast spectrum ADS have promising parameters to be used for this goal. However, there is no international consensus yet and many innovative ADS concepts are contradictory. The principal goals have been to evaluate a maximum rate of waste transmutation in a fast spectrum ADS, to assess minimum TRU and MA Specific Fuel Inventory (SFI) and 'critical mass' corresponding to a given K eff , and to evaluate the sensitivity of fast spectrum ADS neutron balance to the external neutron spectrum (this is one of specific features of ADS). This parameter is correlated with the neutron source importance, which in turn affects the accelerator current demand and, hence, the ADS economics

  14. Development of a photonuclear activation file and measurement of delayed neutron spectra; Creation d'une bibliotheque d'activation photonucleaire et mesures de spectres d'emission de neutrons retardes

    Energy Technology Data Exchange (ETDEWEB)

    Giacri-Mauborgne, M.L

    2005-11-15

    This thesis work consists in two parts. The first part is the description of the creation of a photonuclear activation file which will be used to calculated photonuclear activation. To build this file we have used different data sources: evaluations but also calculations done using several cross sections codes (HMS-ALICE, GNASH, ABLA). This file contains photonuclear activation cross sections for more than 600 nuclides and fission fragments distributions for 30 actinides at tree different Bremsstrahlung energies and the delay neutron spectrum associated. These spectra are not in good agreement with experimental data. That is why we decided to launch measurement of delayed neutrons spectra from photofission. The second part of this thesis consists in demonstrating the possibility to do such measurements at the ELSA accelerator facility. To that purpose, we have developed the detection, the acquisition system and the analysis method of such spectra. These were tested for the measurement of the delayed neutron spectrum of uranium-238 after irradiation in a 2 MeV neutron flux. Finally, we have measured the delayed neutron spectrum of uranium-238 after irradiation in a 15 MeV Bremsstrahlung flux. We compare our results with experimental data. The experiment has allowed us to improve the value of {nu}{sub p}-bar with an absolute uncertainty below 7%, we propose {nu}{sub p}-bar = (3.03 {+-} 0.02) n/100 fissions, and to correct the Nikotin's parameters for the six group representation. Particularly, we have improved the data concerning the sixth group by taking into account results from different irradiation times.

  15. Neutron spectrum adjustment using reaction rate data acquired with a liquid dosimetry system

    International Nuclear Information System (INIS)

    Smith, D.L.; Ikeda, Y.; Uno, Y.; Maekawa, F.

    1997-01-01

    A dosimetry technique based on neutron activation of circulating water with dissolved salts is discussed. The neutron source was the FNS accelerator at JAERI, Tokai, Japan. Yttrium chloride hexahydrate (YCl 3· 6H 2 O) was the salt (264.9 grams dissolved in 16.094 liters of water). Gamma-ray yields were measured with an intrinsic Ge detector. The following reactions were examined: (1) 16 O(n,p) 16 N (E thresh = 10.245 MeV, t 1/2 = 7.13 sec, E γ = 6.129 MeV); (2) 37 Cl(n,p) 37 S (E thresh = 4.194 MeV, t 1/2 = 5.05 min, E γ = 3.104 MeV); (3) 89 Y(n,n') 89m Y (E thresh = 0.919 MeV, t 1/2 = 16.06 sec, E γ = 0.909 MeV). This paper describes use of the generalized least-squares (GLS) method to adjust the neutron spectrum

  16. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  17. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    International Nuclear Information System (INIS)

    Zamani, M.; Kasesaz, Y.; Khalafi, H.; Shayesteh, M.

    2015-01-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  18. Unfolding neutron spectrum with Markov Chain Monte Carlo at MIT research Reactor with He-3 Neutral Current Detectors

    Science.gov (United States)

    Leder, A.; Anderson, A. J.; Billard, J.; Figueroa-Feliciano, E.; Formaggio, J. A.; Hasselkus, C.; Newman, E.; Palladino, K.; Phuthi, M.; Winslow, L.; Zhang, L.

    2018-02-01

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CEνNS) using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2 × 1018 ν/second in its core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped Bonner cylinders around a 32He thermal neutron detector, whose data was then unfolded via a Markov Chain Monte Carlo (MCMC) to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.

  19. Absolute measurements of neutron cross sections. Progress report

    International Nuclear Information System (INIS)

    1984-11-01

    In the photoneutron laboratory, we have completed a major refurbishing of experimental facilities and begun work on measurements of the capture cross section in thorium and U-238. In the 14 MeV neutron experimental bay, work continues on the measurement of 14 MeV neutron induced reactions of interest as standards or because of their technological importance. First results have been obtained over the past year, and we are extending these measurements along the lines outlined in our proposal of a year ago

  20. Optimization of artificial neural networks for the reconstruction of the neutrons spectrum and their equivalent doses

    International Nuclear Information System (INIS)

    Reyes A, A.; Ortiz R, J. M.; Reyes H, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

    2014-08-01

    In this work was used the robust design methodology of artificial neural networks to determine a good topology of net able to solve with efficiency the problems of neutrons spectrometry and dosimetry. For the design of the topology of optimized net 36 different net architectures based on an orthogonal arrangement with a configuration L 9 (3 4 ), L 4 (3 2 ) were trained. For the training of the neural networks, was used a computer code developed in the ambient of Mat lab programming, which automates the process and analysis of the information, reducing the time used in this activity considerably for the investigator. For the training of the propagation nets forward was utilized a neutrons spectrum compendium published by the International Atomic Energy Agency, where of the total 80% was used for the training and 20% for the test, it trained with an inverse propagation algorithm being the entrance data the count rates corresponding to the 7 spheres of the spectrometric system of Bonner spheres, as exit data, the neural network obtains the neutrons spectrum expressed in 60 energy groups and are calculated of simultaneous way 15 dosimetric quantities. (Author)

  1. Experimental determination of spectral ratios and of neutrons energy spectrum in the fuel of the IPEN/MB-01 nuclear reactor

    International Nuclear Information System (INIS)

    Nunes, Beatriz Guimaraes

    2012-01-01

    This study aims to determine the spectral ratios and the neutron energy spectrum inside the fuel of IPEN/MB-01 Nuclear Reactor. These parameters are of great importance to accurately determine spectral physical parameters of nuclear reactors like reaction rates, fuel lifetime and also security parameters such as reactivity. For the experiment, activation detectors in the form of thin metal foils were introduced in a collapsible fuel rod. Then the rod was placed in the central position of the core which has a standard rectangular configuration of 26 x 28 fuel rods. There were used activation detectors from different elements such Au-197, U-238, Sc-45, Ni-58, Mg-24, Ti-47 and In-115 to cover a large range of the neutron energy spectrum. After the irradiation, the activation detectors were submitted to gamma spectrometry using a counting system with high purity Germanium, to obtain the reaction rates (saturation activity) per target nucleus. The spectral ratios were compared with calculated values obtained by the Monte Carlo method using the MCNP-4C code. The neutron energy spectrum was obtained inside the fuel rod using the SANDBP code with an input spectrum obtained by the MCNP-4C code, based on the saturation activity per target nucleus values of the activation detectors irradiated. (author)

  2. Procedure to measure the neutrons spectrum around a lineal accelerator for radiotherapy; Procedimiento para medir el espectro de los neutrones en torno a un acelerador lineal para radioterapia

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Benites R, J. L. [Servicios de Salud de Nayarit, Centro Estatal de Cancerologia, Calzada de la Cruz 116 Sur, 63000 Tepic, Nayarit (Mexico); Salas L, M. A., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Agronomia, Apdo. Postal 336, 98000 Zacatecas (Mexico)

    2013-10-15

    An experimental procedure was developed, by means of Bonner spheres, to measure the neutrons spectrum around Linacs of medical use that only requires of a single shot of the accelerator; to this procedure we denominate Planetary or Isocentric method. One of the problems associated to the neutrons spectrum measurement in a radiotherapy room with lineal accelerator is because inside the room a mixed, intense and pulsed radiation field takes place affecting the detection systems based on active detector; this situation is solved using a passive detector. In the case of the Bonner spheres spectrometer the active detector has been substituted by activation detectors, trace detectors or thermoluminescent dosimeters. This spectrometer uses several spheres that are situated one at a time in the measurement point, this way to have the complete measurements group the accelerator should be operated, under the same conditions, so many times like spheres have the spectrometer, this activity can consume a long time and in occasions due to the work load of Linac to complicate the measurement process too. The procedure developed in this work consisted on to situate all the spectrometer spheres at the same time and to make the reading by means of a single shot, to be able to apply this procedure, is necessary that before the measurements two characteristics are evaluated: the cross-talking of the spheres and the symmetry conditions of the neutron field. This method has been applied to determine the photo-neutrons spectrum produced by a lineal accelerator of medical use Varian ix of 15 MV to 100 cm of the isocenter located to 5 cm of depth of a solid water mannequin of 30 x 30 x 15 cm. The spectrum was used to determine the total flow and the environmental dose equivalent. (Author)

  3. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  4. Impact of statistical uncertainty of the neutron spectrum in the isotopic evolution of fuel

    International Nuclear Information System (INIS)

    Ortega, P.

    2012-01-01

    The results obtained and presented in this study for different calculation conditions (number of stories, number of steps burning, etc.) and their simultaneous impact on neutron spectrum and isotopic composition and a methodology is proposed to determine the minimum parameters for calculation given uncertainty in the results of isotopic composition with high burnup, both UO 2 and MOX fuel.

  5. Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

    International Nuclear Information System (INIS)

    De Leeuw-Gierts, G.; De Leeuw, S.; Hansen, G.E.; Helmick, H.H.

    1979-01-01

    The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de L'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

  6. Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

    International Nuclear Information System (INIS)

    Leeuw-Gierts, G. de; Leeuw, S. de

    1980-01-01

    The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de l'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

  7. High-resolution measurements of the DT neutron spectrum using new CD foils in the Magnetic Recoil neutron Spectrometer (MRS) on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Gatu Johnson, M., E-mail: gatu@psfc.mit.edu; Frenje, J. A.; Li, C. K.; Petrasso, R. D.; Séguin, F. H. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Bionta, R. M.; Casey, D. T.; Eckart, M. J.; Grim, G. P.; Hartouni, E. P.; Hatarik, R.; Sayre, D. B.; Skulina, K.; Yeamans, C. B. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Farrell, M. P.; Hoppe, M.; Kilkenny, J. D.; Reynolds, H. G.; Schoff, M. E. [General Atomics, San Diego, California 92186 (United States)

    2016-11-15

    The Magnetic Recoil neutron Spectrometer (MRS) on the National Ignition Facility measures the DT neutron spectrum from cryogenically layered inertial confinement fusion implosions. Yield, areal density, apparent ion temperature, and directional fluid flow are inferred from the MRS data. This paper describes recent advances in MRS measurements of the primary peak using new, thinner, reduced-area deuterated plastic (CD) conversion foils. The new foils allow operation of MRS at yields 2 orders of magnitude higher than previously possible, at a resolution down to ∼200 keV FWHM.

  8. Neutron detection gamma ray sensitivity criteria

    International Nuclear Information System (INIS)

    Kouzes, Richard T.; Ely, James H.; Lintereur, Azaree T.; Mace, Emily K.; Stephens, Daniel L.; Woodring, Mitchell L.

    2011-01-01

    The shortage of 3 He has triggered the search for effective alternative neutron detection technologies for national security and safeguards applications. Any new detection technology must satisfy two basic criteria: (1) it must meet a neutron detection efficiency requirement, and (2) it must be insensitive to gamma-ray interference at a prescribed level, while still meeting the neutron detection requirement. It is the purpose of this paper to define measureable gamma ray sensitivity criteria for neutron detectors. Quantitative requirements are specified for: intrinsic gamma ray detection efficiency and gamma ray absolute rejection. The gamma absolute rejection ratio for neutrons (GARRn) is defined, and it is proposed that the requirement for neutron detection be 0.9 3 He based neutron detector is provided showing that this technology can meet the stated requirements. Results from tests of some alternative technologies are also reported.

  9. Differential neutron production cross sections and neutron yields from stopping-length targets for 113-MeV protons

    International Nuclear Information System (INIS)

    Meier, M.M.; Amian, W.B.; Clark, D.A.; Goulding, C.A.; McClelland, J.B.; Morgan, G.L.; Moss, C.E.

    1989-03-01

    We have measured differential (P,ξn) cross sections, d 2 σ/dΩdE/sub n/, from thin targets and absolute neutron yields from stopping-length targets at angles of 7.5/degree/, 30/degree/, 60/degree/, and 150/degree/, for the 113--MeV proton bombardment of elemental beryllium, carbon, aluminum, iron, and depleted uranium. Additional cross-section measurements are reported for oxygen, tungsten, and lead. We used time-of-flight techniques to identify and discriminate against backgrounds and to determine the neutron energy spectrum. Comparison of the experimental data with intranuclear-cascade evaporation-model calculations with the code HETC showed discrepancies as high as a factor of 7 in the differential cross sections. These discrepancies in the differential cross sections make it possible to identify some of the good agreement seen in the stopping-length yield comparisons as fortuitous cancellation of incorrect production estimates in different energy regimes. 13 refs., 20 figs., 4 tabs

  10. Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

    Science.gov (United States)

    Cook, J. C.; Barker, J. G.; Rowe, J. M.; Williams, R. E.; Gagnon, C.; Lindstrom, R. M.; Ibberson, R. M.; Neumann, D. A.

    2015-08-01

    The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

  11. Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

    Energy Technology Data Exchange (ETDEWEB)

    Cook, J.C.; Barker, J.G.; Rowe, J.M.; Williams, R.E. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States); Gagnon, C. [Department of Materials Science and Engineering, University of Maryland, College Park, MD 20742 (United States); Lindstrom, R.M. [Scientist Emeritus, Chemical Sciences Division, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 8395, Gaithersburg, MD 20899-8395 (United States); Ibberson, R.M.; Neumann, D.A. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States)

    2015-08-21

    The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

  12. Characterizing ICF Neutron Diagnostics on the nTOF line at SUNY Geneseo

    Science.gov (United States)

    Simone, Angela; Padalino, Stephen; Turner, Ethan; Ginnane, Mary Kate; Dubois, Natalie; Fletcher, Kurtis; Giordano, Michael; Lawson-Keister, Patrick; Harrison, Hannah; Visca, Hannah; Sangster, Craig; Regan, Sean

    2014-10-01

    Charged particle beams from the Geneseo 1.7 MV tandem Pelletron accelerator produce nuclear reactions that emit neutrons in the range of 0.5 to 17.9 MeV via the d(d,n)3He and 11B(d,n)12C reactions. The neutron energy and flux can be adjusted by controlling the accelerator beam current and potential. This adjustable neutron source makes it possible to calibrate ICF and HEDP neutron scintillator diagnostics. However, gamma rays which are often present during an accelerator-based calibration are difficult to differentiate from neutron signals in scintillators. To identify neutrons from gamma rays and to determine their energy, a permanent neutron time-of-flight (nTOF) line is being constructed. By detecting the scintillator signal in coincidence with an associated charged particle (ACP) produced in the reaction, the identity of the neutron can be known and its energy determined by time of flight. Using a 100% efficient surface barrier detector to count the ACPs, the absolute efficiency of the scintillator as a function of neutron energy can be determined. This is done by determining the ratio of the ACP counts in the singles spectrum to coincidence counts for matched solid angles of the SBD and scintillator. Funded in part by a LLE contract through the DOE.

  13. Neutron-deuteron analyzing power data at 19.0 MeV

    International Nuclear Information System (INIS)

    Weisel, G. J.; Tornow, W.; Crowe, B. J. III; Crowell, A. S.; Esterline, J. H.; Howell, C. R.; Kelley, J. H.; Macri, R. A.; Pedroni, R. S.; Walter, R. L.; Witala, H.

    2010-01-01

    Measurements of neutron-deuteron (n-d) analyzing power A y (θ) at E n =19.0 MeV are reported at 16 angles from θ c.m. =46.7 to 152.0 deg. The objective of the experiment is to better characterize the discrepancies between n-d data and the predictions of three-nucleon calculations for neutron energies above 16.0 MeV. The experiment used a shielded neutron source, which produced polarized neutrons via the 2 H(d-vector,n-vector) 3 He reaction, a deuterated liquid scintillator center detector (CD) and liquid-scintillator neutron side detectors. A coincidence between the CD and the side detectors isolated the elastic-scattering events. The CD pulse height spectrum associated with each side detector was sorted by using pulse-shape discrimination, time-of-flight techniques, and by removing accidental coincidences. A Monte Carlo computer simulation of the experiment accounted for effects due to finite geometry, multiple scattering, and CD edge effects. The resulting high-precision data (with absolute uncertainties ranging from 0.0022 to 0.0132) have a somewhat lower discrepancy with the predictions of three-body calculations, as compared to those found at lower energies.

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  15. Secondary standard neutron detector for measuring total reaction cross sections

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  16. Program HEFEST for calculation of neutron spectrum on the basis of the activity of threshold detectors; Progam HEFEST za obradu neutronskog spektra na osnovu aktivnosti prag detektora

    Energy Technology Data Exchange (ETDEWEB)

    Cupac, S; Sokcic-Kostic, M; Pesic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1985-07-01

    Program HEFEST for calculation of neutron spectrum on the basis of the activity of threshold detectors is described in this paper. After testing, program is used for the elaboration of the experimental results in determining the fast neutron spectrum on the coupled fast-thermal system on the reactor RB in IBK. (author)

  17. NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code

    International Nuclear Information System (INIS)

    Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.

    1977-02-01

    The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes

  18. Measurement of the absolute values of cross-sections in neutron photoproduction (1962); Mesure de sections efficaces de photoproduction de neutrons en valeur absolue (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Schuhl, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    The absolute values of photoneutrons production cross-sections for the case of intermediate and heavy nuclei (lanthanium, cerium, tantalum, gold, lead and bismuth) are determined with an error of 15 per cent. The results obtained agree with theories in which the giant resonance is explained by the collective motion of the protons against the neutrons. The effect of the nuclear deformation on the shape of the giant resonance is seen in the case of Ta{sup 181}, it will be possible to determine the quadrupole momenta of deformed nuclei with a good accuracy when we shall increase the statistics of measurements. (author) [French] Les sections efficaces de production de photoneutrons par divers noyaux moyens et lourds (lanthane, cerium, tantale, or, plomb et bismuth) sont determinees en valeur absolue avec une erreur relative de 15 pour cent. Les resultats obtenus s'accordent avec les theories qui interpretent la resonance geante par un mouvement collectif des protons par rapport aux neutrons. L'influence de la deformation du noyau sur la forme de la resonance geante est soulignee dans le cas de {sup 181}Ta pour lequel elle se decompose en deux pics. Une amelioration de la statistique des mesures permettra de determiner les moments quadrupolaires des noyaux deformes avec une meilleure precision. (auteur)

  19. A search for fine structure of the time-of-flight spectrum of the fission neutrons of 252Cf

    International Nuclear Information System (INIS)

    Scobie, J.; Scott, R.D.; Feather, N.; Vass, D.G.

    1977-01-01

    A standard time-of-flight arrangement, in which start pulses were supplied by fission fragments and stop pulses by neutrons, has been employed in an attempt to check recent claims of the existence of fine structures in the time-of-flight spectrum of the fission neutrons of 252 Cf. This structure, in the form of spikes with tails towards longer times, has been attributed to the emission of neutrons of short delay (with half-lives of a few to a hundred or so nanoseconds) in the fission process. It has not been possible to find any convincing evidence for the existence of such structure. (author)

  20. Using an interference spectrum as a short-range absolute rangefinder with fiber and wideband source

    Science.gov (United States)

    Hsieh, Tsung-Han; Han, Pin

    2018-06-01

    Recently, a new type of displacement instrument using spectral-interference has been found, which utilizes fiber and a wideband light source to produce an interference spectrum. In this work, we develop a method that measures the absolute air-gap distance by taking wavelengths at two interference spectra minima. The experimental results agree with the theoretical calculations. It is also utilized to produce and control the spectral switch, which is much easier than other previous methods using other control mechanisms. A scanning mode of this scheme for stepped surface measurement is suggested, which is verified by a standard thickness gauge test. Our scheme is different to one available on the market that may use a curve-fitting method, and some comparisons are made between our scheme and that one.

  1. Measurement of the time dependent neutron energy spectrum in the 'DENA' plasma focus device

    Energy Technology Data Exchange (ETDEWEB)

    Abdollahzadeh, M [Department of Physics, Imam Husein University, PO Box 16575-347, Tehran (Iran, Islamic Republic of); Sadat kiai, S M [Nuclear Science and Technology Research Institute (NSTRI), Nuclear Science Research School, A.E.O.I., PO Box 14155-1339, Tehran (Iran, Islamic Republic of); Babazadeh, A R [Physics Department, Qom University, PO Box 37165, Qom (Iran, Islamic Republic of)

    2008-10-15

    An extended time of flight method is used to determine the time dependent neutron energy spectrum in the Filippove type 'Dena' plasma focus (90 kJ, 25 kV, 288 {mu}F), filled with deuterium gas. An array of 5 detectors containing NE-102 plastic scintillators+photomultipliers is used. The number and position of the detectors are determined by a Monte Carlo program and the MCNP code. This paper briefly describes the simulation method and presents the experimental measurements and their results. The mechanisms of neutron production (thermonuclear and non-thermonuclear) and their time variations are discussed.

  2. Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Suman, Vitisha [Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sarkar, P.K., E-mail: pksarkar02@gmail.com [Manipal Centre for Natural Sciences, Manipal University, Manipal 576104 (India)

    2014-02-11

    A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra.

  3. Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation

    International Nuclear Information System (INIS)

    Suman, Vitisha; Sarkar, P.K.

    2014-01-01

    A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra

  4. Neutrons from Antiproton Irradiation

    DEFF Research Database (Denmark)

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

    the neutron spectrum. Additionally, we used a cylindrical polystyrene loaded with several pairs of thermoluminescent detectors containing Lithium-6 and Lithium-7, which effectively detects thermalized neutrons. The obtained results are compared with FLUKA imulations. Results: The results obtained...... spectrum is very low, and does not pose a problem for radiation therapy. However, the contribution from fast neutrons is much more significant. The dose equivalent contribution from neutrons originate from the patient alone and reaches levels which are found in passive moderated proton therapy. The exact...

  5. Neutron Beam Filters

    International Nuclear Information System (INIS)

    Adib, M.

    2011-01-01

    The purpose of filters is to transmit neutrons with selected energy, while remove unwanted ones from the incident neutron beam. This reduces the background, and the number of spurious. The types of commonly used now-a-day neutron filters and their properties are discussed in the present work. There are three major types of neutron filters. The first type is filter of selective thermal neutron. It transmits the main reflected neutrons from a crystal monochromate, while reject the higher order contaminations accompanying the main one. Beams coming from the moderator always contain unwanted radiation like fast neutrons and gamma-rays which contribute to experimental background and to the biological hazard potential. Such filter type is called filter of whole thermal neutron spectrum. The third filter type is it transmits neutrons with energies in the resonance energy range (En . 1 KeV). The main idea of such neutron filter technique is the use of large quantities of a certain material which have the deep interference minima in its total neutron cross-section. By transmitting reactor neutrons through bulk layer of such material, one can obtain the quasimonochromatic neutron lines instead of white reactor spectrum.

  6. Facility for fast neutron irradiation tests of electronics at the ISIS spallation neutron source

    International Nuclear Information System (INIS)

    Andreani, C.; Pietropaolo, A.; Salsano, A.; Gorini, G.; Tardocchi, M.; Paccagnella, A.; Gerardin, S.; Frost, C. D.; Ansell, S.; Platt, S. P.

    2008-01-01

    The VESUVIO beam line at the ISIS spallation neutron source was set up for neutron irradiation tests in the neutron energy range above 10 MeV. The neutron flux and energy spectrum were shown, in benchmark activation measurements, to provide a neutron spectrum similar to the ambient one at sea level, but with an enhancement in intensity of a factor of 10 7 . Such conditions are suitable for accelerated testing of electronic components, as was demonstrated here by measurements of soft error rates in recent technology field programable gate arrays

  7. 14 MeV calibration of JET neutron detectors—phase 1: calibration and characterization of the neutron source

    Science.gov (United States)

    Batistoni, P.; Popovichev, S.; Cufar, A.; Ghani, Z.; Giacomelli, L.; Jednorog, S.; Klix, A.; Lilley, S.; Laszynska, E.; Loreti, S.; Packer, L.; Peacock, A.; Pillon, M.; Price, R.; Rebai, M.; Rigamonti, D.; Roberts, N.; Tardocchi, M.; Thomas, D.; Contributors, JET

    2018-02-01

    In view of the planned DT operations at JET, a calibration of the JET neutron monitors at 14 MeV neutron energy is needed using a 14 MeV neutron generator deployed inside the vacuum vessel by the JET remote handling system. The target accuracy of this calibration is  ±10% as also required by ITER, where a precise neutron yield measurement is important, e.g. for tritium accountancy. To achieve this accuracy, the 14 MeV neutron generator selected as the calibration source has been fully characterised and calibrated prior to the in-vessel calibration of the JET monitors. This paper describes the measurements performed using different types of neutron detectors, spectrometers, calibrated long counters and activation foils which allowed us to obtain the neutron emission rate and the anisotropy of the neutron generator, i.e. the neutron flux and energy spectrum dependence on emission angle, and to derive the absolute emission rate in 4π sr. The use of high resolution diamond spectrometers made it possible to resolve the complex features of the neutron energy spectra resulting from the mixed D/T beam ions reacting with the D/T nuclei present in the neutron generator target. As the neutron generator is not a stable neutron source, several monitoring detectors were attached to it by means of an ad hoc mechanical structure to continuously monitor the neutron emission rate during the in-vessel calibration. These monitoring detectors, two diamond diodes and activation foils, have been calibrated in terms of neutrons/counts within  ±5% total uncertainty. A neutron source routine has been developed, able to produce the neutron spectra resulting from all possible reactions occurring with the D/T ions in the beam impinging on the Ti D/T target. The neutron energy spectra calculated by combining the source routine with a MCNP model of the neutron generator have been validated by the measurements. These numerical tools will be key in analysing the results from the in

  8. The perturbation of backscattered fast neutrons spectrum caused by the resonances of C, N and O for possible use in pyromaterial detection

    Energy Technology Data Exchange (ETDEWEB)

    Abedin, Ahmad Firdaus Zainal, E-mail: firdaus087@gmail.com; Ibrahim, Noorddin; Zabidi, Noriza Ahmad; Abdullah, Abqari Luthfi Albert [Department of Defence Science, Universiti Pertahanan Nasional Malaysia, Kem Sungai Besi, Kuala Lumpur 57000 (Malaysia)

    2015-04-29

    Neutron radiation is able to determine the signature of land mine detection based on backscattering energy spectrum of landmine. In this study, the Monte Carlo simulation of backscattered fast neutrons was performed on four basic elements of land mine; hydrogen, nitrogen, oxygen and carbon. The moderation of fast neutrons to thermal neutrons and their resonances cross-section between 0.01 eV until 14 MeV were analysed. The neutrons energies were divided into 29 groups and ten million neutrons particles histories were used. The geometries consist of four main components: neutrons source, detectors, landmine and soil. The neutrons source was placed at the origin coordinate and shielded with carbon and polyethylene. Americium/Beryllium neutron source was placed inside lead casing of 1 cm thick and 2.5 cm height. Polyethylene was used to absorb and disperse radiation and was placed outside the lead shield of width 10 cm and height 7 cm. Two detectors were placed between source with distance of 8 cm and radius of 1.9 cm. Detectors of Helium-3 was used for neutron detection as it has high absorption cross section for thermal neutrons. For the anomaly, the physical is in cylinder form with radius of 10 cm and 8.9 cm height. The anomaly is buried 5 cm deep in the bed soil measured 80 cm radius and 53.5 cm height. The results show that the energy spectrum for the four basic elements of landmine with specific pattern which can be used as indication for the presence of landmines.

  9. The perturbation of backscattered fast neutrons spectrum caused by the resonances of C, N and O for possible use in pyromaterial detection

    International Nuclear Information System (INIS)

    Abedin, Ahmad Firdaus Zainal; Ibrahim, Noorddin; Zabidi, Noriza Ahmad; Abdullah, Abqari Luthfi Albert

    2015-01-01

    Neutron radiation is able to determine the signature of land mine detection based on backscattering energy spectrum of landmine. In this study, the Monte Carlo simulation of backscattered fast neutrons was performed on four basic elements of land mine; hydrogen, nitrogen, oxygen and carbon. The moderation of fast neutrons to thermal neutrons and their resonances cross-section between 0.01 eV until 14 MeV were analysed. The neutrons energies were divided into 29 groups and ten million neutrons particles histories were used. The geometries consist of four main components: neutrons source, detectors, landmine and soil. The neutrons source was placed at the origin coordinate and shielded with carbon and polyethylene. Americium/Beryllium neutron source was placed inside lead casing of 1 cm thick and 2.5 cm height. Polyethylene was used to absorb and disperse radiation and was placed outside the lead shield of width 10 cm and height 7 cm. Two detectors were placed between source with distance of 8 cm and radius of 1.9 cm. Detectors of Helium-3 was used for neutron detection as it has high absorption cross section for thermal neutrons. For the anomaly, the physical is in cylinder form with radius of 10 cm and 8.9 cm height. The anomaly is buried 5 cm deep in the bed soil measured 80 cm radius and 53.5 cm height. The results show that the energy spectrum for the four basic elements of landmine with specific pattern which can be used as indication for the presence of landmines

  10. Review of Non-Neutron and Neutron Nuclear Data, 2004

    International Nuclear Information System (INIS)

    Holden, Norman E.

    2005-01-01

    Review articles are in preparation for the 2004 edition of the CRC Handbook of Chemistry and Physics dealing with the evaluation of both non-neutron and neutron nuclear data. Data on the discovery of element 110, Darmstadtium, and element 111 have been officially accepted, while data on element 118 have been withdrawn. Data to be presented include revised values for very short-lived nuclides, long-lived nuclides, and beta-beta decay measurements. There has been a reassessment of the spontaneous fission (sf) half-lives, which distinguishes between sf decay half-lives and cluster decay half-lives, and with cluster-fission decay. New measurements of isotopic abundance values for many elements will be discussed with an emphasis on the minor isotopes of interest for use in neutron activation analysis measurements. Neutron resonance integrals will be discussed emphasizing the differences between the calculated values obtained from the analytical integration over neutron resonances and the measured values in a neutron reactor-spectrum, which does not quite conform to the assumed 1/E neutron energy spectrum. The method used to determine the neutron resonance integral from measurement, using neutron activation analysis, will be discussed

  11. Methods of neutron spectrum calculation from measured reaction rates in SAIPS. Part 2: Software and data input

    International Nuclear Information System (INIS)

    Berzonis, M.A.; Bondars, H.Ya.

    1981-08-01

    A brief description of the SAIPS software and the basic principles of its application is given. SAIPS contains programs needed to unfold spectra, libraries of neutron cross sections and reference spectra, and software for automatic calculation and for system maintenance. SAIPS offers the possibility of determining the reliability of an unfolded neutron spectrum and of planning measurements and calculations by varying different factors: the errors in the reaction rates, the errors in the cross sections used, the detector assembly, the unfolding programs, etc. SAIPS runs on the ES 1022 computer

  12. Absolute spectrophotometry of the β Lyr

    International Nuclear Information System (INIS)

    Burnashev, V.I.; Skul'skij, M.Yu.

    1978-01-01

    In 1974 an absolute spectrophotometry of β Lyr was performed with the scanning spectrophotometer in the 3300-7400 A range. The energy distribution in the β Lyr spectrum is obtained. The β Lyr model is proposed. It is shown, that the continuous spectrum of the β Lyr radiation can be presented by the total radiation of the B8 3 and A5 3 two stars and of the gaseous envelope with Te =20000 K

  13. Influence of coated particle structure in thermal neutron spectrum energy range

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U; Teuchert, E

    1971-02-15

    The heterogenity due to lumping the fuel in coated particles affects the thermal neutron spectrum. A calculation model is discussed which, apart from some simplifying assumptions about the statistical distribution, allows a rigorous computation of effective cross sections for all nuclides of the heterogeneous medium. It is based on an exact computation of the neutron penetration probability through coating and kernel. The model is incorporated in a THERMOS-code providing a double heterogeneous cell calculation, which can be repeated automatically at different time steps in the depletion code system MAFIA-V.S.O.P.. A discussion of the effects of the coated particle structure is given by a comparison of calculations for heterogeneous and homogeneous fuel zones in pebble bed reactor elements. This is performed for enriched UO{sub 2} fuel and for a ThO{sub 2}-PuO{sub 2} mixture in the grains. Depending on the energy dependent total sigmas in the kernels the changes of the cross sections are ranging from 0.1% up to 45%. The influence on the spectrum averaged sigmas of the nuclides in the fresh UO{sub 2} fuel is lower than 1%. For the emerging {sup 240}Pu it increases up to 3.3% during irradiation. For the ThO{sub 2}-PuO{sub 2} fuel the averaged sigmas of the isotopes vary from 0.5% to 5.7% depending on the state of irradiation. Correspondingly there is an influence on the plutonium isotopic composition, on breeding ratios, and on the tilt of k{sub eff} during burnup which will be discussed in detail.

  14. Neutron-deuteron breakup experiment at En=13 MeV: Determination of the 1S0 neutron-neutron scattering length ann

    International Nuclear Information System (INIS)

    Gonzalez Trotter, D.E.; Meneses, F. Salinas; Tornow, W.; Howell, C.R.; Chen, Q.; Crowell, A.S.; Roper, C.D.; Walter, R.L.; Schmidt, D.; Witala, H.; Gloeckle, W.; Tang, H.; Zhou, Z.; Slaus, I.

    2006-01-01

    We report on results of a kinematically complete neutron-deuteron breakup experiment performed at Triangle Universities Nuclear Laboratory using an E n =13 MeV incident neutron beam. The 1 S 0 neutron-neutron scattering length a nn has been determined for four production angles of the neutron-neutron final-state interaction configuration. The absolute cross-section data were analyzed with rigorous three-nucleon calculations. Our average value of a nn =-18.7±0.7 fm is in excellent agreement with a nn =-18.6±0.4 fm obtained from capture experiments of negative pions on deuterons. We also performed a shape analysis of the final-state interaction cross-section enhancements by allowing the normalization of the data to float. From these relative data, we obtained an average value of a nn =-18.8±0.5 fm, in agreement with the result obtained from the absolute cross-section measurements. Our result deviates from the world average of a nn =-16.7±0.5 fm determined from previous kinematically complete neutron-deuteron breakup experiments, including the most recent one carried out at Bonn. However, this low value for a nn is at variance with theoretical expectation and other experimental information about the sign of charge-symmetry breaking of the nucleon-nucleon interaction. In agreement with theoretical predictions, no evidence was found of significant three-nucleon force effects on the neutron-neutron final-state interaction cross sections

  15. Development of Coincidence Method for Determination Thermal Neutron Flux on RSG-GAS

    International Nuclear Information System (INIS)

    Bakhri, Syaiful; Hamzah, Amir

    2004-01-01

    The research to develop detection radiation system using coincidence method has been done to determine thermal neutron flux in RS1 and RS2 irradiation facilities RSG-GAS. At this research has arranged beta-gamma coincidence equipment system and parameter of measurement according to Au-198 beta-gamma spectrum. Gold foils that have irradiated for period of time, counted, and the activities of radiation is analyzed to get neutron flux. Result of research indicate that systems measurement of absolute activity with gamma beta coincidence method functioning well and can be applied at activity measurement of gold foil for irradiation facility characterization. The results show that thermal neutron flux in RS1 and RS2, respectively is 2.007E+12 n/cm 2 s and 2.147E+12 n/cm 2 s. To examine the system performance, the result was compared to measure activity using high resolution of Hp Ge detector and achieved discrepancy is about 1.26% and 6.70%. (author)

  16. Measurement of neutron captured cross-sections in 1-2 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gi Dong; Kim, Young Sek; Kim, Jun Kon; Yang, Tae Keun [Korea Institutes of Geoscience and Mineral Resources, Taejeon (Korea)

    2001-04-01

    The measurement of neutron captured reaction cross sections was performed to build the infra system for the production of nuclear data. MeV neutrons were produced with TiT target and {sup 3}T(p,n){sup 3}He reaction. The characteristics of TiT thin film was analyzed with ERD-TOF and RBS. The results was published at Journal of the Korea Physical Society (SCI registration). The energy, the energy spread and the flux of the produced neutron were measured. The neutron excitation functions of {sup 12}C and {sup 16}O were obtained to confirm the neutron energy and neutron energy spread. The neutron energy spread found to be 1.3 % at the neutron energy of 2.077 MeV. The {sup 197}Au(n,{gamma}) reaction was performed to obtain the nerutron flux. The maximum neutron flux found to be 1 x 10{sup 8} neutrons/sec at the neutron energy of 2 MeV. The absolute efficiency of liquid scintillation detector was obtained in the neutron energy of 1 - 2 MeV. The fast neutron total reaction cross sections of Cu, Fe, and Au were measured with sample in-out method. Also the neutron captured reaction cross sections of {sup 63}Cu were measured with fast neutron activation method. The measurement of neutron total reaction cross sections and the neutron captured reaction cross sections with fast neutrons were first tried in Korea. The beam pulsing system was investigated and the code of calculating the deposition spectrums for primary gamma rays was made to have little errors at nuclear data. 25 refs., 28 figs., 14 tabs. (Author)

  17. Some neutronics of innovative subcritical assembly with fast neutron spectrum

    International Nuclear Information System (INIS)

    Kiyavitskaya, H.; Fokov, Yu.; Rutkovskaya, Ch.; Sadovich, S.; Kasuk, D.; Gohar, Y.; Bolshinsky, I.

    2013-01-01

    Conclusion: • New assembly can be used to: • develop the experimental techniques and adapt the existing ones for monitoring the sub-criticality level, neutron spectra measurements, etc; • study the spatial kinetics of sub-critical and critical systems with fast neutron spectra; • measure the transmutation reaction rates of minor-actinides etc

  18. Calculated neutron spectrum from 800-MeV protons incident on a copper beam stop

    International Nuclear Information System (INIS)

    Perry, D.G.

    1975-10-01

    A Monte Carlo calculation was performed to obtain the neutron spectrum generated by 800-MeV protons incident on the LAMPF main copper beam stop. The total flux is calculated to be of the order of 10 13 n/cm 2 -sec-mA at full-beam intensity of 1 mA, with flux spectra calculated for angles of 20 0 , 30 0 , 60 0 , 90 0 , 120 0 , and 150 0 . (auth)

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

    International Nuclear Information System (INIS)

    Amir Hamzah

    2004-01-01

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

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

    CERN Document Server

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

    2002-01-01

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

  1. Neutron radiography with sub-15 {mu}m resolution through event centroiding

    Energy Technology Data Exchange (ETDEWEB)

    Tremsin, Anton S., E-mail: ast@ssl.berkeley.edu [Space Sciences Laboratory, University of California at Berkeley, Berkeley, CA 94720 (United States); McPhate, Jason B.; Vallerga, John V.; Siegmund, Oswald H.W. [Space Sciences Laboratory, University of California at Berkeley, Berkeley, CA 94720 (United States); Bruce Feller, W. [NOVA Scientific, Inc. 10 Picker Road, Sturbridge, MA 01566 (United States); Lehmann, Eberhard; Kaestner, Anders; Boillat, Pierre; Panzner, Tobias; Filges, Uwe [Spallation Neutron Source Division, Paul Scherrer Institute, CH-5232 Villigen (Switzerland)

    2012-10-01

    Conversion of thermal and cold neutrons into a strong {approx}1 ns electron pulse with an absolute neutron detection efficiency as high as 50-70% makes detectors with {sup 10}B-doped Microchannel Plates (MCPs) very attractive for neutron radiography and microtomography applications. The subsequent signal amplification preserves the location of the event within the MCP pore (typically 6-10 {mu}m in diameter), providing the possibility to perform neutron counting with high spatial resolution. Different event centroiding techniques of the charge landing on a patterned anode enable accurate reconstruction of the neutron position, provided the charge footprints do not overlap within the time required for event processing. The new fast 2 Multiplication-Sign 2 Timepix readout with >1.2 kHz frame rates provides the unique possibility to detect neutrons with sub-15 {mu}m resolution at several MHz/cm{sup 2} counting rates. The results of high resolution neutron radiography experiments presented in this paper, demonstrate the sub-15 {mu}m resolution capability of our detection system. The high degree of collimation and cold spectrum of ICON and BOA beamlines combined with the high spatial resolution and detection efficiency of MCP-Timepix detectors are crucial for high contrast neutron radiography and microtomography with high spatial resolution. The next generation of Timepix electronics with sparsified readout should enable counting rates in excess of 10{sup 7} n/cm{sup 2}/s taking full advantage of high beam intensity of present brightest neutron imaging facilities.

  2. Study on the dose distribution of the mixed field with thermal and epi-thermal neutrons for neutron capture therapy

    International Nuclear Information System (INIS)

    Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

    1994-01-01

    Simulation calculations using DOT 3.5 were carried out in order to confirm the characteristics of depth-dependent dose distribution in water phantom dependent on incident neutron energy. The epithermal neutrons mixed to thermal neutron field is effective improving the thermal neutron depth-dose distribution for neutron capture therapy. A feasibility study on the neutron energy spectrum shifter was performed using ANISN-JR for the KUR Heavy Water Facility. The design of the neutron spectrum shifter is feasible, without reducing the performance as a thermal neutron irradiation field. (author)

  3. 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)

  4. The measurement of tripartition alpha particle low energy spectrum in 235U fission induced by thermal neutrons

    International Nuclear Information System (INIS)

    El Hage Sleiman, F.

    1980-01-01

    The energy spectrum of the α particles emitted in the thermal neutron induced fission of 235 U was measured from 11.5 MeV down to 2 MeV using the parabola mass spectrometer Lohengrin at the ILL high flux reactor. A Monte Carlo program, that simulates the α particle motion to the spectrometer, has been developed. Numerical results of Monte Carlo calculations for differents values of parameter are reported. The overall energy spectrum is slightly asymmetric at low energy. The possible reasons for the existence of this asymmetry are discussed [fr

  5. Sensitivity studies of the neutron multiplicity spectrum in the spallation of Pb targets

    International Nuclear Information System (INIS)

    Sinha, A.; Garg, S.B.; Srinivasan, M.

    1986-01-01

    The number of neutrons produced per incident proton in the spallation of Pb targets is of direct relevance to the design of accelerator breeders. The nuclear cascade initiated by high-energy protons in spallation targets is usually described by an intranuclear cascade evaporation (INCE) model. Even though this model describes various average nuclear properties of spallation targets fairly well, differential quantities such as energy spectra, angular spectra etc., are not reproduced within the limits of experimental uncertainty. One of the reasons for this is the uncertainty in the magnitude of the parameters involved in the model, notably the level density parameter Bsub(O) whose magnitude is quoted by different workers to be in the range of 8-20 MeV. The accuracy of Bsub(O) could be improved if we could experimentally determine a quantity which is much more sensitive to Bsub(O) than the average neutron yield. In this paper we discuss one such quantity, namely the neutron multiplicity spectrum (MS). We compute the MS due to the spallation of Pb targets of different sizes at proton energies of 1.5, 1.0 and 0.59 GeV using the Monte Carlo code HETC. It is noticed that for the 1.5 GeV proton case the probability P(ν) for leakage of ν neutrons for ν in the range of 60-65, changes by about 70% when Bsub(O) is varied from 8 to 20 MeV. The corresponding change in the average neutron yield is <20%. It is therefore suggested that an accurate measurement of the MS can serve as a useful tool to narrow down the range of uncertainty in the Bsub(O) parameter. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  7. A comparison of the predicted and observed reaction rates of various neutron detectors in a thermal reactor spectrum

    International Nuclear Information System (INIS)

    Hardiman, J.P.; Maunders, E.J.

    1963-08-01

    A number of the detectors commonly used in integral neutron spectrum measurements have been irradiated in the pitch moderator position of a Calder Hall lattice where the detailed energy spectrum is known from time of flight measurements. Predicted and observed reaction rates are generally in good agreement although they are brought into better agreement by a small modification to the spectrum. The predicted cadmium ratios are quite sensitive to the value adopted for the effective cadmium cut off energy, values of which were determined for various detectors using the Ferranti Mercury computer. The values varied over a wide range, although in every case only 40 mil. cadmium filters were used. (author)

  8. Absolute measurement of a standard thermal-neutron flux using gold-detector activation; Mesure absolue d'un flux etalon de neutrons thermiques par activation de detecteurs d'or

    Energy Technology Data Exchange (ETDEWEB)

    Paternot, Y [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-02-15

    The density of neutrons in a standard stacking is determined between the zero-energy and the cut-off energy of 1 mm thick cadmium unit using a gold detector. Its absolute activity is measured using a 4 {pi} {beta} counter calibrated for 4 {pi} {beta},{gamma} coincidence by counting strongly active sources. The correction factor F due to the disintegration process for the gold is determined experimentally. {phi}{sub 0} = N{sub 0} {sup E}Cd{sub V0} = 6495 {+-} 1.5 per cent n/cm{sup 2}/s. (author) [French] La densite de neutrons dans un empilement etalon est determinee entre l'energie zero et l'energie de coupure d'un boitier de cadmium de 1 mm d'epaisseur au moyen d'un detecteur d'or. Son activite absolue est mesuree a l'aide d'un compteur 4 {pi} {beta} etalonne en coincidence 4 {pi} {beta},{gamma} par comptage de sources fortement actives. Le facteur de correction F d au schema de desintegration de l'or est determine experimentalement. {phi}{sub 0} = N{sub 0} {sup E}Cd{sub V0} = 6495 {+-} 1.5 pour cent n/cm{sup 2}/s. (auteur)

  9. Possible dark energy imprints in the gravitational wave spectrum of mixed neutron-dark-energy stars

    Energy Technology Data Exchange (ETDEWEB)

    Yazadjiev, Stoytcho S. [Department of Theoretical Physics, Faculty of Physics, St. Kliment Ohridski University of Sofia, James Bourchier Blvd. 5, 1164 Sofia (Bulgaria); Doneva, Daniela D., E-mail: yazad@phys.uni-sofia.bg, E-mail: daniela.doneva@uni-tuebingen.de [Theoretical Astrophysics, IAAT, Eberhard-Karls University of Tübingen, Auf der Morgenstelle 10, 72076 Tübingen (Germany)

    2012-03-01

    In the present paper we study the oscillation spectrum of neutron stars containing both ordinary matter and dark energy in different proportions. Within the model we consider, the equilibrium configurations are numerically constructed and the results show that the properties of the mixed neuron-dark-energy star can differ significantly when the amount of dark energy in the stars is varied. The oscillations of the mixed neuron-dark-energy stars are studied in the Cowling approximation. As a result we find that the frequencies of the fundamental mode and the higher overtones are strongly affected by the dark energy content. This can be used in the future to detect the presence of dark energy in the neutron stars and to constrain the dark-energy models.

  10. Neutron Therapy Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Neutron Therapy Facility provides a moderate intensity, broad energy spectrum neutron beam that can be used for short term irradiations for radiobiology (cells)...

  11. Neutron dynamics of fast-spectrum dedicated cores for waste transmutation

    International Nuclear Information System (INIS)

    Massara, S.

    2002-04-01

    Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

  12. Dependence of the Ratio between the Resonance Integral and Thermal Neutron Cross Section on the Deviation of the Epithermal Neutron Spectrum from the 1/E Law

    International Nuclear Information System (INIS)

    Soliman, N.F.

    2012-01-01

    In k 0 - Neutron Activation Analysis (k 0 -NAA), the conversion from the tabulated Q 0 (ratio of the resonance integral to thermal neutron cross-section)to Q 0 (α) (α is the shape factor of the epithermal neutron flux, indicating the deviation of the epithermal neutron spectrum from the ideal 1/E shape) are calculated using a FORTRAN program. The calculations are done for most elements that can be detected by neutron activation using different values of the parameter (α) ranging from -0.1≤α≤+0.1. The obtained data are used to study the dependence of the values (α) on the irradiation position factor in (k 0 -NAA)equation for some selected isotopes differ in their resonance energy and its Q 0 values. The results show that, the irradiation factor is affective mainly for low thermal tro epithermal flux ratio f especially for Q 0 value greater than 50. so consequently determining the irradiation parameters α value is not needed for irradiation positions that rich with thermal neutron. But for high f values the irradiation position factor should be taken into account. On the other hand the constructed FORTRAN program can be used to calculate the value Q 0 (α) directly for different value of α

  13. Technical preparations for the in-vessel 14 MeV neutron calibration at JET

    International Nuclear Information System (INIS)

    Batistoni, P.; Popovichev, S.; Crowe, R.; Cufar, A.; Ghani, Z.; Keogh, K.; Peacock, A.; Price, R.; Baranov, A.; Korotkov, S.; Lykin, P.; Samoshin, A.

    2017-01-01

    Highlights: • The JET 14 MeV neutron calibration requires a neutron generator to be deployed inside the vacuum vessel by means of the remote handling system. • A neutron generator of suitable intensity and compliant with physics, remote handling and safety requirements has been identified and procured.The scientific programme of the preparatory phase devoted to fully characterizing the selected 14 MeV neutron generator is discussed. • The aim is to measure the absolute neutron emission rate within (± 5%) and the energy spectrum of emitted neutron as a function of angles. • The physics preparations, source issues, safety and engineering aspects required to calibrate directly the JET neutron detectors are discussed. - Abstract: The power output of fusion devices is measured from their neutron yields which relate directly to the fusion yield. In this paper we describe the devices and methods that have been prepared to perform a new in situ 14 MeV neutron calibration at JET in view of the new DT campaign planned at JET in the next years. The target accuracy of this calibration is ±10% as required for ITER, where a precise neutron yield measurement is important, e.g., for tritium accountancy. In this paper, the constraints and early decisions which defined the main calibration approach are discussed, e.g., the choice of 14 MeV neutron source and the deployment method. The physics preparations, source issues, safety and engineering aspects required to calibrate directly the JET neutron detectors are also discussed. The existing JET remote-handling system will be used to deploy the neutron source inside the JET vessel. For this purpose, compatible tooling and systems necessary to ensure safe and efficient deployment have been developed. The scientific programme of the preparatory phase is devoted to fully characterizing the selected 14 MeV neutron generator to be used as the calibrating source, obtain a better understanding of the limitations of the

  14. Technical preparations for the in-vessel 14 MeV neutron calibration at JET

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P., E-mail: paola.batistoni@enea.it [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044, Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Popovichev, S. [CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Crowe, R. [Remote Applications in Challenging Environments (RACE), Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Cufar, A. [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000, Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Ghani, Z. [CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Keogh, K. [Remote Applications in Challenging Environments (RACE), Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Peacock, A. [JET Exploitation Unit, Abingdon, Oxon, OX14 3DB (United Kingdom); Price, R. [Remote Applications in Challenging Environments (RACE), Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Baranov, A.; Korotkov, S.; Lykin, P.; Samoshin, A. [All-Russia Research Institute of Automatics (VNIIA), 22, Sushchevskaya str., 127055, Moscow (Russian Federation)

    2017-04-15

    Highlights: • The JET 14 MeV neutron calibration requires a neutron generator to be deployed inside the vacuum vessel by means of the remote handling system. • A neutron generator of suitable intensity and compliant with physics, remote handling and safety requirements has been identified and procured.The scientific programme of the preparatory phase devoted to fully characterizing the selected 14 MeV neutron generator is discussed. • The aim is to measure the absolute neutron emission rate within (± 5%) and the energy spectrum of emitted neutron as a function of angles. • The physics preparations, source issues, safety and engineering aspects required to calibrate directly the JET neutron detectors are discussed. - Abstract: The power output of fusion devices is measured from their neutron yields which relate directly to the fusion yield. In this paper we describe the devices and methods that have been prepared to perform a new in situ 14 MeV neutron calibration at JET in view of the new DT campaign planned at JET in the next years. The target accuracy of this calibration is ±10% as required for ITER, where a precise neutron yield measurement is important, e.g., for tritium accountancy. In this paper, the constraints and early decisions which defined the main calibration approach are discussed, e.g., the choice of 14 MeV neutron source and the deployment method. The physics preparations, source issues, safety and engineering aspects required to calibrate directly the JET neutron detectors are also discussed. The existing JET remote-handling system will be used to deploy the neutron source inside the JET vessel. For this purpose, compatible tooling and systems necessary to ensure safe and efficient deployment have been developed. The scientific programme of the preparatory phase is devoted to fully characterizing the selected 14 MeV neutron generator to be used as the calibrating source, obtain a better understanding of the limitations of the

  15. Absolute measurement of 152Eu

    International Nuclear Information System (INIS)

    Baba, Hiroshi; Baba, Sumiko; Ichikawa, Shinichi; Sekine, Toshiaki; Ishikawa, Isamu

    1981-08-01

    A new method of the absolute measurement for 152 Eu was established based on the 4πβ-γ spectroscopic anti-coincidence method. It is a coincidence counting method consisting of a 4πβ-counter and a Ge(Li) γ-ray detector, in which the effective counting efficiencies of the 4πβ-counter for β-rays, conversion electrons, and Auger electrons were obtained by taking the intensity ratios for certain γ-rays between the single spectrum and the spectrum coincident with the pulses from the 4πβ-counter. First, in order to verify the method, three different methods of the absolute measurement were performed with a prepared 60 Co source to find excellent agreement among the results deduced by them. Next, the 4πβ-γ spectroscopic coincidence measurement was applied to 152 Eu sources prepared by irradiating an enriched 151 Eu target in a reactor. The result was compared with that obtained by the γ-ray spectrometry using a 152 Eu standard source supplied by LMRI. They agreed with each other within the error of 2%. (author)

  16. 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.)

  17. Design considerations for neutron activation and neutron source strength monitors for ITER

    International Nuclear Information System (INIS)

    Barnes, C.W.; Jassby, D.L.; LeMunyan, G.; Roquemore, A.L.

    1997-01-01

    The International Thermonuclear Experimental Reactor will require highly accurate measurements of fusion power production in time, space, and energy. Spectrometers in the neutron camera could do it all, but experience has taught us that multiple methods with redundancy and complementary uncertainties are needed. Previously, conceptual designs have been presented for time-integrated neutron activation and time-dependent neutron source strength monitors, both of which will be important parts of the integrated suite of neutron diagnostics for this purpose. The primary goals of the neutron activation system are: to maintain a robust relative measure of fusion energy production with stability and wide dynamic range; to enable an accurate absolute calibration of fusion power using neutronic techniques as successfully demonstrated on JET and TFTR; and to provide a flexible system for materials testing. The greatest difficulty is that the irradiation locations need to be close to plasma with a wide field of view. The routing of the pneumatic system is difficult because of minimum radius of curvature requirements and because of the careful need for containment of the tritium and activated air. The neutron source strength system needs to provide real-time source strength vs. time with ∼1 ms resolution and wide dynamic range in a robust and reliable manner with the capability to be absolutely calibrated by in-situ neutron sources as done on TFTR, JT-60U, and JET. In this paper a more detailed look at the expected neutron flux field around ITER is folded into a more complete design of the fission chamber system

  18. Experimental research of neutron yield and spectrum from deuterium gas-puff z-pinch on the GIT-12 generator at current above 2 MA

    Science.gov (United States)

    Cherdizov, R. K.; Fursov, F. I.; Kokshenev, V. A.; Kurmaev, N. E.; Labetsky, A. Yu; Ratakhin, N. A.; Shishlov, A. V.; Cikhardt, J.; Cikhardtova, B.; Klir, D.; Kravarik, J.; Kubes, P.; Rezac, K.; Dudkin, G. N.; Garapatsky, A. A.; Padalko, V. N.; Varlachev, V. A.

    2017-05-01

    The Z-pinch experiments with deuterium gas-puff surrounded by an outer plasma shell were carried out on the GIT-12 generator (Tomsk, Russia) at currents of 2 MA. The plasma shell consisting of hydrogen and carbon ions was formed by 48 plasma guns. The deuterium gas-puff was created by a fast electromagnetic valve. This configuration provides an efficient mode of the neutron production in DD reaction, and the neutron yield reaches a value above 1012 neutrons per shot. Neutron diagnostics included scintillation TOF detectors for determination of the neutron energy spectrum, bubble detectors BD-PND, a silver activation detector, and several activation samples for determination of the neutron yield analysed by a Sodium Iodide (NaI) and a high-purity Germanium (HPGe) detectors. Using this neutron diagnostic complex, we measured the total neutron yield and amount of high-energy neutrons.

  19. Neutron spectra produced by moderating an isotopic neutron source

    International Nuclear Information System (INIS)

    Carrillo Nunnez, Aureliano; Vega Carrillo, Hector Rene

    2001-01-01

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

  20. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.; Luis L, M. A. [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Division de Ciencias Basicas, Av. San Pablo No. 180, Col. Reynosa Tamaulipas, 02200 Mexico D. F. (Mexico); Raya A, R.; Cruz G, H. S., E-mail: roberto.raya@inin.gob.mx [ININ, Departamento del Reactor, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10{sup -10} to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of {sup 197}Au, {sup 58}Ni, {sup 115}In, {sup 24}Mg, {sup 27}Al, {sup 58}Fe, {sup 59}Co and {sup 63}Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

  1. A prototype detector using the neutron image intensifier and multi-anode type photomultiplier tube for pulsed neutron imaging

    International Nuclear Information System (INIS)

    Ishikawa, Hirotaku; Sato, Hirotaka; Hara, Kaoru Y.; Kamiyama, Takashi

    2016-01-01

    We developed a neutron two-dimensional (2-D) detector for pulsed neutron imaging as a prototype detector, which was composed of a neutron image intensifier and a multi-anode type photomultiplier tube. A neutron transmission spectrum of α-Fe plate was measured by the prototype detector, and compared with the one measured by a typical neutron 2-D detector. The spectrum was in reasonable agreement with the one measured by the typical detector in the neutron wavelength region above 0.15 nm. In addition, a neutron transmission image of a cadmium indicator was obtained by the prototype detector. The usefulness of the prototype detector for pulsed neutron imaging was demonstrated. (author)

  2. DS86 neutron dose. Monte Carlo analysis for depth profile of {sup 152}Eu activity in a large stone sample

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Satoru; Hoshi, Masaharu; Takada, Jun [Hiroshima Univ. (Japan). Research Inst. for Radiation Biology and Medicine; Iwatani, Kazuo; Oka, Takamitsu; Shizuma, Kiyoshi; Imanaka, Tetsuji; Fujita, Shoichiro; Hasai, Hiromi

    1999-06-01

    The depth profile of {sup 152}Eu activity induced in a large granite stone pillar by Hiroshima atomic bomb neutrons was calculated by a Monte Carlo N-Particle Transport Code (MCNP). The pillar was on the Motoyasu Bridge, located at a distance of 132 m (WSW) from the hypocenter. It was a square column with a horizontal sectional size of 82.5 cm x 82.5 cm and height of 179 cm. Twenty-one cells from the north to south surface at the central height of the column were specified for the calculation and {sup 152}Eu activities for each cell were calculated. The incident neutron spectrum was assumed to be the angular fluence data of the Dosimetry System 1986 (DS86). The angular dependence of the spectrum was taken into account by dividing the whole solid angle into twenty-six directions. The calculated depth profile of specific activity did not agree with the measured profile. A discrepancy was found in the absolute values at each depth with a mean multiplication factor of 0.58 and also in the shape of the relative profile. The results indicated that a reassessment of the neutron energy spectrum in DS86 is required for correct dose estimation. (author)

  3. The effect of neutron spectrum on the mechanical and physical properties of pure copper and copper alloys

    International Nuclear Information System (INIS)

    Fabritsiev, S.A.; Pokrovsky, A.S.; Sandakov, V.A.; Zinkle, S.J.; Rowcliffe, A.F.; Edwards, D.J.; Garner, F.A.; Singh, B.N.; Barabash, V.R.

    1996-01-01

    The electrical resistivity and tensile properties of copper and oxide dispersion strengthened (DS) copper alloys have been measured before and after fission neutron irradiation to damage levels of 0.5 to 5 displacements per atom (dps) at ∼100 to 400 degrees C. Some of the specimens were irradiated inside a 1.5 mm Cd shroud in order to reduce the thermal neutron flux. The electrical resistivity data could be separated into two components, a solid transmutation component Δρ tr which was proportional to thermal neutron fluence and a radiation defect component Δρ rd which was independent of the displacement dose. The saturation value for Δρ rd was ∼1.2 nanohm-meters for pure copper and ∼1.6 nanohm-meters for the DS copper alloys irradiated at 100 degrees C in positions with a fast-to-thermal neutron flux ratio of 5. Considerable radiation hardening was observed in all specimens at irradiation temperatures below 200 degrees C. The yield strength was relatively insensitive to neutron spectrum in specimens strengthened by dispersoids or cold- working. 17 refs., 7 figs., 1 tab

  4. Absolute values of inelastic neutron scattering cross-sections calculated with account taken of the pre-equilibrium mechanism

    International Nuclear Information System (INIS)

    Jahn, H.

    1980-01-01

    Absolute values of secondary energy-dependent inelastic neutron scattering cross sections can be calculated either with the master equation pre-equilibrium formalism of Cline and Blann or with Blann's more recent geometry-dependent hybrid model. The master equation formalism was used at Dubna and Dresden to reproduce experimental results for 14 MeV incident energy. The geometry-dependent hybrid model was used at Karlsruhe to cover for a number of materials the whole range from 5 to 14 MeV incident energy and to reproduce smoothed experimental spectra at 7.45 and 14 MeV. Only the geometry-dependent hybrid model accounts for scattering in the diffuse nuclear surface and thus for a certain average over the direct interaction. It is also free of any fit parameters other than those of the usual optical model. The master equation calculations, on the other hand, are based on nucleon-nucleon scattering cross sections inserted into the high-energy approximation of Kikuchi and Kawai for the intranuclear transition rate. Other approaches require either mass- or energy-dependent or more global fit parameters for a satisfactory reproduction of experimental results, but a genuine prediction of the incident-energy dependence of the inelastic neutron cross section, especially below 14 MeV, is needed for transport and shielding calculations for instance in connection with fusion reactor design studies. (author)

  5. On the spectrum of the one-speed slab-geometry discrete ordinates operator in neutron transport theory

    International Nuclear Information System (INIS)

    Abreu, Marcos Pimenta de

    1998-01-01

    We describe a numerical method applied to the first-order form of one-speed slab-geometry discrete ordinates equations modelling time-independent neutron transport problems with anisotropic scattering, with no interior source and defined in a nonmultiplying homogeneous host medium. Our numerical method is concerned with the generation of the spectrum and of a vector basis for the null space of the one-speed slab-geometry discrete ordinates operator. Moreover, it allows us to overcome the difficulties introduced in previous methods by anisotropic scattering and by angular quadrature sets of high order. To illustrate the positive features of our numerical method, we present numerical results for one-speed slab-geometry neutron transport model problems with anisotropic scattering

  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. Advancing Absolute Calibration for JWST and Other Applications

    Science.gov (United States)

    Rieke, George; Bohlin, Ralph; Boyajian, Tabetha; Carey, Sean; Casagrande, Luca; Deustua, Susana; Gordon, Karl; Kraemer, Kathleen; Marengo, Massimo; Schlawin, Everett; Su, Kate; Sloan, Greg; Volk, Kevin

    2017-10-01

    We propose to exploit the unique optical stability of the Spitzer telescope, along with that of IRAC, to (1) transfer the accurate absolute calibration obtained with MSX on very bright stars directly to two reference stars within the dynamic range of the JWST imagers (and of other modern instrumentation); (2) establish a second accurate absolute calibration based on the absolutely calibrated spectrum of the sun, transferred onto the astronomical system via alpha Cen A; and (3) provide accurate infrared measurements for the 11 (of 15) highest priority stars with no such data but with accurate interferometrically measured diameters, allowing us to optimize determinations of effective temperatures using the infrared flux method and thus to extend the accurate absolute calibration spectrally. This program is integral to plans for an accurate absolute calibration of JWST and will also provide a valuable Spitzer legacy.

  8. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

    International Nuclear Information System (INIS)

    Selby, H.D.; Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C.

    2010-01-01

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99 Mo, 95 Zr, 137 Cs, 140 Ba, 141,143 Ce, and 147 Nd. Modest incident-energy dependence exists for the 147 Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ∼5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except

  9. Preliminary engineering assessment of the HCLL and HCPB Neutron Activation System

    Energy Technology Data Exchange (ETDEWEB)

    Calderoni, Pattrick; Leichtle, Dieter [Fusion for Energy, Barcelona, (Spain); Angelone, Maurizio [ENEA, Unita Tecnica Fusione, Frascati, (Italy); Klix, Axel [KIT, Eggenstein-Leopoldshafen, (Germany)

    2015-07-01

    The Neutron Activation System (NAS) is one of the four types of neutronics sensors considered for the testing of the HCLL and HCPB Test Blanket Module (TBM) in ITER. It measures the absolute neutron flux intensity with information on the neutron spectrum in selected positions of the TBM. The working principle of the NAS is as follows: the system moves small activation probes (capsules) into selected positions in the TBM (irradiation ends) by means of pneumatic transport with pressurized helium gas; the capsules are irradiated for a selected period, depending on their materials composition (several tens of seconds up to the full plasma pulse length); immediately after the irradiation they are extracted and transported to a gamma spectrometer by means of the same pneumatic transport system; the gamma spectrometer determines the induced gamma activity; the neutron flux and neutron fluence is calculated from the measured gamma activity and the known activation cross section of the materials in the activation probe; after the measurement the capsule is sent either to a disposal or storage (for later measurement). This paper summarizes the results of the feasibility assessment of the TBM NAS in the conceptual design phase, including design justification, identification of requirements based on the expected operating conditions in ITER and preliminary engineering assessment of the activation materials, irradiation ends integration in the modules design and the counting station. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

  11. Improved Bonner sphere neutron spectrometry measurements for the nuclear industry

    Science.gov (United States)

    Roberts, N. J.; Thomas, D. J.; Visser, T. P. P.

    2017-11-01

    A novel, two-stage approach has been developed for producing the a priori spectrum for Bonner sphere unfolding in a case where neutrons are produced by spontaneous fission and (α,n) reactions, e.g. in UF6. The code SOURCES 4C is first used to obtain the energy spectrum of the neutrons inside the material, which is then fed into a MCNP model of the entire geometry to derive the neutron spectrum at the location of the Bonner sphere. Using this as the a priori spectrum produces a much more detailed unfolded Bonner sphere spectrum retaining fine structure from the calculation that would not be present if a simple estimated spectrum had been used as the a priori spectrum. This is illustrated using a Bonner sphere measurement of the neutron energy spectrum produced by a 48Y cylinder of UF6. From the unfolded spectrum an estimate has been made of the neutron ambient dose equivalent, i.e. the quantity which a neutron survey instrument should measure. The difference in the ambient dose equivalent of the unfolded spectrum is over 10% when using the novel approach instead of using a simpler estimate consisting of a single high energy peak, 1/E continuum, and thermal peak.

  12. Measurement of photoneutron spectrum at Pohang Neutron Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G.N.; Kovalchuk, V.; Lee, Y.S.; Skoy, V.; Cho, M.H.; Ko, I.S.; Namkung, W. [POSTECH, Pohang Accelerator Laboratory, Pohang, Kyungbuk (Korea)

    2001-03-01

    Pohang Neutron Facility, which is the pulsed neutron facility based on the 100-MeV electron linear accelerator, was constructed for nuclear data production in Korea. The Pohang Neutron Facility consists of an electron linear accelerator, a water-cooled Ta target with a water moderator and a time-of-flight path with an 11 m length. The neutron energy spectra are measured for different water levels inside the moderator and compared with the MCNP calculation. The optimum size of the water moderator is determined on the base of this result. The time dependent spectra of neutrons in the water moderator are investigated with the MCNP calculation. (author)

  13. On the Proton Spectrum in Free Neutron beta-decay

    CERN Document Server

    Bunatian, G G

    2000-01-01

    We consider the calculations which are appropriate to acquire with a high precision, of ~1% or better, the general characteristics of weak interactions from the experiments on the free neutron beta-decay; the principle emphasis is placed on the phenomena associated with the recoil of protons. The part played by electromagnetic interactions in beta-decay is visualized, with special attention drawn to the influence of the gamma-radiation on the momentum distribution of the particles in the final state. The effect of electromagnetic interactions on the proton recoil spectrum is studied, in the light of the experiments which are carried out and planned for now. The results of the calculations, which are to be confronted with the experimental data, are presented upright in terms of the effective Lagrangian underlying the inquiry. Owing to electromagnetic interactions, the corrections to the energy distribution of protons prove to amount to the value of a few per cent. Nowadays, this is substantial to obtain with a...

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2008-09-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  16. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2008-01-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  17. Rapid method of calculating the fluence and spectrum of neutrons from a critical assembly and the resulting dose

    International Nuclear Information System (INIS)

    Bessis, J.

    1977-01-01

    The proposed calculation method is unsophisticated but rapid. The first part (computer code CRITIC), which is based on the Fermi age equation, evaluates the number of neutrons per fission emitted from a moderated critical assembly and their energy spectrum. The second part (computer code NARCISSE), which uses the current albedo for concrete, evaluates the product of neutron reflection on the walls and calculates the fluence resulting at any point in the room and its energy distribution by 21 groups. The results obtained are shown to compare satisfactorily with those obtained through the use of a Monte Carlo program

  18. A gamma-ray discriminating neutron scintillator

    International Nuclear Information System (INIS)

    Eschbach, P.A.; Miller, S.D.; Cole, M.C.

    1994-01-01

    A neutron scintillator has been developed at Pacific Northwest Laboratory which responds directly to as little as 10 mrem/hour dose equivalent rate fast neutron fields. The scintillator is composed of CaF 2 :Eu or of NaI grains within a silicone rubber or polystyrene matrix, respectively. Neutrons colliding with the plastic matrix provide knockon protons, which in turn deposit energy within the grains of phosphor to produce pulses of light. Neutron interactions are discriminated from gamma-ray events on the basis of pulse height. Unlike NE-213 liquid scintillators, this solid scintillator requires no pulseshape discrimination and therefore requires less hardware. Neutron events are anywhere from two to three times larger than the gamma-ray exposures are compared to 0.7 MeV gamma-ray exposures. The CaF 2 :Eu/silicone rubber scintillator is nearly optically transparent, and can be made into a very sizable detector (4 cm x 1.5 cm) without degrading pulse height. This CaF 2 :Eu scintillator has been observed to have an absolute efficiency of 0.1% when exposed to 5-MeV accelerator-generated neutrons (where the absolute efficiency is the ratio of observed neutron events divided by the number of fast neutrons striking the detector)

  19. Determination of intensity and energy spectrum of neutrons by bombardment of thallium-203 thick target and its copper substrate with 28.5 MeV protons

    International Nuclear Information System (INIS)

    Hajiloo, N.; Raisali, Gh.; Hamidi, S.; Aslani, Gh.

    2007-01-01

    In this research we have determined neutrons spectrum and the intensity that produced from thallium target bombardment. We have applied SRIM and ALICE computer codes to thallium target and its copper substrate for 145 μA of 28.5 MeV incident proton beam from cyclotron Cyclone30. Because of the energy degradation of protons while passing through the thallium target and its copper substrate, the average energy of protons in different depths has been calculated by using SRIM computer code. Then, by applying ALICE computer code for each sub-layer, the neutron production cross sections and their energy spectrum have been calculated to determine the total neutron intensity and spectrum. Using the calculated neutron intensity of 1.22x10 13 n/s as the source, the equivalent dose rate at the distance 6 meters from the target has been calculated by MCNP computer code and the result has been compared with the measured value. The Pb 201 activity has also been calculated as 13.5 Curies. The measured Pb 201 activity by Curie meter CAPINTEC CRC-712 is 13.1 Ci which is in reasonable agreement with the calculated value, bearing in mind the uncertainties in the proposed models and the measurements

  20. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.

    2014-07-01

    This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au{sup 197}, Ni{sup 58}, In{sup 115}, Mg{sup 24}, Al{sup 27}, Fe{sup 58}, Co{sup 59} and Cu{sup 63}, they were selected to cover the full range the energies (10{sup -10} to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or

  1. Absolute isotopic abundances of Ti in meteorites

    International Nuclear Information System (INIS)

    Niederer, F.R.; Papanastassiou, D.A.; Wasserburg, G.J.

    1985-01-01

    The absolute isotope abundance of Ti has been determined in Ca-Al-rich inclusions from the Allende and Leoville meteorites and in samples of whole meteorites. The absolute Ti isotope abundances differ by a significant mass dependent isotope fractionation transformation from the previously reported abundances, which were normalized for fractionation using 46 Ti/ 48 Ti. Therefore, the absolute compositions define distinct nucleosynthetic components from those previously identified or reflect the existence of significant mass dependent isotope fractionation in nature. We provide a general formalism for determining the possible isotope compositions of the exotic Ti from the measured composition, for different values of isotope fractionation in nature and for different mixing ratios of the exotic and normal components. The absolute Ti and Ca isotopic compositions still support the correlation of 50 Ti and 48 Ca effects in the FUN inclusions and imply contributions from neutron-rich equilibrium or quasi-equilibrium nucleosynthesis. The present identification of endemic effects at 46 Ti, for the absolute composition, implies a shortfall of an explosive-oxygen component or reflects significant isotope fractionation. Additional nucleosynthetic components are required by 47 Ti and 49 Ti effects. Components are also defined in which 48 Ti is enhanced. Results are given and discussed. (author)

  2. Calibration issues for neutron diagnostics

    International Nuclear Information System (INIS)

    Sadler, G.J.; Adams, J.M.; Barnes, C.W.

    1997-01-01

    The performance of diagnostic systems are limited by their weakest constituents, including their calibration issues. Neutron diagnostics are notorious for problems encountered while determining their absolute calibrations, due mainly to the nature of the neutron transport problem. In order to facilitate the determination of an accurate and precise calibration, the diagnostic design should be such as to minimize the scattered neutron flux. ITER will use a comprehensive set of neutron diagnostics--comprising radial and vertical neutron cameras, neutron spectrometers, a neutron activation system and internal and external fission chambers--to provide accurate measurements of fusion power and power densities as a function of time. The calibration of such an important diagnostic system merits careful consideration. Some thoughts have already been given to this subject during the conceptual design phase in relation to the time-integrated neutron activation and time-dependent neutron yield monitors. However, no overall calibration strategy has been worked out so far. This paper represents a first attempt to address this vital issue. Experience gained from present large tokamaks (JET, TFTR and JT60U) and proposals for ITER are reviewed. The need to use a 14-MeV neutron generator as opposed to radioactive sources for in-situ calibration of D-T diagnostics will be stressed. It is clear that the overall absolute determination of fusion power will have to rely on a combination of nuclear measuring techniques, for which the provision of accurate and independent calibrations will constitute an ongoing process as ITER moves from one phase of operation to the next

  3. Unfolding of neutron spectra from Godiva type critical assemblies

    International Nuclear Information System (INIS)

    Harvey, J.T.; Meason, J.L.; Wright, H.L.

    1976-01-01

    The results from three experiments conducted at the White Sands Missile Range Fast Burst Reactor Facility are discussed. The experiments were designed to measure the ''free-field'' neutron leakage spectrum and the neutron spectra from mildly perturbed environments. SAND-II was used to calculate the neutron spectrum utilizing several different trial input spectra for each experiment. Comparisons are made between the unfolded neutron spectrum for each trial input on the basis of the following parameters: average neutron energy (above 10 KeV), integral fluence (above 10 KeV), spectral index and the hardness parameter, phi/sub eq//phi

  4. Rigidity spectrum of Forbush decrease calculated by neutron monitors data corrected and uncorrected for geomagnetic disturbances

    International Nuclear Information System (INIS)

    Alania, M V; Wawrzynczak, A; Sdobnov, V E; Kravtsova, M V

    2013-01-01

    Forbush decreases (Fd) of the galactic cosmic ray (GCR) intensity and geomagnetic storms are observed almost at the same time. Geomagnetic storm is a reason of significant disturbances of the magnetic cut off rigidity causing the distortion of the time profile of the Fd of the GCR intensity. We show some differences in the temporal changes of the rigidity spectra of Fd calculated by neutron monitors experimental data corrected and uncorrected for the changes of the geomagnetic cut off rigidity. Nevertheless, the general features of the temporal changes of the rigidity spectrum of Fd maintain as it was found in our previous investigations. Namely, at the beginning phase of Fd rigidity spectrum is relatively soft and gradually becomes hard up to reaching the minimum level of the GCR intensity; then the rigidity spectrum gradually becomes soft during the recovery phase of Fd. We also confirm that for the established temporal profiles of the rigidity spectrum of Fd a structural changes of the interplanetary magnetic field turbulence in the range of frequencies, 10 −-6 ÷10 −-5 Hz are responsible.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  6. Performance analysis of fusion nuclear-data benchmark experiments for light to heavy materials in MeV energy region with a neutron spectrum shifter

    International Nuclear Information System (INIS)

    Murata, Isao; Ohta, Masayuki; Miyamaru, Hiroyuki; Kondo, Keitaro; Yoshida, Shigeo; Iida, Toshiyuki; Ochiai, Kentaro; Konno, Chikara

    2011-01-01

    Nuclear data are indispensable for development of fusion reactor candidate materials. However, benchmarking of the nuclear data in MeV energy region is not yet adequate. In the present study, benchmark performance in the MeV energy region was investigated theoretically for experiments by using a 14 MeV neutron source. We carried out a systematical analysis for light to heavy materials. As a result, the benchmark performance for the neutron spectrum was confirmed to be acceptable, while for gamma-rays it was not sufficiently accurate. Consequently, a spectrum shifter has to be applied. Beryllium had the best performance as a shifter. Moreover, a preliminary examination of whether it is really acceptable that only the spectrum before the last collision is considered in the benchmark performance analysis. It was pointed out that not only the last collision but also earlier collisions should be considered equally in the benchmark performance analysis.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  8. Neutron spectrometer for improved SNM search.

    Energy Technology Data Exchange (ETDEWEB)

    Vance, Andrew L.; Aigeldinger, Georg

    2007-03-01

    With the exception of large laboratory devices with very low sensitivities, a neutron spectrometer have not been built for fission neutrons such as those emitted by special nuclear materials (SNM). The goal of this work was to use a technique known as Capture Gated Neutron Spectrometry to develop a solid-state device with this functionality. This required modifications to trans-stilbene, a known solid-state scintillator. To provide a neutron capture signal we added lithium to this material. This unique triggering signal allowed identification of neutrons that lose all of their energy in the detector, eliminating uncertainties that arise due to partial energy depositions. We successfully implemented a capture gated neutron spectrometer and were able to distinguish an SNM like fission spectrum from a spectrum stemming from a benign neutron source.

  9. Fast-neutron detecting system with n, γ discrimination

    International Nuclear Information System (INIS)

    Ouyang Xiaoping; Huang Bao; Cao Jinyun

    1997-11-01

    In the present work, a new type neutron detecting system is reported, which can absolutely measure neutron parameters in n + γ mixed fields and has a long continuance of static high vacuum of 10 -4 Pa. The detecting system, with middle neutron-detecting sensitivity, short time response and big linear current output, has applied successfully in pulsed neutron beam measurement

  10. The Thermal Neutron Beam Option for NECTAR at MLZ

    Science.gov (United States)

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

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

  11. Methods of neutron spectrum calculation from measured reaction velocities in SAIPS

    International Nuclear Information System (INIS)

    Berzonis, M.A.; Bondars, Kh.Ya.

    1981-01-01

    When a user (physicist) needs to perform calculations, he faces a number of problems: obtaining or generating calculation programs, comparing these programs, generating a library of reference spectra, study of calculated spectra and so on. This means routine work which is duplicated in many laboratories. To help solve these problems a computerized information system called SAIPS has been developed, some aspects of which are dealt with in references. The present paper gives a short description of data input into SAIPS and the basic principles of its utilization. SAIPS is based on the ES 1022 computer controlled by the operational system OS ES version 4.1. It contains the programs needed for unfolding spectra, neutron cross-section and reference spectrum libraries and the software for the main system and for computerized calculations

  12. Thermal neutron polarisation

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  14. Influence of Neutron Spectra Unfolding Method on Fast Neutron Dose Determination

    International Nuclear Information System (INIS)

    Marinkovic, P.

    1991-01-01

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

  15. 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.

  16. Optimization of artificial neural networks for the reconstruction of the neutrons spectrum and their equivalent doses; Optimizacion de redes neuronales artificiales para la reconstruccion del espectro de neutrones y sus dosis equivalentes

    Energy Technology Data Exchange (ETDEWEB)

    Reyes A, A.; Ortiz R, J. M.; Reyes H, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R., E-mail: art8291@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Lopez Velarde No. 801, Col. Centro, 98000 Zacatecas (Mexico)

    2014-08-15

    In this work was used the robust design methodology of artificial neural networks to determine a good topology of net able to solve with efficiency the problems of neutrons spectrometry and dosimetry. For the design of the topology of optimized net 36 different net architectures based on an orthogonal arrangement with a configuration L{sub 9}(3{sup 4}), L{sub 4}(3{sup 2}) were trained. For the training of the neural networks, was used a computer code developed in the ambient of Mat lab programming, which automates the process and analysis of the information, reducing the time used in this activity considerably for the investigator. For the training of the propagation nets forward was utilized a neutrons spectrum compendium published by the International Atomic Energy Agency, where of the total 80% was used for the training and 20% for the test, it trained with an inverse propagation algorithm being the entrance data the count rates corresponding to the 7 spheres of the spectrometric system of Bonner spheres, as exit data, the neural network obtains the neutrons spectrum expressed in 60 energy groups and are calculated of simultaneous way 15 dosimetric quantities. (Author)

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-15

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

  19. 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

  20. On the absolute calibration of a DT fusion neutron yield diagnostic

    Directory of Open Access Journals (Sweden)

    Ruiz C.L.

    2013-11-01

    Full Text Available Recent advances in Inertial Confinement Fusion (ICF experiments at Lawrence Livermore National Laboratory's National Ignition Facility (NIF have underscored the need for accurate total yield measurements of DT neutrons because yield measurements provide a measure of the predicted performance of the experiments. Future gas-puff DT experiments at Sandia National Laboratory's Z facility will also require similar measurements. For ICF DT experiments, the standard technique for measuring the neutron (14.1 MeV yield, counts the activity (counts/minute induced in irradiated copper samples. This activity occurs by the 63Cu(n,2n62Cu reaction where 62Cu decays by positrons (β+ with a half-life of 9.67 minutes. The calibrations discussed here employ the associated-particle method (APM, where the α (4He particles from the T(d,n4He reaction are measured to infer neutron fluxes on a copper sample. The flux induces 62Cu activity, measured in a coincidence counting system. The method leads to a relationship between a DT neutron yield and copper activity known as the F-factor. The goal in future experiments is to apply this calibration to measure the yield at NIF with a combined uncertainty approaching 5%.

  1. 20070607 NATO Advanced Study Institute on the Electromagnetic Spectrum of Neutron Stars Marmaris, Turkey 07 - 18 Jun 2004 2004 marmaris20040607 TR 20040618

    CERN Document Server

    Baykal, Altan; Inam, Sitki C; Grebenev, Sergei

    2005-01-01

    Neutron stars hold a central place in astrophysics, not only because they are made up of the most extreme states of the condensed matter, but also because they are, along with white dwarfs and black holes, one of the stable configurations that stars reach at the end of stellar evolution. Neutron stars posses the highest rotation rates and strongest magnetic fields among all stars. They radiate prolifically, in high energy electromagnetic radiation and in the radio band. This book is devoted to the selected lectures presented in the 6th NATO-ASI series entitled "The Electromagnetic Spectrum of Neutron Stars" in Marmaris, Turkey, on 7-18 June 2004. This ASI is devoted to the spectral properties of neutron stars. Spectral observations of neutron stars help us to understand the magnetospheric emission processes of isolated radio pulsars and the emission processes of accreting neutron stars. This volume includes spectral information from the neutron stars in broadest sense, namely neutrino and gravitational radiat...

  2. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2009-08-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  3. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2009-01-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  4. Neutron activation analysis of absolutely-dated tree rings

    International Nuclear Information System (INIS)

    Uenlue, K.; Hauck, D.K.; Kuniholm, P.I.; Chiment, J.J.

    2005-01-01

    Gold concentration was determined for dendrochronologically-dated wood samples using neutron activation analysis (NAA) and correlation sought with known environmental changes, e.g., volcanic activities, during historic periods. Uptake of gold is sensitive to soil pH for many plants. Data presented are from a single, cross-dated tree that grew in Greece. Using NAA, gold was measured with parts-per-billion sensitivity in individual tree rings from 1411 to 1988 AD. (author)

  5. Neutron converter at reactor RB; Konvertor neutrona na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Sotic, P; Ninkovic, M; Pesic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1977-07-01

    A neutron converter at Reactor RB in the 'Boris Kidric' Institute of Nuclear Sciences - Vinca has been constructed. Preliminary measurements have been shown that the converted neutron spectrum is very similar to the fission neutron spectrum. For the same integral reactor power, the measured neutron radiation dose has been for about ten times larger with the neutron converter. The neutron converter offers wide possibilities, as in investigations in the reactor physics, where the fission neutron spectra have been required, as well as in the field of neutron dosimetry and biological irradiations (author)

  6. Integral test on activation cross section of tag gas nuclides using fast neutron spectrum fields

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Takafumi; Suzuki, Soju [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-03-01

    Activation cross sections of tag gas nuclides, which will be used for the failed fuel detection and location in FBR plants, were evaluated by the irradiation tests in the fast neutron spectrum fields in JOYO and YAYOI. The comparison of their measured radioactivities and the calculated values using the JENDL-3.2 cross section set showed that the C/E values ranged from 0.8 to 2.8 for the calibration tests in YAYOI and that the present accuracies of these cross sections were confirmed. (author)

  7. Study of the influence of the fast neutron spectrum on the production of defects in solids and liquids

    International Nuclear Information System (INIS)

    Mas, P.; Droulers, Y.

    1964-01-01

    In the first part of this work a calculation has been made of the number of defects formed in graphite by a given neutron flux having various spectral distributions. The defect formation function is that of KINCHIN and PEASE; its formulation is briefly given. An efficiency function is then defined for a fast neutron spectrum. This defects produced in a light water reactor and those produced in a graphite reactor. Finally an application of this method is given for comparing the defect forming tendency in graphite in the case of the reactor Melusine and of the reactor G-2 and G-3. In the second part are calculated the integrals for the energy release brought about by fast neutrons in carbon oxygen and hydrogen. In a region of 25 cm around the core of a swimming-pool type reactor these energy release integrals are approximately proportional to the neutron flux above 1 MeV. The determination of the energy released as a result of the passage of neutrons in organic liquids can therefore be reduced to the measurement of the flux above 1 MeV for the real spectral distribution. A calorimetric verification has been carried out in the case of water. (authors) [fr

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

    International Nuclear Information System (INIS)

    Kiyanagi, Yoshiaki

    2012-01-01

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

  9. 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.

  10. Effect of Fast Neutron to MA/PU Burning/Transmutation Characteristic Using a Fast Reactor

    International Nuclear Information System (INIS)

    Marsodi; Lasman, As Natio; Kimamoto, A.; Marsongkohadi; Zaki, S.

    2003-01-01

    MA/Pu burning/transmutation has been studied and evaluated using fast neutrons. Generally, neutron density at this fast burner reactor and transmutation has spectrum energy level around 0.2 MeV with wide enough variation, i.e. from low neutron spectrum to its peak is 0.2 MeV. This neutron spectrum energy level depends on the kind of cooler material or fuel used. Neutron spectrum higher than fast power reactor neutron spectrum is found by means of changing oxide fuel by metallic fuel and changing natrium cooler material by metallic or gas cooler material. This evaluation is conducted by various variations in accordance with the kind of fuel or cooler, MA/Pu fractions and fuel comparison fraction with respect to its cooler in order to get better neutron usage and MA/Pu burning speed. Reactor calculation evaluation in this paper was conducted with 26-group nuclear data cross section energy spectrum. The main purpose of the discussion is to know the effect of fast neutrons to burning/transmutation MA/Pu using fast neutrons

  11. Measurement of neutron-production double-differential cross sections for continuous neutron-incidence reaction up to 100 MeV

    International Nuclear Information System (INIS)

    Kunieda, Satoshi; Watanabe, Takehito; Shigyo, Nobuhiro; Ishibashi, Kenji; Satoh, Daiki; Nakamura, Takashi; Haight, Robert C.

    2004-01-01

    The inclusive measurements of neutron-incident neutron-production double-differential cross sections in intermediate energy range is now being carried out. Spallation neutrons are used as incident particles. As a part of this, the experiment was performed by using of NE213 liquid organic scintillators to detect outgoing-neutrons. Incident-neutron energy was determined by time-of-flight technique, and outgoing-neutron energy spectrum was derived by unfolding light-output spectrum of NE213 with response functions calculated by SCINFUL-R. Preliminary cross sections were obtained up to about 100 MeV, and were compared with calculations by the GNASH code. It is hoped to get pure measurements by using measured response functions for our detectors used in this study. (author)

  12. OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.; Roche, Kenneth J.; Kurtz, Richard J.; Wirth, Brian D.

    2015-12-31

    The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

  13. Determination of the neutron spectrum in the well Naiade attached to the reactor Nereide

    International Nuclear Information System (INIS)

    Capgras, Andree; Clement, Christophe; Sueur, Maurice.

    1977-11-01

    The spectral distribution of neutrons in the centre of the well Naiade attached to the Fontenay-aux-Roses reactor Nereide is studied. In the thermal, epithermal and over 2.2 MeV regions, activation detectors are used: 197 Au and 55 Mn (bare and under cadmium), and 58 Ni. In the energy band from a few keV to 2.2 MeV two recoil proton proportional counters are employed. Under these conditions the whole spectrum is studied, but some comments are made on the difficulties of interpreting the results obtained by either of these methods [fr

  14. Absolute calibration in vivo measurement systems

    International Nuclear Information System (INIS)

    Kruchten, D.A.; Hickman, D.P.

    1991-02-01

    Lawrence Livermore National Laboratory (LLNL) is currently investigating a new method for obtaining absolute calibration factors for radiation measurement systems used to measure internally deposited radionuclides in vivo. Absolute calibration of in vivo measurement systems will eliminate the need to generate a series of human surrogate structures (i.e., phantoms) for calibrating in vivo measurement systems. The absolute calibration of in vivo measurement systems utilizes magnetic resonance imaging (MRI) to define physiological structure, size, and composition. The MRI image provides a digitized representation of the physiological structure, which allows for any mathematical distribution of radionuclides within the body. Using Monte Carlo transport codes, the emission spectrum from the body is predicted. The in vivo measurement equipment is calibrated using the Monte Carlo code and adjusting for the intrinsic properties of the detection system. The calibration factors are verified using measurements of existing phantoms and previously obtained measurements of human volunteers. 8 refs

  15. Many channel spectrum unfolding

    International Nuclear Information System (INIS)

    Najzer, M.; Glumac, B.; Pauko, M.

    1980-01-01

    The principle of the ITER unfolding code as used for the many channel spectrum unfolding is described. Its unfolding ability is tested on seven typical neutron spectra. The effect of the initial spectrum approximation upon the solution is discussed

  16. Determination of neutron energy spectrum at a pneumatic rabbit station of a typical swimming pool type material test research reactor

    International Nuclear Information System (INIS)

    Malkawi, S.R.; Ahmad, N.

    2002-01-01

    The method of multiple foil activation was used to measure the neutron energy spectrum, experimentally, at a rabbit station of Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type material test research reactor. The computer codes MSITER and SANDBP were used to adjust the spectrum. The pre-information required by the adjustment codes was obtained by modelling the core and its surroundings in three-dimensions by using the one dimensional transport theory code WIMS-D/4 and the multidimensional finite difference diffusion theory code CITATION. The input spectrum covariance information required by MSITER code was also calculated from the CITATION output. A comparison between calculated and adjusted spectra shows a good agreement

  17. Deuterium-tritium neutron yield measurements with the 4.5 m neutron-time-of-flight detectors at NIF.

    Science.gov (United States)

    Moran, M J; Bond, E J; Clancy, T J; Eckart, M J; Khater, H Y; Glebov, V Yu

    2012-10-01

    The first several campaigns of laser fusion experiments at the National Ignition Facility (NIF) included a family of high-sensitivity scintillator∕photodetector neutron-time-of-flight (nTOF) detectors for measuring deuterium-deuterium (DD) and DT neutron yields. The detectors provided consistent neutron yield (Y(n)) measurements from below 10(9) (DD) to nearly 10(15) (DT). The detectors initially demonstrated detector-to-detector Y(n) precisions better than 5%, but lacked in situ absolute calibrations. Recent experiments at NIF now have provided in situ DT yield calibration data that establish the absolute sensitivity of the 4.5 m differential tissue harmonic imaging (DTHI) detector with an accuracy of ± 10% and precision of ± 1%. The 4.5 m nTOF calibration measurements also have helped to establish improved detector impulse response functions and data analysis methods, which have contributed to improving the accuracy of the Y(n) measurements. These advances have also helped to extend the usefulness of nTOF measurements of ion temperature and downscattered neutron ratio (neutron yield 10-12 MeV divided by yield 13-15 MeV) with other nTOF detectors.

  18. 1987 calibration of the TFTR neutron spectrometers

    International Nuclear Information System (INIS)

    Barnes, C.W.; Strachan, J.D.; Princeton Univ., NJ

    1989-12-01

    The 3 He neutron spectrometer used for measuring ion temperatures and the NE213 proton recoil spectrometer used for triton burnup measurements were absolutely calibrated with DT and DD neutron generators placed inside the TFTR vacuum vessel. The details of the detector response and calibration are presented. Comparisons are made to the neutron source strengths measured from other calibrated systems. 23 refs., 19 figs., 6 tabs

  19. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    International Nuclear Information System (INIS)

    Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

    2010-01-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  20. 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.

  1. Neutron metrology in LAMPF, USA

    International Nuclear Information System (INIS)

    Ketema, D.J.; Nolthenius, H.J.

    1990-08-01

    The characterization of appropriate materials for fusion reactors requires a high intensity neutron source which simulates the neutron spectrum and the radiation conditions at the positions of interest in a fusion reactor (first wall). A neutron spectrum of interest is found in the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF). Various ceramic materials and some polycrystalline graphites were irradiated in this facility during two intervals of time in 1986 and 1987. The specimens were accompanied by several sets with activation detectors. This report presents the saturation activities per atom obtained from these sets. (author). 3 refs.; 8 figs.; 10 tabs

  2. The feasibility study of Dragon Ⅰ using for temperature measurement of resonance neutron

    International Nuclear Information System (INIS)

    Xiang Yanjun; Ma Jingfang; Ai Jie; Fan Ruifeng

    2010-01-01

    The temperature measurement using neutron resonance spectrum can be used for temperature measurement of shock wave, but the high intensity pulsed neutron source is needed. This paper calculates the neutron transmission spectrum through resonance sample (contained 182 W), which produced by the current electron beam of Dragon Ⅰ impacting uranium target. The 4.155 eV and 21.06 eV resonance drop of 182 W can be seen from the transmission spectrum. Then, according to the experiment condition of Los Alamos, the neutron resonance spectrum of Dragon Ⅰ have been computed. Dragon Ⅰ can be used for temperature measurement using neutron spectrum, comparing this simulated result and the experiment result of Los Alamos. (authors)

  3. Absolute calibration system of neutron sources by the manganese sulphate bath

    International Nuclear Information System (INIS)

    Fonseca, E.S. da; Sachett, I.A.

    1990-01-01

    The calibration system consists of deep the neutron source, protected by plastic container, at the center of spherical polietilene tank, in a concentrated solution of manganese sulphate. The neutrons emitted by the source are moderated and when reach the termal energy are catched by manganese atoms activating the solution. After the saturation activity has been reached the source is removed and one scintilation detector (NaI(Tl) 3' x 3') is put in the same place to follow the decay activity. The gama couting rate (845 KeV 54 Mn photopeak), after the corrections is used to estimate the saturation activity, and calculate the neutron source emission rate. These calculations are executed by one computer program. The uncertainties in the final value of emission rate are about 2.5 - 3.0 % to AmBe sources in the 1.11 x 10 10 Bq (0,3 Ci) - 3.7 x 10 11 Bq (10 Ci) range. (author) [pt

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

    CERN Document Server

    Matsumoto, T

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. Neutron fluence spectrometry using disk activation

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  7. Limiting rotational period of neutron stars

    Science.gov (United States)

    Glendenning, Norman K.

    1992-11-01

    We seek an absolute limit on the rotational period for a neutron star as a function of its mass, based on the minimal constraints imposed by Einstein's theory of relativity, Le Chatelier's principle, causality, and a low-density equation of state, uncertainties in which can be evaluated as to their effect on the result. This establishes a limiting curve in the mass-period plane below which no pulsar that is a neutron star can lie. For example, the minimum possible Kepler period, which is an absolute limit on rotation below which mass shedding would occur, is 0.33 ms for a M=1.442Msolar neutron star (the mass of PSR1913+16). A still lower curve, based only on the structure of Einstein's equations, limits any star whatsoever to lie in the plane above it. Hypothetical stars such as strange stars, if the matter of which they are made is self-bound in bulk at a sufficiently large equilibrium energy density, can lie in the region above the general-relativistic forbidden region, and in the region forbidden to neutron stars.

  8. Limiting rotational period of neutron stars

    International Nuclear Information System (INIS)

    Glendenning, N.K.

    1992-01-01

    We seek an absolute limit on the rotational period for a neutron star as a function of its mass, based on the minimal constraints imposed by Einstein's theory of relativity, Le Chatelier's principle, causality, and a low-density equation of state, uncertainties in which can be evaluated as to their effect on the result. This establishes a limiting curve in the mass-period plane below which no pulsar that is a neutron star can lie. For example, the minimum possible Kepler period, which is an absolute limit on rotation below which mass shedding would occur, is 0.33 ms for a M=1.442M circle-dot neutron star (the mass of PSR1913+16). A still lower curve, based only on the structure of Einstein's equations, limits any star whatsoever to lie in the plane above it. Hypothetical stars such as strange stars, if the matter of which they are made is self-bound in bulk at a sufficiently large equilibrium energy density, can lie in the region above the general-relativistic forbidden region, and in the region forbidden to neutron stars

  9. Determination of energy distribution for photon and neutron microdosimetry

    International Nuclear Information System (INIS)

    Todo, A.S.

    1989-01-01

    This work was undertaken to provide basic physical data for use in both microdosimetry and dosimetry of high energy photons and also in the neutron radiation field. It is described the formalism to determine the initial electron energy spectra in water irradiated by photons with energies up to 1 GeV. Calculations were performed with a Monte Carlo computer code, PHOEL-3, which is also described. The code treats explicitly the production of electron-positron pairs, Compton scattering, photoelectric absorption, and the emission of Auger electrons following the occurrence of K-shell vacancies in oxygen. The tables give directly the information needed to specify the absolute single-collision kerma in water, which approximates tissue, at each photon energy. Results for continuous photon energy spectra can be obtained by using linear interpolation with the tables. The conditions under which first-collision kerma approximate absorbed dose are discussed. A formula is given for estimating bremsstrahlung energy loss, one of the principal differences between kerma and absorbed dose in practical case. A study has been carried out, on the use of cylindrical, energy-proportional pulse-height detector for determining microdosimetric quantities, as neutron fractional dose spectra, D (L), in function of linear energy transfer, TLE. In the present study the Hurst detector was used and this device satisfies the requirement of the Bragg-Gray principle. It is developed a Monte Carlo Method to obtain the D(L) spectrum from a measured pulse-height spectrum H(h), and the knowledge of the distribution of recoil-particle track lenght, P(T) in the sensitive volume of the detector. These developed programs to find P(T) and D(L) are presented. The distribution of D(L) in LET were obtained using a known distribution of P(T) and the measured H(h) spectrum from sup(252)Cf neutron source. All the results are discussed and the conclusions are presented. (author)

  10. Neutron absorbed dose in a pacemaker CMOS

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  11. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-15

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

  12. Comparison of Thermal Neutron Flux Measured by Uranium 235 Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

    International Nuclear Information System (INIS)

    Fourmentel, D.; Filliatre, P.; Barbot, L.; Villard, J.-F.; Lyoussi, A.; Geslot, B.; Malo, J.-Y.; Carcreff, H.; Reynard-Carette, C.

    2013-06-01

    Thermal neutron flux is one of the most important nuclear parameter to be measured on-line in Material Testing Reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with uranium 235 coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay - France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor. (authors)

  13. Translation of selected reports on neutron spectrum unfolding

    International Nuclear Information System (INIS)

    Berzonis, M.; Bondars, Kh.Ya.; Taimina, D.

    1982-05-01

    The paper provides the information needed by users of the SAIPS information system on the neutron cross-section libraries accessible and on the principles upon which they are based. Neutron cross-section integrals in fission and fusion spectra are given. (author)

  14. Determination of the Spectral Index in the Fission Spectrum Energy Regime

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Amy Sarah [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-16

    Neutron reaction cross sections play a vital role in tracking the production and destruction of isotopes exposed to neutron fluence. They are central to the process of reconciling the initial and final atom inventories. Measurements of irradiated samples by radiochemical methods in tangent with an algorithm are used to evaluate the fluence a sample is exposed to over the course of the irradiation. This algorithm is the Isotope Production Code (IPC) created and used by the radiochemistry data assessment team at Los Alamos National Laboratory (LANL). An integral result is calculated by varying the total neutron fluence seen by a sample. A sample, irradiated in a critical assembly, will be exposed to a unique neutron flux defined by the neutron source and distance of the sample from the source. Neutron cross sections utilized are a function of the hardness of the neutron spectrum at the location of irradiation. A spectral index is used an indicator of the hardness of the neutron spectrum. Cross sections fit forms applied in IPC are collapsed from a LANL 30-group energy structure. Several decades of research and development have been performed to formalize the current IPC cross section library. Basis of the current fission spectrum neutron reaction cross section library is rooted in critical assembly experiments performed from the 1950’s through the early 1970’s at LANL. The focus of this report is development of the spectral index used an indicator of the hardness of the neutron spectrum in the fission spectrum energy regime.

  15. ATHENA2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum

    International Nuclear Information System (INIS)

    1999-01-01

    neutronics and CFD equations are solved by successive-over relaxation (SOR) iteration. Chebychev extrapolation is available for the neutronics equations, although this option has not been thoroughly tested. The CFD equations use the semi-implicit method for pressure-linked equations (SIMPLE) iteration to achieve self-consistent solutions for mass, momentum and energy. Asymptotic acceleration of the pressure-correction equation (part of the SIMPLE iteration) is available. The one-dimensional thermal hydraulic fuel pin models employ fast tri-diagonal matrix inversion for a single time step. 3 - Restrictions on the complexity of the problem: ATHENA 2 D assumes moderator boiling occurs before fuel remelt, as there is no mechanism to handle fuel remelt and relocation. However, even the most severe transients simulated during the course of the PhD work showed that this is reasonable. In a thermal neutron spectrum, time constants are longer than those in a fast spectrum. Given typical fuel piece dimensions, moderator boiling always occurred before the peak fuel temperature reached the melting point. Another limitation is the single (liquid) phase CFD model. Boiling is treated, but only insofar as to calculate local void fractions that feed back to the neutronics equations through local cross section interpolation based on reduced moderator density. The reactivity effects of voids being explicitly transported away from their point of origin is not treated. However, these effects are believed to be small as the introduction of voids tends to be a primary shutdown mechanism for these severe transients. Improved two-phase modeling would only affect the details of the shutdown phase of the transient, not the total energy release. Other Limitations: Reactivity feedback effects arising from any potential fluidized bed motion of fuel particles in the debris bed is not treated. No fuel motion is modeled. Blackbody (radiative) heat transfer is not modeled. Radiolytic gas bubble formation and

  16. Absolute measurement of the cross sections of neutron radiative capture for 23Na, Cr, 55Mn, Fe, Ni, 103Rh, Ta, 197Au and 238U in the 10-600keV energy range

    International Nuclear Information System (INIS)

    Le Rigoleur, Claude; Arnaud, Andre; Taste, Jean.

    1976-10-01

    The total energy weighting technique has been applied to measuring absolute neutron capture cross sections for 23 Na, Cr, 55 Mn, Fe, Ni, 103 Rh, Ta, 197 Au, 238 U in the 10-600keV energy range. A non hydrogeneous liquid scintillator was used to detect the gamma from the cascade. The neutron flux was measured with a 10 B INa(Tl) detector or a 6 Li glass scintillator of well known efficiency. The fast time-of-flight technique was used with on line digital computer data processing [fr

  17. Recent improvements in the calculation of prompt fission neutron spectra: Preliminary results

    International Nuclear Information System (INIS)

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

    1989-01-01

    We consider three topics in the refinement and improvement of our original calculations of prompt fission neutron spectra. These are an improved calculation of the prompt fission neutron spectrum N(E) from the spontaneous fission of 252 Cf, a complete calculation of the prompt fission neutron spectrum matrix N(E,E n ) from the neutron-induced fission of 235 U, at incident neutron energies ranging from 0 to 15 MeV, and an assessment of the scission neutron component of the prompt fission neutron spectrum. Preliminary results will be presented and compared with experimental measurements and an evaluation. A suggestion is made for new integral cross section measurements. (author). 45 refs, 12 figs, 1 tab

  18. Research on neutron energy spectrum of the beryllium, iron and polyethylene shells assemblies injected by D-T neutron

    International Nuclear Information System (INIS)

    An, Li; Guo, Haiping; Wang, Xinhua

    2009-04-01

    To test a simulation code, the multi-shell assemblies were established, which were made of beryllium stainless steel and polyethylene from the interior to the outer. The symmetry axes are all in the line of the D + beam. The neutron energy spectra above 1 MeV were obtained in medium by the detector of stilbene crystal of φ18 min x 20 mm. The distance between source and the spherical surface was 30 cm and 50 cm. The measurement channels are in the angle 0 degree and 120 degree relative to D + beam direction. The measurement positions are 0 cm, 9.7 cm, 12.8 cm and 17.3 cm away from the center of the assembly in both directions. The spectrum in different positions of the multi-shell assemblies in medium were compared and analyzed. (authors)

  19. Spectrum of the multigroup neutron transport operator for bounded spatial domains

    International Nuclear Information System (INIS)

    Larsen, E.W.

    1979-01-01

    The spectrum of the multigroup neutron transport operator A is studied for bounded spatial regions D which consist of a finite number of material subregions. Our main results provide simple conditions on the material cross sections which guarantee that (1) A possesses eigenvalues in the finite plane; (2) A possesses a ''leading'' eigenvalue lambda 0 which is real, not less than the real part of any other eigenvalue, and to which there corresponds at least one nonnegative eigenfunction psi/sub lambda/0; and (3) A possesses a ''dominant'' eigenvalue lambda 0 which is real, simple, greater than the real part of any other eigenvalue, and whose eigenfunction psi/sub lambda/0 satisfies psi/sub lambda/0> or =0 and ∫psi/sub lambda/0d 2 Ω>0. We give examples to illustrate the results and to show that a leading eigenvalue need not be simple, nor its eigenfunction(s) positive

  20. Compilation of neutron flux density spectra and reaction rates in different neutron fields

    International Nuclear Information System (INIS)

    Ertek, C.

    1979-07-01

    Upon the recommendation of International Working Group of Reactor Radiation Measurements (IWGRRM), the compilation of neutron flux density spectra and the reaction rates obtained by activation and fission foils in different neutron fields is presented. The neutron fields considered are as follows: 1/E; iron block; LWR core and pressure vessel; LMFBR core and blanket; CTR first wall and blanket; fission spectrum

  1. Neutron fluence spectrometry using disk activation

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-15

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

  2. Plasma-erosion-enhanced neutron emission in fiber-generated dense Z-pinches

    International Nuclear Information System (INIS)

    Mosher, D.; Colombant, D.

    1990-01-01

    Experiments in which dense z-pinches are created from high-current discharges through frozen deuterium fibers have reported neutron yields far in excess of those expected from thermal processes. A simple analysis based on pinch collapse due to the sausage instability has successfully predicted the relative variation of neutron yield with discharge current, but model assumptions precluded prediction of absolute values of the yield. A pinch-collapse model derived from a 2-dimensional, nonlinear treatment of the sausage instability, combined with space-charged-limited (SCL) ion orbital dynamic for the vacuum region above the pinches and between the expanding flares, leads to neutron yields four or more orders-of-magnitude below experimental values. Here, the same pinch-collapse model is used in conjunction with a low-density plasma background above the collapsing pinches. Ions are accelerated across the space-charge sheath separating the background plasma from the flares, which electron emission from the flares is strongly insulated by the z-discharge magnetic field. The sheath gap increases in time, i.e., the background plasma erodes, at a rate determined by its density and the SCL ion current density which, in turn, depends on the z-discharge dynamics and the associated induced electromagnetic fields. A model incorporating the above processes is used to determine the accelerated ion energy spectrum and associated neutron yield as functions of the discharge, instability, and background parameters

  3. Neutronic analysis of JET external neutron monitor response

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom (United Kingdom); Batistoni, Paola [ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden)

    2016-11-01

    Highlights: • We model JET tokamak containing JET remote handling system. • We investigate effect of remote handling system on external neutron monitor response. • Remote handling system correction factors are calculated. • Integral correction factors are relatively small, i.e up to 8%. - Abstract: The power output of fusion devices is measured in terms of the neutron yield which relates directly to the fusion yield. JET made a transition from Carbon wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) during 2010–11. Absolutely calibrated measurement of the neutron yield by JET neutron monitors was ensured by direct measurements using a calibrated {sup 252}Cf neutron source (NS) deployed by the in-vessel remote handling system (RHS) inside the JET vacuum vessel. Neutronic calculations were required in order to understand the neutron transport from the source in the vacuum vessel to the fission chamber detectors mounted outside the vessel on the transformer limbs of the tokamak. We developed a simplified computational model of JET and the JET RHS in Monte Carlo neutron transport code MCNP and analyzed the paths and structures through which neutrons reach the detectors and the effect of the JET RHS on the neutron monitor response. In addition we performed several sensitivity studies of the effect of substantial massive structures blocking the ports on the external neutron monitor response. As the simplified model provided a qualitative picture of the process only, some calculations were repeated using a more detailed full 3D model of the JET tokamak.

  4. Experiment and analysis of neutron spectra in a concrete assembly bombarded by 14 MeV neutrons

    International Nuclear Information System (INIS)

    Oishi, Koji; Tomioka, Kazuyuki; Ikeda, Yujiro; Nakamura, Tomoo.

    1988-01-01

    Neutron spectrum in concrete bombarded by 14 MeV neutrons was measured using a miniature NE213 spectrometer and multi-foil activation method. A good agreement between those two experimental methods was obtained within experimental errors. The measured spectrum was compared with calculated ones using two-dimensional transport code DOT3.5 with 125 group structure cross section libraries based on ENDF/B-IV, JENDL-2, and JENDL-3T (the testing version of JENDL-3.) In the D-T neutron peak region, measured and calculated neutron spectra agreed well with each other for those libraries. However, disagreements of about -10 % to +50 % and -30 % to +40 % were obtained in the MeV region and still lower neutron energy range, respectively. As a result, it was concluded that those discrepancies were caused by the overestimation of secondary neutrons emitted by inelastic scattering from O, Si, and/or Ca which were the main components of concrete. (author)

  5. Absolutely continuous spectrum and spectral transition for some ...

    Indian Academy of Sciences (India)

    Proc. Indian Acad. Sci. (Math. Sci.) Vol. 122, No. 2, May 2012, pp. 243–255. c Indian ... we also know the existence of dense pure point spectrum for some disorder thus exhibit- ing spectral .... When β varies from 1 to 0, the growth behaviour of N(R) changes from ...... Student Texts 37 (Cambridge University Press) (1997).

  6. Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

    International Nuclear Information System (INIS)

    Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

    1989-01-01

    Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

  7. Characterization of a scintillating lithium glass ultra-cold neutron detector

    Energy Technology Data Exchange (ETDEWEB)

    Jamieson, B.; Rebenitsch, L.A.; Hansen-Romu, S.; Mammei, R.; Martin, J.W. [University of Winnipeg, Department of Physics, Winnipeg (Canada); Lauss, B. [Paul Scherrer Institute, Laboratory for Particle Physics, Villigen (Switzerland); Lindner, T. [TRIUMF, Vancouver (Canada); University of Winnipeg, Department of Physics, Winnipeg (Canada); Pierre, E. [TRIUMF, Vancouver (Canada); Osaka University, Research Centre for Nuclear Physics, Osaka (Japan)

    2017-01-15

    A {sup 6}Li-glass-based scintillation detector developed for the TRIUMF neutron electric dipole moment experiment was characterized using the ultra-cold neutron source at the Paul Scherrer Institute (PSI). The data acquisition system for this detector was demonstrated to perform well at rejecting backgrounds. An estimate of the absolute efficiency of background rejection of 99.7±0.1% is made. For variable ultra-cold neutron rate (varying from < 1 kHz to approx. 100 kHz per channel) and background rate seen at the Paul Scherrer Institute, we estimate that the absolute detector efficiency is 89.7{sup +1.3}{sub -1.9}%. Finally a comparison with a commercial Cascade detector was performed for a specific setup at the West-2 beamline of the ultra-cold neutron source at PSI. (orig.)

  8. Measurement of the cosmic-induced neutron yield at the Modane underground laboratory

    International Nuclear Information System (INIS)

    Kluck, Holger Martin

    2013-01-01

    uncertainties. The experimental value of (3.2_-_0_._3"+"0"."5) . 10"-"1 neutrons per day is reproduced by MC to within 15 %. Also the measured absolute multiplicity spectrum is well reproduced by our model. The neutron production yield in lead at LSM is, for the first time, derived to be left angle Y right angle =(2.71_-_0_._1_0"+"0"."9"8) . 10"-"3 cm"2 g"-"1 for muon energies left angle E_μ right angle =267 GeV. This work demonstrates, that Geant4 can reliably model the production and detection of muon-induced neutrons once all relevant production processes and a detailed description of the detector response and geometry are implemented in the model. Thus, one of the most prominent background sources for Dark Matter search can be accurately modelled and eventually suppressed.

  9. Different spectra with the same neutron source

    International Nuclear Information System (INIS)

    Vega C, H. R.; Ortiz R, J. M.; Hernandez D, V. M.; Martinez B, M. R.; Hernandez A, B.; Ortiz H, A. A.; Mercado, G. A.

    2010-01-01

    Using as source term the spectrum of a 239 Pu-Be source several neutron spectra have been calculated using Monte Carlo methods. The source term was located in the centre of spherical moderators made of light water, heavy water and polyethylene of different diameters. Also a 239 Pu-Be source was used to measure its neutron spectrum, bare and moderated by water. The neutron spectra were measured at 100 cm with a Bonner spheres spectrometer. Monte Carlo calculations were used to calculate the neutron spectra of bare and water-moderated spectra that were compared with those measured with the spectrometer. Resulting spectra are similar to those found in power plants with PWR, BWR and Candu nuclear reactors. Beside the spectra the dosimetric features were determined. Using moderators and a single neutron source can be produced neutron spectra alike those found in workplaces, this neutron fields can be utilized to calibrate neutron dosimeters and area monitors. (Author)

  10. Remarks on the comparison of cross section libraries for neutron metrology

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.; Appelman, K.H.

    1977-01-01

    Cross section libraries in a 620 group structure were available from different origin: CCC-112B, DETAN-74 and ENDF/B-IV. For a few well known neutron spectra (CFRMF spectrum, ΣΣ spectrum, fission neutron spectrum, HFR neutron spectrum) a comparison was made of the available experimental reaction rates in foil detectors and the reaction rates as calculated with the different cross section libraries. This investigation is dealing with the consistency of cross section data within a library, and the consistency of activity data in actual reaction rate determinations. Some preliminary conclusions are given

  11. Source Correlated Prompt Neutron Activation Analysis for Material Identification and Localization

    Science.gov (United States)

    Canion, Bonnie; McConchie, Seth; Landsberger, Sheldon

    2017-07-01

    This paper investigates the energy spectrum of photon signatures from an associated particle imaging deuterium tritium (API-DT) neutron generator interrogating shielded uranium. The goal is to investigate if signatures within the energy spectrum could be used to indirectly characterize shielded uranium when the neutron signature is attenuated. By utilizing the correlated neutron cone associated with each pixel of the API-DT neutron generator, certain materials can be identified and located via source correlated spectrometry of prompt neutron activation gamma rays. An investigation is done to determine if fission neutrons induce a significant enough signature within the prompt neutron-induced gamma-ray energy spectrum in shielding material to be useful for indirect nuclear material characterization. The signature deriving from the induced fission neutrons interacting with the shielding material was slightly elevated in polyethylene-shielding depleted uranium (DU), but was more evident in some characteristic peaks from the aluminum shielding surrounding DU.

  12. Neutronic moderator design for the Spallation Neutron Source (SNS)

    International Nuclear Information System (INIS)

    Charlton, L.A.; Barnes, J.M.; Johnson, J.O.; Gabriel, T.A.

    1998-01-01

    Neutronics analyses are now in progress to support the initial selection of moderator design parameters for the Spallation Neutron Source (SNS). The results of the initial optimization studies involving moderator poison plate location, moderator position, and premoderator performance for the target system are presented in this paper. Also presented is an initial study of the use of a composite moderator to produce a liquid methane like spectrum

  13. Use of sapphire as a neutron damage monitor for pressure vessel steels

    International Nuclear Information System (INIS)

    Pells, G.P.; Fudge, A.J.; Murphy, M.J.; Watt, S.

    1989-01-01

    Single crystal α-Al 2 O 3 (sapphire) has been neutron irradiated over a range of dose, dose rate and neutron energy spectra at temperatures from 60 to 310 0 C. Values of optical absorption at 400 nm, the peak of the aluminum vacancy absorption band, were plotted against damage dose expressed in terms of dpa of Al in sapphire obtained from measurements of induced radio-activity in activation foils irradiated with the sapphires and from calculation of the neutron energy spectrum at the irradiation position. The neutron energy spectrum was calculated using modern neutron transport computer codes and adjusted in the light of measurements obtained from multiple foil activation experiments. A simple response curve was obtained for all sapphires irradiated at temperatures between 220 to 310 0 C and for sapphires irradiated below 200 0 C which had been annealed at 290 0 C irrespective of dose rate or neutron beam energy spectrum. The single response curve for irradiations performed in a variety of neutron energy spectra validate the neutron energy spectrum computational procedures

  14. The magnetic recoil spectrometer (MRSt) for time-resolved measurements of the neutron spectrum at the National Ignition Facility (NIF)

    Energy Technology Data Exchange (ETDEWEB)

    Frenje, J. A., E-mail: jfrenje@psfc.mit.edu; Wink, C. W.; Gatu Johnson, M.; Li, C. K.; Séguin, F. H.; Petrasso, R. D. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Hilsabeck, T. J.; Kilkenny, J. D. [General Atomics, San Diego, California 92186 (United States); Bell, P.; Bionta, R.; Cerjan, C. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2016-11-15

    The next-generation magnetic recoil spectrometer for time-resolved measurements of the neutron spectrum has been conceptually designed for the National Ignition Facility. This spectrometer, called MRSt, represents a paradigm shift in our thinking about neutron spectrometry for inertial confinement fusion applications, as it will provide simultaneously information about the burn history and time evolution of areal density (ρR), apparent ion temperature (T{sub i}), yield (Y{sub n}), and macroscopic flows during burn. From this type of data, an assessment of the evolution of the fuel assembly, hotspot, and alpha heating can be made. According to simulations, the MRSt will provide accurate data with a time resolution of ∼20 ps and energy resolution of ∼100 keV for total neutron yields above ∼10{sup 16}. At lower yields, the diagnostic will be operated at a higher-efficiency, lower-energy-resolution mode to provide a time resolution of ∼20 ps.

  15. The magnetic recoil spectrometer (MRSt) for time-resolved measurements of the neutron spectrum at the National Ignition Facility (NIF).

    Science.gov (United States)

    Frenje, J A; Hilsabeck, T J; Wink, C W; Bell, P; Bionta, R; Cerjan, C; Gatu Johnson, M; Kilkenny, J D; Li, C K; Séguin, F H; Petrasso, R D

    2016-11-01

    The next-generation magnetic recoil spectrometer for time-resolved measurements of the neutron spectrum has been conceptually designed for the National Ignition Facility. This spectrometer, called MRSt, represents a paradigm shift in our thinking about neutron spectrometry for inertial confinement fusion applications, as it will provide simultaneously information about the burn history and time evolution of areal density (ρR), apparent ion temperature (T i ), yield (Y n ), and macroscopic flows during burn. From this type of data, an assessment of the evolution of the fuel assembly, hotspot, and alpha heating can be made. According to simulations, the MRSt will provide accurate data with a time resolution of ∼20 ps and energy resolution of ∼100 keV for total neutron yields above ∼10 16 . At lower yields, the diagnostic will be operated at a higher-efficiency, lower-energy-resolution mode to provide a time resolution of ∼20 ps.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  18. 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.)

  19. Methods of neutron spectrometry

    International Nuclear Information System (INIS)

    Doerschel, B.

    1981-01-01

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

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

    Science.gov (United States)

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

    2016-10-01

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

  1. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of {sup 252}Cf, {sup 241}AmBe and {sup 239}PuBe neutron sources measured with a Bonner spheres system.

  2. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252 Cf, 241 AmBe and 239 PuBe neutron sources measured with a Bonner spheres system

  3. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called "Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres", (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the "Robust design of artificial neural networks methodology" and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252Cf, 241AmBe and 239PuBe neutron sources measured with a Bonner spheres system.

  4. Genetic algorithms - A new technique for solving the neutron spectrum unfolding problem

    International Nuclear Information System (INIS)

    Freeman, David W.; Edwards, D. Ray; Bolon, Albert E.

    1999-01-01

    A new technique utilizing genetic algorithms has been applied to the Bonner sphere neutron spectrum unfolding problem. Genetic algorithms are part of a relatively new field of 'evolutionary' solution techniques that mimic living systems with computer-simulated 'chromosome' solutions. Solutions mate and mutate to create better solutions. Several benchmark problems, considered representative of radiation protection environments, have been evaluated using the newly developed UMRGA code which implements the genetic algorithm unfolding technique. The results are compared with results from other well-established unfolding codes. The genetic algorithm technique works remarkably well and produces solutions with relatively high spectral qualities. UMRGA appears to be a superior technique in the absence of a priori data - it does not rely on 'lucky' guesses of input spectra. Calculated personnel doses associated with the unfolded spectra match benchmark values within a few percent

  5. Use of the associated particle technique in the fast neutron spectroscopy

    International Nuclear Information System (INIS)

    Aquirre O, G.A.

    1978-01-01

    Selecting a neutrons monoenergetic source it was found that the nuclear reaction D(d,n) 3 He can be used to measure nuclear sections and differentials in elastic nuclear reactions through the associated particle technique; the neutron beam energy is directly determined in time of flight spectrum of the neutron. The flux is determined by the number of 3 He ions observed in the charged particle spectrum. The neutron flux can be increased increasing the solid angle of the neutrons beam in two magnitude orders according to the results of neutrons beam profile measures. (author)

  6. Superresolution of a compact neutron spectrometer at energies relevant for fusion diagnostics

    International Nuclear Information System (INIS)

    Reginatto, M.; Zimbal, A.

    2011-01-01

    The ability to achieve resolution that is better than the instrument resolution (i.e., superresolution) is well known in optics, where it has been extensively studied. Unfortunately, there are only a handful of theoretical studies concerning superresolution of particle spectrometers, even though experimentalists are familiar with the enhancement of resolution that is achievable when appropriate methods of data analysis are used, such as maximum entropy and Bayesian methods. Knowledge of the superresolution factor is in many cases important. For example, in applications of neutron spectrometry to fusion diagnostics, the temperature of a burning plasma is an important physical parameter which may be inferred from the width of the peak of the neutron energy spectrum, and the ability to determine this width depends on the superresolution factor. Kosarev has derived an absolute limit for resolution enhancement using arguments based on a well known theorem of Shannon. Most calculations of superresolution factors in the literature, however, are based on the assumption of Gaussian, translationally invariant response functions and therefore not directly applicable to neutron spectrometers which typically have response functions not satisfying these requirements. In this work, we develop a procedure that allows us to overcome these difficulties and we derive estimates of superresolution for liquid scintillator spectrometers of a type commonly used for neutron measurements. Theoretical superresolution factors are compared to experimental results.

  7. The Prompt Fission Neutron Spectrum: From Experiment to the Evaluated Data and its Impact on Critical Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Laboratory

    2015-06-10

    After a brief introduction concerning nuclear data, prompt fission neutron spectrum (PFNS) evaluations and the limited PFNS covariance data in the ENDF/B-VII library, and the important fact that cross section uncertainties ~ PFNS uncertainties, the author presents background information on the PFNS (experimental data, theoretical models, data evaluation, uncertainty quantification) and discusses the impact on certain well-known critical assemblies with regard to integral quantities, sensitivity analysis, and uncertainty propagation. He sketches recent and ongoing research and concludes with some final thoughts.

  8. A method for prediction of prompt fission neutron spectra

    International Nuclear Information System (INIS)

    Grashin, A.F.; Lepeshkin, M.V.

    1988-01-01

    Three-parameter formula for the prompt-fission-neutron integral spectrum is derived from a thermodynamical model. Two parameters, scission-neutron weight p = 11 % and anisotropy factor for accelerated fragments b = 10 %, are determined from experimental data, the same values being assumed for any type of fission. The thermodynamical theory provides the value of the third parameter, temperature τ, thus prognozing neutron spectrum and average energy with an error about 1 %. (author)

  9. Intense neutron source facility for the fusion energy program

    International Nuclear Information System (INIS)

    Armstrong, D.D.; Emigh, C.R.; Meier, K.L.; Meyer, E.A.; Schneider, J.D.

    1975-01-01

    The Intense Neutron Source Facility, INS, has been proposed to provide a neutronic environment similar to that anticipated in a fully operational fusion-power reactor. The neutron generator will produce an intense flux of 14-MeV neutrons greater than 10 14 neutrons per cm 2 /sec from the collision of two intersecting beams, one of 1.1 A of 270 keV tritium ions and the other of a supersonic jet of deuterium gas. Using either the pure 14-MeV primary neutron spectrum or by tailoring the spectrum with appropriate moderators, crucial radiation-damage effects which are likely to occur in fusion reactors can be thoroughly explored and better understood

  10. Fast neutron dosimetry in research reactors

    International Nuclear Information System (INIS)

    Eckert, R.

    1960-01-01

    This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [fr

  11. 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

  12. Neutron response study

    International Nuclear Information System (INIS)

    Endres, G.W.R.; Fix, J.J.; Thorson, M.R.; Nichols, L.L.

    1981-01-01

    Neutron response of the albedo type dosimeter is strongly dependent on the energy of the incident neutrons as well as the moderating material on the backside of the dosimeter. This study characterizes the response of the Hanford dosimeter for a variety of neutron energies for both a water and Rando phantom (a simulated human body consisting of an actual human skeleton with plastic for body muscles and certain organs). The Hanford dosimeter response to neutrons of different energies is typical of albedo type dosimeters. An approximate two orders of magnitude difference in response is observed between neutron energies of 100 keV and 10 MeV. Methods were described to compensate for the difference in dosimeter response between a laboratory neutron spectrum and the different spectra encountered at various facilities in the field. Generally, substantial field support is necessary for accurate neutron dosimetry

  13. Status of ITER neutron diagnostic development

    International Nuclear Information System (INIS)

    Sasao, M.; Krasilnikov, A.V.; Kaschuck, Yu.A.; Nishitani, T.; Batistoni, P.; Zaveryaev, V.S.; Popovichev, S.; Jarvis, O.N.; Iguchi, T.; Kaellne, J.; Fiore, C.L.; Roquemore, A.L.; Heidbrink, W.W.; Fisher, R.; Gorini, G.; Donne, A.J.H.; Costley, A.E.; Walker, C.I.

    2005-01-01

    Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be well measured by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include: radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors, neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The neutron flux monitors need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented. (author)

  14. Neutron flux monitoring device

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro.

    1995-01-01

    In a neutron flux monitoring device, there are disposed a neutron flux measuring means for outputting signals in accordance with the intensity of neutron fluxes, a calculation means for calculating a self power density spectrum at a frequency band suitable to an object to be measured based on the output of the neutron flux measuring means, an alarm set value generation means for outputting an alarm set value as a comparative reference, and an alarm judging means for comparing the alarm set value with the outputted value of the calculation means to judge requirement of generating an alarm and generate an alarm in accordance with the result of the judgement. Namely, the time-series of neutron flux signals is put to fourier transformation for a predetermined period of time by the calculation means, and from each of square sums for real number component and imaginary number component for each of the frequencies, a self power density spectrum in the frequency band suitable to the object to be measured is calculated. Then, when the set reference value is exceeded, an alarm is generated. This can reliably prevent generation of erroneous alarm due to neutron flux noises and can accurately generate an alarm at an appropriate time. (N.H.)

  15. Source characterization of Purnima Neutron Generator (PNG)

    International Nuclear Information System (INIS)

    Bishnoi, Saroj; Patel, T.; Paul, Ram K.; Sarkar, P.S.; Adhikari, P.S.; Sinha, Amar

    2011-01-01

    The use of 14.1 MeV neutron generators for the applications such as elemental analysis, Accelerated Driven System (ADS) study, fast neutron radiography requires the characterization of neutron source i.e neutron yield (emission rate in n/sec), neutron dose, beam spot size and energy spectrum. In this paper, a series of experiments carried out to characterize this neutron source. The neutron source has been quantified with neutron emission rate, neutron dose at various source strength and beam spot size at target position

  16. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  17. Measurement of fast neutron spectrum using CR-39 detectors and a new image analysis program (autoTRAKn)

    International Nuclear Information System (INIS)

    Paul, Sabyasachi; Tripathy, S.P.; Sahoo, G.S.; Bandyopadhyay, T.; Sarkar, P.K.

    2013-01-01

    An attempt is made to estimate the neutron spectrum using the CR-39 (Solid state nuclear track) detector and a new image analyzing program. For this purpose the earlier developed program (autoTRAK) is modified by introducing the required features such as angular correction for the recoil particles, fluence-to-dose conversion coefficient, detection sensitivity of CR-39 detectors, etc. to make it applicable for neutron spectrometry and dosimetry. This upgraded program (autoTRAK n ) is tested with a mono-energetic source (D–T) and two other standard neutron sources, viz. 241 Am–Be and 252 Cf. The program is validated by reproducing these standard spectra, and comparing with the spectra reported by other investigators using different measuring techniques. The ratios of dose equivalent (H ⁎ (10)) to fluence (Φ) are also estimated from the spectra and are compared with the reference values for these neutron sources. An additional feature of this program is explored for counting high density overlapping tracks more precisely and effectively compared to other commonly used image analyzing softwares. This method is found to be simple and promising, which can always be used as a supplementary measuring technique. The details of the modified program, reproduction and comparison of the neutron spectra, reproducibility of the methodology and example of overlapping track counting are presented and discussed. -- Highlights: •A novel image analysis technique (autoTRAK n ) is developed to evaluate CR-39 detectors used for neutron spectrometry and dosimetry. •The methodology is tested to reproduce three standard neutron spectra, (a) D–T, (b) 241 Am–Be, and (c) 252 Cf. •A good matching is observed between dosimetric values obtained by the program and the available reference values. •The program autoTRAK n is also observed to be efficient to distinguish high density overlapping tracks without any segregation procedure. •The methodology seems to be simple, which

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  19. ATR neutron spectral characterization

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-11-01

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

  20. Measurement of thermal neutron spectrum by chopper at the RA reactor in the 'Boris Kidric' Institute; Merenje termickog neutronskog spektra iz reaktora RA u Institutu 'Boris Kidric' pomocu copera

    Energy Technology Data Exchange (ETDEWEB)

    Maglic, R [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    Measurement of neutron spectrum described in this paper is one of the first measurements of neutron energy distribution at the reactor by time-of-flight method performed in our Institute. Measurement was done by applying the mechanical spectrometer (chopper) designed and constructed in 1961. Spectrometer was calibrated at the end of 1962.

  1. Status of ITER neutron diagnostic development

    Science.gov (United States)

    Krasilnikov, A. V.; Sasao, M.; Kaschuck, Yu. A.; Nishitani, T.; Batistoni, P.; Zaveryaev, V. S.; Popovichev, S.; Iguchi, T.; Jarvis, O. N.; Källne, J.; Fiore, C. L.; Roquemore, A. L.; Heidbrink, W. W.; Fisher, R.; Gorini, G.; Prosvirin, D. V.; Tsutskikh, A. Yu.; Donné, A. J. H.; Costley, A. E.; Walker, C. I.

    2005-12-01

    Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be measured well by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors (NFMs), neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The NFMs need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented.

  2. Status of ITER neutron diagnostic development

    International Nuclear Information System (INIS)

    Krasilnikov, A.V.; Sasao, M.; Kaschuck, Yu.A.; Nishitani, T.; Batistoni, P.; Zaveryaev, V.S.; Popovichev, S.; Iguchi, T.; Jarvis, O.N.; Kaellne, J.; Fiore, C.L.; Roquemore, A.L.; Heidbrink, W.W.; Fisher, R.; Gorini, G.; Prosvirin, D.V.; Tsutskikh, A.Yu.; Donne, A.J.H.; Costley, A.E.; Walker, C.I.

    2005-01-01

    Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be measured well by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors (NFMs), neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The NFMs need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented

  3. A coaxial double cylindrical TEPC for the microdosimetry of selected neutron energy bands in mixed fields of fast neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Saion, E.B.; Watt, D.E. (Saint Andrews Univ. (UK). Dept. of Physics); East, B.W. (Scottish Universities Research and Reactor Centre, Glasgow (UK)); Colautti, P. (Istituto Nazionale di Fisica Nucleare, Padua (Italy))

    1990-01-01

    A new low pressure tissue-equivalent proportional counter (TEPC) in a coaxial double cylindrical form has been developed to measure separately the microdose spectrum from any desired energy band of neutrons in the presence of mixed fields of faster neutrons, by selecting the thickness of the common TE dividing wall to be equivalent to the corresponding maximum proton ranges and by appropriate use of coincidence/anti-coincidence pulse arrangements. This thickness ensures charged particle equilibrium for the relevant neutron energy. Event spectra due to recoils generated by faster neutrons which interact with both the counters are removed completely by anti-coincidence techniques, thereby optimising the sensitivity of the inner microdosemeter to the event spectra of interest. The ability of this counter to discriminate in favour of events due to neutrons of energy <850 keV was achieved in microdosimetric measurements from mixed fields of a nuclear reactor. Mean values of lineal energy and quality factor for neutrons of energy <850 keV from a nuclear reactor were determined from the anti-coincidence spectrum. Good discrimination against {gamma} ray induced events is also achieved for the spectrum recorded in the anti-coincidence mode. This is an advantageous feature for other applications and requires further investigation. (author).

  4. A coaxial double cylindrical TEPC for the microdosimetry of selected neutron energy bands in mixed fields of fast neutrons

    International Nuclear Information System (INIS)

    Saion, E.B.; Watt, D.E.; Colautti, P.

    1990-01-01

    A new low pressure tissue-equivalent proportional counter (TEPC) in a coaxial double cylindrical form has been developed to measure separately the microdose spectrum from any desired energy band of neutrons in the presence of mixed fields of faster neutrons, by selecting the thickness of the common TE dividing wall to be equivalent to the corresponding maximum proton ranges and by appropriate use of coincidence/anti-coincidence pulse arrangements. This thickness ensures charged particle equilibrium for the relevant neutron energy. Event spectra due to recoils generated by faster neutrons which interact with both the counters are removed completely by anti-coincidence techniques, thereby optimising the sensitivity of the inner microdosemeter to the event spectra of interest. The ability of this counter to discriminate in favour of events due to neutrons of energy <850 keV was achieved in microdosimetric measurements from mixed fields of a nuclear reactor. Mean values of lineal energy and quality factor for neutrons of energy <850 keV from a nuclear reactor were determined from the anti-coincidence spectrum. Good discrimination against γ ray induced events is also achieved for the spectrum recorded in the anti-coincidence mode. This is an advantageous feature for other applications and requires further investigation. (author)

  5. Neutron and gamma-ray spectra of 239PuBe and 241AmBe

    International Nuclear Information System (INIS)

    Vega-Carrillo, H.R.; Manzanares-Acuna, Eduardo; Becerra-Ferreiro, A.M.; Carrillo-Nunez, Aureliano

    2002-01-01

    Neutron and gamma-ray spectra of 239 PuBe and 241 AmBe were measured and their dosimetric features were calculated. Neutron spectra were measured using a multisphere neutron spectrometer with a 6 LiI(Eu) scintillator. The 239 PuBe neutron spectrum was measured in an open environment, while the 241 AmBe neutron spectrum was measured in a closed environment. Gamma-ray spectra were measured using a NaI(Tl) scintillator using the same experimental conditions for both sources. The effect of measuring conditions for the 241 AmBe neutron spectrum indicates the presence of epithermal and thermal neutrons. The low-resolution neutron spectra obtained with the multisphere spectrometer allows one to calculate the dosimetric features of neutron sources. At 100 cm both sources produce approximately the same count rate as that of the 4.4 MeV gamma-ray per unit of alpha emitter activity

  6. Distinguishing Pu Metal from Pu Oxide and Determining alpha-ratio using Fast Neutron Counting

    Energy Technology Data Exchange (ETDEWEB)

    Verbeke, J. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chapline, G. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nakae, L. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sheets, S. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-01-07

    We describe a new method for determining the ratio of the rate of (α, n) source neutrons to the rate of spontaneous fission neutrons, the so called α-ratio. This method is made possible by fast neutron counting with liquid scintillator detectors, which can determine the shape of the fast neutron spectrum. The method utilizes the spectral difference between fission spectrum neutrons from Pu metal and the spectrum of (α, n) neutrons from PuO2. Our method is a generalization of the Cifarelli-Hage method for determining keff for fissile assemblies, and also simultaneously determines keff along with the α-ratio.

  7. Benchmark experiment on vanadium assembly with D-T neutrons. Leakage neutron spectrum measurement

    Energy Technology Data Exchange (ETDEWEB)

    Kokooo; Murata, I.; Nakano, D.; Takahashi, A. [Osaka Univ., Suita (Japan); Maekawa, F.; Ikeda, Y.

    1998-03-01

    The fusion neutronics benchmark experiments have been done for vanadium and vanadium alloy by using the slab assembly and time-of-flight (TOF) method. The leakage neutron spectra were measured from 50 keV to 15 MeV and comparison were done with MCNP-4A calculations which was made by using evaluated nuclear data of JENDL-3.2, JENDL-Fusion File and FENDL/E-1.0. (author)

  8. The ration/gsub(A)/gsub(V) derived from the proton spectrum in free-neutron decay

    International Nuclear Information System (INIS)

    Stratowa, Ch.; Dobrozemsky, R.; Weinzierl, P.

    1978-08-01

    The electron-neutrino angular correlation coefficient was determined by measuring the shape of the proton recoil spectrum from free-neutron decay. The protons leaving a highly evacuated tangential reactor beam tube were analysed by a spherical condenser spectrometer and counted in an ion-electron converter detector. The design of the apparatus, the possible disturbing influences and the measures to reduce their effects are discussed. The remaining corrections were either calculated or determined by auxiliary measurements and applied to the spectral shape. The sources of systematic errors are considered and included in the final results. We obtained a- is equal to -0.1017+-0.0051 giving

  9. Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

    Science.gov (United States)

    Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

    2014-02-01

    The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

  10. Sound response of superheated drop bubble detectors to neutrons

    International Nuclear Information System (INIS)

    Gao Size; Chen Zhe; Liu Chao; Ni Bangfa; Zhang Guiying; Zhao Changfa; Xiao Caijin; Liu Cunxiong; Nie Peng; Guan Yongjing

    2012-01-01

    The sound response of the bubble detectors to neutrons by using 252 Cf neutron source was described. Sound signals were filtered by sound card and PC. The short-time signal energy. FFT spectrum, power spectrum, and decay time constant were got to determine the authenticity of sound signal for bubbles. (authors)

  11. Radionuclide 252Cf neutron source

    International Nuclear Information System (INIS)

    Kolevatov, Yu.I.; Trykov, L.A.

    1979-01-01

    Characteristics of radionuclide neutron sourses of 252 Cf base with the activity from 10 6 to 10 9 n/s have been investigated. Energetic distributions of neutrons and gamma-radiation have been presented. The results obtained have been compared with other data available. The hardness parameter of the neutron spectrum for the energy range from 3 to 15 MeV is 1.4 +- 0.02 MeV

  12. NSDUAZ unfolding package for neutron spectrometry and dosimetry with Bonner spheres

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

    NSDUAZ (Neutron Spectrometry and Dosimetry for the Universidad Autonoma de Zacatecas) is a user friendly neutron unfolding package for Bonner sphere spectrometer with {sup 6}Lil(Eu) developed under Lab View environment. Unfolding is carried out using a recursive iterative procedure with the SPUNIT algorithm, where the starting spectrum is obtained from a library initial guess spectrum to start the iterations, the package include a statistical procedure based on the count rates relative to the count rate in the 8 inches-diameter sphere to select the initial spectrum. Neutron spectrum is unfolded in 32 energy groups ranging from 10{sup -8} up to 231.2 MeV. (Author)

  13. The study of prompt neutron spectra of 238U fission induced by fast neutron

    International Nuclear Information System (INIS)

    Li Anli; Bai Xixiang; Wang Yufeng; Wang Xiaozhong; Men Jiangchen; Huang Shengnian

    1990-01-01

    The measurements of prompt neutron time-of-flight spectra of U fission induced by 11 MeV neutrons were carried out at HI-13 Tandem Van de Graaff Accelerator Laboratory in 1989. The block diagram of the electronics is shown. A fission neutron TOF spectrum for the sixth section of the fission plates and the left detector at low bias is given. The data accumulation time is 60 h

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

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

  15. Evaluation of the 235U prompt fission neutron spectrum including a detailed analysis of experimental data and improved model information

    Science.gov (United States)

    Neudecker, Denise; Talou, Patrick; Kahler, Albert C.; White, Morgan C.; Kawano, Toshihiko

    2017-09-01

    We present an evaluation of the 235U prompt fission neutron spectrum (PFNS) induced by thermal to 20-MeV neutrons. Experimental data and associated covariances were analyzed in detail. The incident energy dependence of the PFNS was modeled with an extended Los Alamos model combined with the Hauser-Feshbach and the exciton models. These models describe prompt fission, pre-fission compound nucleus and pre-equilibrium neutron emissions. The evaluated PFNS agree well with the experimental data included in this evaluation, preliminary data of the LANL and LLNL Chi-Nu measurement and recent evaluations by Capote et al. and Rising et al. However, they are softer than the ENDF/B-VII.1 (VII.1) and JENDL-4.0 PFNS for incident neutron energies up to 2 MeV. Simulated effective multiplication factors keff of the Godiva and Flattop-25 critical assemblies are further from the measured keff if the current data are used within VII.1 compared to using only VII.1 data. However, if this work is used with ENDF/B-VIII.0β2 data, simulated values of keff agree well with the measured ones.

  16. COOLC, Ne-213 Liquid Scintillation Detector Neutron Spectra Unfolding

    International Nuclear Information System (INIS)

    1971-01-01

    1 - Nature of physical problem solved: COOLC is designed to calculate a neutron energy spectrum from a pulse-height spectrum produced by a detector system using the liquid scintillator NE-213. 2 - Method of solution: The program estimates the counts which would be observed in an ideal detector system having a response which is specified by the user. The solution implicitly takes into account the non-negativity of the desired neutron spectrum. The solution is obtained by finding a nearly optimal combination of slices through the spectrometer response functions such that their sum approximates the response of a channel of the ideal analyzer, and then uses the coefficients so determined to obtain an estimate of the desired neutron spectrum. 3 - Restrictions on the complexity of the problem: There are none noted

  17. Specification of fast neutron radiation quality from cell transformation data

    International Nuclear Information System (INIS)

    Coppola, M.

    1992-01-01

    Experimental data on the neoplastic transformation of C3H 10T1/2 cells measured at Casaccia after neutron and X-ray irradiation were used to determine neutron RBE values for the RSV-Tapiro fast reactor energy spectrum and for monoenergetic neutrons of 0.5, 1, and 6 MeV. In parallel, micro-dosimetric measurements provided the actual lineal energy distributions and related mean parameters for the reactor radiation. From these experiments, values of the neutron quality factor were derived for the reactor neutron energy spectrum and, in turn, for the other neutron energies tested. A mathematical expression giving a smooth dependence on neutron energy was also determined for the effective quality factor in the entire energy range examined. The results were compared with other proposals

  18. Integral test of JENDL dosimetry file using fast neutron field in the Experimental Fast Reactor JOYO

    International Nuclear Information System (INIS)

    Aoyama, Takafumi; Sekine, Takashi

    1999-09-01

    In order to evaluate the applicability of the JENDL dosimetry file, an integral test using a fast neutron spectrum field in the Experimental Fast Reactor JOYO Mark-II core was performed. The dosimeter set consisting of eight reactions of 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 58 Fe(n,γ) 59 Fe, 58 Ni(n,p) 58 Co, 59 Co(n,γ) 60 Co, 63 Cu(n,α) 60 Co, 238 U fission and 237 Np fission was irradiated for approximately 30 days near the core center of the JOYO Mk-II. Neutron flux at the dosimeter position was calculated using the two dimensional discrete ordinate transport code 'DORT'. The core configuration was modeled in XY geometry, and the 100 group cross section set of JSD-J2 / JFT-J2, which was processed from JENDL-2, was utilized. The absolute value of neutron flux was normalized so that the 235 U fission rate using the calculated neutron spectrum agreed with the measured reaction rate. The 103 group cross section data were processed by 'NJOY' code for nuclides to be used in the JOYO dosimetry. As the results of integral test for JENDL/D-99 (new file) and JENDL/D-91 (previous file), calculated values by JENDL/D-99 agreed well with the experimental values, and the C/E ratios ranged from 0.95 to 1.22. By comparing the results between JENDL/D-99 and JENDL/D-91, small differences exist, except for 58 Fe(n, γ) 59 Fe reaction, which was improved significantly in JENDL/D-99. (author)

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

    Science.gov (United States)

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

    2017-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-01

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

  1. Importance of the neutron spectrum for determination of radiation damage

    International Nuclear Information System (INIS)

    Hehn, G.; Stiller, P.; Mattes, M.

    1977-01-01

    Since the radiation effects of neutrons depend strongly on the neutron energy, the correlation between the induced damage and the fluence of the fast neutrons shows appreciable disadvantages. The measured values of changes in material properties resulted in large differences for the same fast neutron fluence, being partly due to different neutron spectra. The uncertainties in damage data led to strong overdesign of important structural components. Different neutron environment at surveillance sample position may give an underestimation of the embrittlement in the reactor pressure vessel, which has to be avoided. The application of damage functions combined with accurately calculated neutron spectra, promise to be a reasonable solution. The damage function has the advantage of a phenomenological quantity that all spectral effects are included. But the correlation quantity has to be determined of high experimental costs. Therefore approximations of its energy distributions are very important. For the keV energy region the kerma function is reasonably good. For the MeV energy region a higher effort is needed to calculate the displacement cross section. The same holds for the low energy part. In all three parts the formation of stable material property levels may vary, so that the final correlation can be determined only by measurements of material properties in different neutron spectra. In material samples the spectra distribution of the displacement production rate was determined at different local positions outside the reactor core of a PWR and a fast breeder showing the most important energy regions of both reactors. (orig.) [de

  2. Absolute instabilities of travelling wave solutions in a Keller-Segel model

    OpenAIRE

    Davis, P. N.; van Heijster, P.; Marangell, R.

    2016-01-01

    We investigate the spectral stability of travelling wave solutions in a Keller-Segel model of bacterial chemotaxis with a logarithmic chemosensitivity function and a constant, sublinear, and linear consumption rate. Linearising around the travelling wave solutions, we locate the essential and absolute spectrum of the associated linear operators and find that all travelling wave solutions have essential spectrum in the right half plane. However, we show that in the case of constant or sublinea...

  3. 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

  4. Non-destructive assay of fissile materials by detection and multiplicity analysis of spontaneous neutrons

    International Nuclear Information System (INIS)

    Prosdocimi, A.

    1979-01-01

    A method for determining the absolute reaction rate of nuclear events giving rise to neutron emission, according to their neutron multiplicity, is proposed. A typical application is the measurement of the (α, n) and spontaneous fission rates in a fissile material sample, particularly of Pu oxide composition. An analysis of random and correlated neutron pulses is carried out on the basis of sequential order without requiring any time interval analysis, then the primary nuclear events are sorted versus their neutron multiplicity. Suitable theoretical relationships enable to derive the absolute (α, n) and SF reaction rates when the physical parameters of the neutron detector and the multiplicity spectrumm of pulses are known. A typical device is described and the results of experiments leading to Pu-239 and Pu-240 assay are given

  5. Neutron activation studies on JET

    International Nuclear Information System (INIS)

    Loughlin, M.J.; Forrest, R.A.; Edwards, J.E.G.

    2001-01-01

    Extensive neutron transport calculations have been performed to determine the neutron spectrum at a number of points throughout the JET torus hall. The model has been bench-marked against a set of foil activation measurements which were activated during an experimental campaign with deuterium/tritium plasmas. The model can predict the neutron activation of the foils on the torus hall walls to within a factor of three for reactions with little sensitivity to thermal neutrons. The use of scandium foils with and without a cadmium thermal neutron absorber was a useful monitor of the thermal neutron flux. Conclusions regarding the usefulness of other foils for benchmarking the calculations are also given

  6. CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident

    International Nuclear Information System (INIS)

    Bessis, J.

    1976-01-01

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

  7. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    1999-12-31

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

  8. Fast neutron dosimetry in research reactors; Dosimetrie en neutrons rapides dans les reacteurs de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Eckert, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [French] On etudie principalement la mesure des flux de neutrons rapides a l'aide de detecteurs a seuil. On montre d'abord que les sections efficaces a utiliser pour les mesures par detecteurs a seuil, dependent grandement du spectre des neutrons, c'est-a-dire de l'emplacement ou s'effectue la mesure. La determination du spectre est effectuee par le calcul pour plusieurs emplacements des piles EL2 et EL3; on en deduit les sections efficaces a utiliser pour les mesures effectuees a ces emplacements. Dans la derniere partie du rapport, on indique quelles methodes sont possibles pour la determination experimentale du spectre. (auteur)

  9. Concept on coupled spectrum B/T (burning and/or transmutation) reactor for treatment of minor actinides by thermal and fast neutrons

    International Nuclear Information System (INIS)

    Aziz, Ferhat; Kitamoto, Asashi

    1996-01-01

    A conceptual design of B/T (burning and/or transmutation) reactor based on a modified conventional 1150 MWe-PWR system, with core consisted of two concentric regions for thermal and fast neutrons, was proposed herein for B/T treatment of MA (minor actinides). The B/T fuel considered was supposed su