Analytic computation of average energy of neutrons inducing fission

International Nuclear Information System (INIS)

Clark, Alexander Rich

2016-01-01

The objective of this report is to describe how I analytically computed the average energy of neutrons that induce fission in the bare BeRP ball. The motivation of this report is to resolve a discrepancy between the average energy computed via the FMULT and F4/FM cards in MCNP6 by comparison to the analytic results.

Resonance structure of 32S+n from measurements of neutron total and capture cross sections

International Nuclear Information System (INIS)

Halperin, J.; Johnson, C.H.; Winters, R.R.; Macklin, R.L.

1980-01-01

Neutron total and capture cross sections of 32 S have been measured up to 1100 keV neutron energy [E/sub exc/( 33 S) =9700 keV]. Spin and parity assignments have been made for 28 of the 64 resonances found in this region. Values of total radiation widths, reduced neutron widths, level spacings, and neutron strength functions have been evaluated for s/sub 1/2/, p/sub 1/2/, p/sub 3/2/, and d/sub 5/2/ levels. Single particle contributions using the valency model account for a significant portion of the total radiation width only for the p/sub 1/2/-wave resonances. A significant number of resonances can be identified with reported levels excited in 32 S(d,p) and 29 Si(α,n) reactions. A calculation of the Maxwellian average cross section appropriate to stellar interiors indicates an average capture cross section at 30 keV, sigma-bar approx. = 4.2(2) mb, a result that is relatively insensitive to the assumed stellar temperature. Direct (potential) capture and the s-wave resonance capture contributions to the thermal capture cross section do not fully account for the reported thermal cross section (530 +- 40 mb) and a bound state is invoked to account for the discrepancy

Energy dependence of average half-life of delayed neutron precursors in fast neutron induced fission of 235U and 236U

International Nuclear Information System (INIS)

Isaev, S.G.; Piksaikin, L.E.; Kazakov, L.E.; Tarasko, M.Z.

2000-01-01

The measurements of relative abundances and periods of delayed neutrons from fast neutron induced fission of 235 U and 236 U have been made at the electrostatic accelerator CG-2.5 at IPPE. The preliminary results were obtained and discussed in the frame of the systematics of the average half-life of delayed neutron precursors. It was shown that the average half-life value in both reactions depends on the energy of primary neutrons [ru

Determination of average activating thermal neutron flux in bulk samples

International Nuclear Information System (INIS)

Doczi, R.; Csikai, J.; Doczi, R.; Csikai, J.; Hassan, F. M.; Ali, M.A.

2004-01-01

A previous method used for the determination of the average neutron flux within bulky samples has been applied for the measurements of hydrogen contents of different samples. An analytical function is given for the description of the correlation between the activity of Dy foils and the hydrogen concentrations. Results obtained by the activation and the thermal neutron reflection methods are compared

Influence of orientation averaging on the anisotropy of thermal neutrons scattering on water molecules

International Nuclear Information System (INIS)

Markovic, M. I.; Radunovic, J. B.

1976-01-01

Determination of spatial distribution of neutron flux in water, most frequently used moderator in thermal reactors, demands microscopic scattering kernels dependence on cosine of thermal neutrons scattering angle when solving the Boltzmann equation. Since spatial orientation of water molecules influences this dependence it is necessary to perform orientation averaging or rotation-vibrational intermediate scattering function for water molecules. The calculations described in this paper and the obtained results showed that methods of orientation averaging do not influence the anisotropy of thermal neutrons scattering on water molecules, but do influence the inelastic scattering

MADNIX a code to calculate prompt fission neutron spectra and average prompt neutron multiplicities

International Nuclear Information System (INIS)

Merchant, A.C.

1986-03-01

A code has been written and tested on the CDC Cyber-170 to calculate the prompt fission neutron spectrum, N(E), as a function of both the fissioning nucleus and its excitation energy. In this note a brief description of the underlying physical principles involved and a detailed explanation of the required input data (together with a sample output for the fission of 235 U induced by 14 MeV neutrons) are presented. Weisskopf's standard nuclear evaporation theory provides the basis for the calculation. Two important refinements are that the distribution of fission-fragment residual nuclear temperature and the cooling of the fragments as neutrons are emitted approximately taken into account, and also the energy dependence of the cross section for the inverse process of compound nucleus formation is included. This approach is then used to calculate the average number of prompt neutrons emitted per fission, v-bar p . At high excitation energies, where fission is still possible after neutron emission, the consequences of the competition between first, second and third chance fission on N(E) and v-bar p are calculated. Excellent agreement with all the examples given in the original work of Madland and Nix is obtained. (author) [pt

Determination of average fission fraction produced by 14 MeV neutrons in assemblies with large volume of depleted uranium

International Nuclear Information System (INIS)

Wang Dalun; Li Benci; Wang Xiuchun; Li Yijun; Zhang Shaohua; He Yongwu

1991-07-01

The average fission fraction of 238 U caused by 14 MeV neutrons in assemblies with large volume depleted uranium has been determined. The measured value of p f 238U (R ∞ depleted ) 14 was 0.897 ± 0.036. Measurements were also completed for neutron flux distribution and average fission fraction of 235 U isotope in depleted uranium sphere. Values of p f 238U (R depleted ) have been obtained by using a series of uranium spheres. For a sphere with Φ 600 the p f 23 '8 U (R 300 depleted ) is 0.823 ± 0.041, the density of depleted uranium assembly is 18.8g/cm 3 and total weight of assembly is about 2.8t

Method for analysis of averages over transmission energy of resonance neutrons

International Nuclear Information System (INIS)

Komarov, A.V.; Luk'yanov, A.A.

1981-01-01

Experimental data on transmissions on iron specimens in different energy groups have been analyzed on the basis of an earlier developed theoretical model for the description of resonance neutron averages in transmission energy, as the functions of specimen thickness and mean resonance parameters. The parameter values obtained agree with the corresponding data evaluated in the theory of mean neutron cross sections. The method suggested for the transmission description permits to reproduce experimental results for any thicknesses of specimens [ru

Modified model of neutron resonance widths distribution. Results of total gamma-widths approximation

International Nuclear Information System (INIS)

Sukhovoj, A.M.; Khitrov, V.A.

2011-01-01

Functional dependences of probability to observe given Γ n 0 value and algorithms for determination of the most probable magnitudes of the modified model of resonance parameter distributions were used for analysis of the experimental data on the total radiative widths of neutron resonances. As in the case of neutron widths, precise description of the Γ γ spectra requires a superposition of three and more probability distributions for squares of the random normally distributed values with different nonzero average and nonunit dispersion. This result confirms the preliminary conclusion obtained earlier at analysis of Γ n 0 that practically in all 56 tested sets of total gamma widths there are several groups noticeably differing from each other by the structure of their wave functions. In addition, it was determined that radiative widths are much more sensitive than the neutron ones to resonance wave functions structure. Analysis of early obtained neutron reduced widths distribution parameters for 157 resonance sets in the mass region of nuclei 35 ≤ A ≤ 249 was also performed. It was shown that the experimental values of widths can correspond with high probability to superposition of several expected independent distributions with their nonzero mean values and nonunit dispersion

Total neutron cross sections of berkelium-249 and californium-249 below 100 eV

International Nuclear Information System (INIS)

Benjamin, R.W.; Harvey, J.A.; Hill, N.W.; Pandey, M.S.; Carlton, R.F.

1979-01-01

The neutron total cross sections of 249 Bk and 249 Cf have been measured from 0.03 to 100 eV using the Oak Ridge Electron Linear Accelerator (ORELA) as a source of pulsed neutrons. The 1.6 mm dia. cylindrical transmission samples contained initially up to 5.3 mg of 98% 249 Bk and 2% 249 Cf: 4.5 years later, when the final measurements were made, the composition of the samples had become 2.5% 249 Bk, 96.9% 249 Cf, and 0.6% 245 Cm. Samples were cooled with liquid nitrogen to reduce Doppler broadening. Thirty-nine resonances were identified in 249 Bk and analyzed using a single-level Breit-Wigner formalism. Fifty-five resonances were identified in 249 Cf and analyzed using an R-matrix multilevel formalism. Fifty-five resonances were identified in 249 Cf and analyzed using an R-matrix multilevel formalism. The resonance parameters obtained have been used to determine the average level spacings and the s-wave neutron and fission strength functions. Where possible, bound-level parameters were derived to fit the thermal neutron total cross section data

What Do s- and p-Wave Neutron Average Radiative Widths Reveal

Energy Technology Data Exchange (ETDEWEB)

Mughabghab, S.F.

2010-04-30

A first observation of two resonance-like structures at mass numbers 92 and 112 in the average capture widths of the p-wave neutron resonances relative to the s-wave component is interpreted in terms of a spin-orbit splitting of the 3p single-particle state into P{sub 3/2} and P{sub 1/2} components at the neutron separation energy. A third structure at about A = 124, which is not correlated with the 3p-wave neutron strength function, is possibly due to the Pygmy Dipole Resonance. Five significant results emerge from this investigation: (i) The strength of the spin-orbit potential of the optical-model is determined as 5.7 {+-} 0.5 MeV, (ii) Non-statistical effects dominate the p-wave neutron-capture in the mass region A = 85 - 130, (iii) The background magnitude of the p-wave average capture-width relative to that of the s-wave is determined as 0.50 {+-} 0.05, which is accounted for quantitatively in tenns of the generalized Fermi liquid model of Mughabghab and Dunford, (iv) The p-wave resonances arc partially decoupled from the giant-dipole resonance (GDR), and (v) Gamma-ray transitions, enhanced over the predictions of the GDR, are observed in the {sup 90}Zr - {sup 98}Mo and Sn-Ba regions.

The definition and computation of average neutron lifetimes

International Nuclear Information System (INIS)

Henry, A.F.

1983-01-01

A precise physical definition is offered for a class of average lifetimes for neutrons in an assembly of materials, either multiplying or not, or if the former, critical or not. A compact theoretical expression for the general member of this class is derived in terms of solutions to the transport equation. Three specific definitions are considered. Particular exact expressions for these are derived and reduced to simple algebraic formulas for one-group and two-group homogeneous bare-core models

Research of isolated resonances using the average energy shift method for filtered neutron beam

International Nuclear Information System (INIS)

Gritzay, O.O.; Grymalo, A.K.; Kolotyi, V.V.; Mityushkin, O.O.; Venediktov, V.M.

2010-01-01

This work is devoted to detailed description of one of the research directions in the Neutron Physics Department (NPD), namely, to research of resonance parameters of isolated nuclear level at the filtered neutron beam on the horizontal experimental channel HEC-8 of the WWR-M reactor. Research of resonance parameters is an actual problem nowadays. This is because there are the essential differences between the resonance parameter values in the different evaluated nuclear data library (ENDL) for many nuclei. Research of resonance parameter is possible due to the set of the neutron cross sections received at the same filter, but with the slightly shifted filter average energy. The shift of the filter average energy is possible by several processes. In this work this shift is realized by neutron energy dependence on scattering angle. This method is provided by equipment.

One-speed neutron transport in spheres with totally absorbing cores

International Nuclear Information System (INIS)

Sjoestrand, N.G.

1988-01-01

Stationary and time-dependent transport of neutrons of one speed has been studied in spheres with totally absorbing cores. For stationary, critical reactors the number of secondaries per collision has been calculated numerically for various inner and outer radii. In the time-dependent case, the decay constant has been calculated for spherical shells of different inner radii and thicknesses. For a fixed ratio between shell thickness and inner radius, the curve of the decay constant versus shell thickness crosses the Corngold limit in the same way as the curve for a homogeneous sphere. When the ratio goes to zero the curve approaches that for an infinite slab. The behaviour is discussed in view of a new result from collision theory, viz. that the following condition must be fulfilled for a body at the point where the decay constant curve crosses the Corngold limit: the average exit distance of the neutrons is equal to the mean free path for scattering

Neutron total cross section measurements on 249Cf

International Nuclear Information System (INIS)

Carlton, R.F.; Harvey, J.A.; Hill, N.W.; Pandey, M.S.; Benjamin, R.W.

1979-01-01

Neutron total cross section measurements were performed on a sample of 249 Cf (5.65 mg total weight) with the ORELA as a source of pulsed neutrons. The sample, the inverse thickness of which was 1542 barns/atom, consisted of 85.3% 249 Cf and 14.4% 249 Bk, and was cooled to liquid nitrogen temperature. Analyses were also made of data from a thin sample (l/n = 17430) of 65% 249 Cf in the region of the large fission resonance at 0.7 eV. Fifty-five resonances in 249 Cf were observed and analyzed over the energy range 0.1 eV to 90 eV by use of an R-matrix multilevel formalism. The resonance parameters obtained were used to determine the level spacing and the s-wave neutron and fission strength functions. Thermal total cross section measurements were also performed. 5 figures, 3 tables

Total variation-based neutron computed tomography

Science.gov (United States)

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

2018-05-01

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

Measurement of neutron total cross-sections for {sup nat}Dy at Pohang Neutron Facility

Energy Technology Data Exchange (ETDEWEB)

Shin, S. G.; Kye, Y. U.; Shvetsov, Valery; Cho, M. H. [POSTECH, Pohang (Korea, Republic of); Namkung, W.; Cho, M. H. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Kim, G. N. [Kyungpook National Univ., Daegu (Korea, Republic of); Lee, M. W. [Dongnam Inst. of radiological and Medical Science, Busan (Korea, Republic of)

2013-05-15

There are few measurements for Dy below 100 eV. Moreover, there exist discrepancies among the measurements. In the present work, the total neutron cross-sections for {sup nat}Dy were measured by using the time-of-flight (TOF) method at the Pohang Neutron Facility (PNF). The PNF consists of an electron linac, a water-cooled Ta target, and an 11-m-long TOF path. The characteristics of PNF are described elsewhere. We also briefly discuss the future plan to verify our experimental result. We have measured the total neutron cross-sections of {sup nat}Dy in the neutron energy region from 0.1 eV to 100 eV with the TOF method at the Po hang Neutron Facility. The present result is in good agreement with the previous data and the evaluated data in ENDF/B-VI. We would like to get resonance parameters by using SAMMY or REFIT codes.

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)

Standard Test Method for Measuring Neutron Fluence and Average Energy from 3H(d,n)4He Neutron Generators by Radioactivation Techniques 1

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This test method covers a general procedure for the measurement of the fast-neutron fluence rate produced by neutron generators utilizing the 3H(d,n)4He reaction. Neutrons so produced are usually referred to as 14-MeV neutrons, but range in energy depending on a number of factors. This test method does not adequately cover fusion sources where the velocity of the plasma may be an important consideration. 1.2 This test method uses threshold activation reactions to determine the average energy of the neutrons and the neutron fluence at that energy. At least three activities, chosen from an appropriate set of dosimetry reactions, are required to characterize the average energy and fluence. The required activities are typically measured by gamma ray spectroscopy. 1.3 The measurement of reaction products in their metastable states is not covered. If the metastable state decays to the ground state, the ground state reaction may be used. 1.4 The values stated in SI units are to be regarded as standard. No oth...

Neutron total scattering cross sections of elemental antimony

Energy Technology Data Exchange (ETDEWEB)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-11-01

Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V.

Neutron total scattering cross sections of elemental antimony

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-11-01

Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V

Porosity effects in the neutron total cross section of graphite

International Nuclear Information System (INIS)

Santisteban, J. R; Dawidowski, J; Petriw, S. N

2009-01-01

Graphite has been used in nuclear reactors since the birth of the nuclear industry due to its good performance as a neutron moderator material. Graphite is still an option as moderator for generation IV reactors due to its good mechanical and thermal properties at high operation temperatures. So, there has been renewed interest in a revision of the computer libraries used to describe the neutron cross section of graphite. For sub-thermal neutron energies, polycrystalline graphite shows a larger total cross section (between 4 and 8 barns) than predicted by existing theoretical models (0.2 barns). In order to investigate the origin of this discrepancy we measured the total cross section of graphite samples of three different origins, in the energy range from 0.001 eV to 10 eV. Different experimental arrangements and sample treatments were explored, to identify the effect of various experimental parameters on the total cross section measurement. The experiments showed that the increase in total cross section is due to neutrons scattered around the forward direction. We associate these small-angle scattered neutrons (SANS) to the porous structure of graphite, and formulate a very simple model to compute its contribution to the total cross section of the material. This results in an analytic expression that explicitly depends on the density and mean size of the pores, which can be easily incorporated in nuclear library codes. [es

The total neutron cross-section of Nb at different temperatures for neutrons with energies below 1 eV

International Nuclear Information System (INIS)

Adib, M.; Abdel-Kawy, A.; Maayouf, R.M.A.; Fayek, M.; Mostafa, M.; Hamouda, I.

1981-09-01

Total neutron cross-section measurements have been performed for natural Nb at liquid nitrogen, room and 425 0 K temperatures in the energy range from 2 MeV - 1 eV. The measurements were performed using two time-of-flight spectrometers installed in front of two of the ET-RR-1 reactor horizontal channels. The neutron diffraction pattern of Nb, at room temperature, was obtained using a double axis crystal spectrometer installed also at the ET-RR-1 reactor. The obtained total neutron cross-sections were analyzed using the single level Breit-Wigner formula. The coherent scattering amplitude was determined from the Bragg reflections observed in the total neutron cross-section of Nb and the analysis of its neutron diffraction pattern. The incoherent and thermal inelastic scattering cross-sections of Nb were determined from the analysis of the total cross-section of Nb beyond the cut-off wavelength. The following results have been obtained: sigmasub(t) = (6.30+-0.20)b; sigmasub(coh) = (6.0+-0.3)b; sigmasub(incoh) = (2.0+-1.0)b; bsub(coh) = (6.91+-0.08)fm

Fast-neutron total and scattering cross sections of 58Ni

International Nuclear Information System (INIS)

Jorgensen, C.B.; Guenther, P.T.; Smith, A.B.; Whalen, J.F.

1981-09-01

Neutron total cross sections of 58 Ni were measured at 25 keV intervals from 0.9 to 4.5 MeV with 50 to 100 keV resolutions. Attention was given to self-shielding corrections to the observed total cross sections. Differential elastic- and inelastic-scattering cross sections were measured at 50 keV intervals from 1.35 to 4.0 MeV with 50 to 100 keV resolutions. Inelastic excitation of levels at 1.458 +- 0.009, 2.462 +- 0.010, 2.791 +- 0.015, 2.927 +- 0.012 and 3.059 +- 0.025 MeV was observed. The experimental results were interpreted in terms of optical-statistical and coupled-channels models. A spherical optical-statistical model was found generally descriptive of an energy-average of the experimental results. However, detailed considerations suggested significant contributions from direct-vibrational interactions, particularly associated with the excitation of the first 2+ level

4π-spectrometer technique for measurements of secondary neutron average number in nuclear fission by 252Cf neutrons

International Nuclear Information System (INIS)

Vasil'ev, Yu.A.; Barashkov, Yu.A.; Golovanov, O.A.; Sidorov, L.V.

1977-01-01

A method for determining the average number of secondary neutrons anti ν produced in nuclear fission by the neutrons of the 252 Cf fission spectra by means of a 4π time-of-flight spectrometer is described. Layers of 252 Cf and an isotope studied are placed close to each other; if the isotope layer density is 1 mg/cm 2 probability of its fission is about 10 -5 per one spontaneous fission of californium. Fission fragments of 252 Cf and the isotope investigated have been detected by two surface-barrier counters with an efficiency close to 100%. The layers and the counters are situated in a measuring chamber placed in the center of the 4π time-of-flight spectrometer. The latter is utilized as a neutron counter because of its fast response. The method has been verified by carrying out measurements for 235 U and 239 Pu. A comparison of the experimental and calculated results shows that the method suggested can apply to determine the number of secondary neutrons in fission of isotopes that have not been investigated yet

Total neutron cross section of lead

International Nuclear Information System (INIS)

Kanda, K.; Aizawa, O.

1976-01-01

The total thermal-neutron cross section of natural lead under various physical conditions was measured by the transmission method. It became clear that the total cross section at room temperature previously reported is lower than the present data. The total cross section at 400, 500, and 600 0 C, above the melting point of lead, 327 0 C, was also measured, and the changes in the cross section as a function of temperature were examined, especially near and below the melting point. The data obtained for the randomly oriented polycrystalline state at room temperature were in reasonable agreement with the theoretical values calculated by the THRUSH and UNCLE-TOM codes

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

ORLIB: a computer code that produces one-energy group, time- and spatially-averaged neutron cross sections

International Nuclear Information System (INIS)

Blink, J.A.; Dye, R.E.; Kimlinger, J.R.

1981-12-01

Calculation of neutron activation of proposed fusion reactors requires a library of neutron-activation cross sections. One such library is ACTL, which is being updated and expanded by Howerton. If the energy-dependent neutron flux is also known as a function of location and time, the buildup and decay of activation products can be calculated. In practice, hand calculation is impractical without energy-averaged cross sections because of the large number of energy groups. A widely used activation computer code, ORIGEN2, also requires energy-averaged cross sections. Accordingly, we wrote the ORLIB code to collapse the ACTL library, using the flux as a weighting function. The ORLIB code runs on the LLNL Cray computer network. We have also modified ORIGEN2 to accept the expanded activation libraries produced by ORLIB

Estimation of total as well as bioaccessible levels and average daily dietary intake of iodine from Japanese edible seaweeds by epithermal neutron activation analysis

International Nuclear Information System (INIS)

Fukushima, M.; Chatt, A.

2012-01-01

An epi-thermal instrumental neutron activation analysis (EINAA) method in conjunction with Compton suppression spectrometry (EINAA-CSS) was used for the determination of total iodine in eight different species of edible seaweeds from Japan. This method gave an absolute detection limit of about 2 μg. The accuracy of the method was evaluated using various reference materials and found to be generally in agreement within ±6% of the certified values. The longitudinal distributions of iodine at different growing stages in Japanese sea mustard and tangle seaweeds were investigated. For a 150-cm-high tangle, the highest concentration (5,360 mg/kg) of iodine was found at the root, then decreased slowly to 780 mg/kg in the middle portion (60-75 cm), and increased to 2,300 mg/kg at the apex. On the other hand, for a 190-cm-high sea mustard the highest levels of iodine were found both at the roots (164 mg/kg) and apex (152 mg/kg) with lower values (98 mg/kg) in the middle section. In order to estimate the bioaccessible fraction of iodine, seaweeds were digested by an in vitro enzymolysis method, dietary fibre separated from residue, and both fractions analyzed by EINAA-CSS. The average daily dietary intakes of total (0.14 mg) as well as bioaccessible fraction (0.12 mg) of iodine from the consumption of sea mustards were estimated. (author)

Analysis of total hydrogen content in palm oil and palm kernel oil using thermal neutron moderation method

International Nuclear Information System (INIS)

Akaho, E.H.K.; Dagadu, C.P.K.; Maaku, B.T.; Anim-Sampong, S.; Kyere, A.W.K.; Jonah, S.A.

2001-01-01

A fast and non-destructive technique based on thermal neutron moderation has been used for determining the total hydrogen content in two types of red palm oil (dzomi and amidze) and palm kernel oil produced by traditional methods in Ghana. An equipment consisting of an 241 Am-Be neutron source and 3 He neutron detector was used in the investigation. The equipment was originally designed for detection of liquid levels in petrochemical and other process industries. Standards in the form of liquid hydrocarbons were used to obtain calibration lines for thermal neutron reflection parameter as a function of hydrogen content. Measured reflection parameters with respective hydrogen content with or without heat treatment of the three edible palm oils available on the market were compared with a brand cooking oil (frytol). The average total hydrogen content in the local oil samples prior to heating was measured to be 11.62 w% which compared well with acceptable value of 12 w% for palm oils in the sub-region. After heat treatment, the frytol oil (produced through bleaching process) had the least loss of hydrogen content of 0.26% in comparison with palm kernel oil of 0.44% followed by dzomi of 1.96% and by amidze of 3.22%. (author)

Measurements of the total neutron cross-sections of poly- and mono-germanium crystals at neutron energies below 1 eV

International Nuclear Information System (INIS)

Maayouf, R.M.A.; Abdel-Kawy, A.; Abbas, Y.; Habib, N.; Adib, M.; Hamouda, I.

1983-12-01

Total neutron cross-section measurements have been performed for poly and mono-germanium crystals in the energy range from 2 meV-1eV. The measurements were performed using two TOF and a double axis crystal spectrometer installed at the ET-RR-1 reactor. The obtained neutron cross-sections were analyzed using the single level Breit-Wigner formula. The coherent scattering amplitude was determined from the Bragg reflections observed in the total neutron cross-section of Ge and the analysis of its neutron diffraction pattern. The incoherent and thermal diffuse scattering cross-sections of Ge were estimated from the analysis of the total cross-section data obtained for Ge mono-crystal

The evaluation of neutron total cross section for natural iron and aluminium

International Nuclear Information System (INIS)

Liu Shirui; Wang Chunhao; Zhao Defang

1990-05-01

The experimental data of total cross section were collected and evaluated for natural iron in the energy region from 1 keV to 20 MeV and for natural aluminium from 4.07 keV to 20 MeV. The evaluated data were recommended in the regions for them. The minimum values of Fe total cross section in the keV region were specially recommended. The resonance structures were briefly discussed for both Fe and Al. To make the evaluation better, all experimental measurements of neutron total cross section relative to Fe and Al were studied. Considering the resonance feature of medium weight nuclides, two criteria for selecting total cross section were presented: 1) the correlation between the precission of total cross section and neutron source; 2) the correlation between the accuracy of total cross section and the resolving power of the neutron spectrometer

Determination of the fission-neutron averaged cross sections of some high-energy threshold reactions of interest for reactor dosimetry

International Nuclear Information System (INIS)

Arribere, M.A.; Kestelman, A.J.; Korochinsky, S.; Blostein, J.J.

2003-01-01

For three high threshold reactions, we have measured the cross sections averaged over a 235 U fission neutron spectrum. The measured reactions, and corresponding averaged cross sections found, are: 127 I(n,2n) 126 I, (1.36±0.12) mb; 90 Zr(n,2n) 89m Zr, (13.86±0.83) μb; and 58 Ni(n,d+np+pn) 57 Co, (274±15) μb; all referred to the well known standard of (111±3) mb for the 58 Ni(n,p) 58m+g Co averaged cross section. The measured cross sections are of interest in nuclear engineering for the characterization of the fast neutron component in the energy distribution of reactor neutrons. (author)

Neutron total cross section measurements of gold and tantalum at the nELBE photoneutron source

CERN Document Server

Hannaske, Roland; Beyer, Roland; Junghans, Arnd; Bemmerer, Daniel; Birgersson, Evert; Ferrari, Anna; Grosse, Eckart; Kempe, Mathias; Kögler, Toni; Marta, Michele; Massarczyk, Ralph; Matic, Andrija; Schramm, Georg; Schwengner, Ronald; Wagner, Andreas

2014-01-01

Neutron total cross sections of 197 Au and nat Ta have been measured at the nELBE photoneutron source in the energy range from 0.1 - 10 MeV with a statistical uncertainty of up to 2 % and a total systematic uncertainty of 1 %. This facility is optimized for the fast neutron energy range and combines an excellent t ime structure of the neutron pulses (electron bunch width 5 ps) with a short flight path of 7 m. Because of the low instantaneous neutron flux transmission measurements of neutron total cross sections are possible, that exhibit very different beam and back ground conditions than found at other neutron sources.

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

High resolution neutron total and capture cross-sections in separated isotopes of copper (6365Cu)

International Nuclear Information System (INIS)

Pandey, M.S.

1975-01-01

High resolution neutron total and capture cross section measurements have been performed on separated isotopes of copper ( 63 65 Cu). Measurements for capture cross section were made from about 1 keV to a few hundreds of keV. The total cross section measurements were made in the energy interval of approximately 10 keV to 150 keV. The resulting capture data have been analyzed by a generalized least square peak fitting computer code in the energy interval of 2.5 keV to 50 keV. Photon strengths are determined using the data up to approximately 250 keV. The resulting total cross section data have been analyzed by area-analysis on the transmission values and by R-matrix multilevel code on cross section values. Average s- and p-wave level spacing and s- and p-wave strength function values are determined. From the resonance parameters thus obtained, by the analysis, statistical distribution is studied for s- and p-wave level spacings and reduced neutron widths. A comparison has been made for adjacent level spacings with the theoretical predictions of level repulsion (of same J/sup π/) by Wigner considering levels with various spin states separately for s-wave resonances where confident spin assignment has been possible. Reduced neutron widths are compared with the Porter-Thomas distribution. Optical model formulated by Feshbach, Porter and Weiskopf describes the neutron-nucleus interaction. A comparison has been made between experimentally determined values of the s- and p-wave strength functions and that obtainable from optical model calculations, thereby determining the appropriate optical model parameters. The experimental arrangement, pertinent theoretical discussion, and the processes of data reduction and the analyses along with the comparison of the previously reported results with the present work are presented in detail

Structure analysis of liquids and disordered materials using pulsed neutron diffraction and total scattering

International Nuclear Information System (INIS)

Suzuya, Kentaro

2011-01-01

Neutron diffraction·total scattering at pulsed neutron source is a powerful method to analyze the complex structure of disordered materials: liquids, glasses, amorphous materials and disordered crystals. The basic idea of the structure of disordered materials, the fundamental diffraction theory for disordered materials, and structure analysis of disordered materials using pulsed neutron diffraction·total scattering technique (TOF method) are described in detail. In addition, the precise information of the world highest class J-PARC MLF spallation neutron source and typical J-PARC neutron total scattering instrument NOVA are also given. Recent structural modelling methods of disordered materials such like reverse Monte Carlo (RMC) simulation method is briefly described using an example of the analysis of a typical disordered material silica glass. (author)

EDITAR: a module for reaction rate editing and cross-section averaging within the AUS neutronics code system

International Nuclear Information System (INIS)

Robinson, G.S.

1986-03-01

The EDITAR module of the AUS neutronics code system edits one and two-dimensional flux data pools produced by other AUS modules to form reaction rates for materials and their constituent nuclides, and to average cross sections over space and energy. The module includes a Bsub(L) flux calculation for application to cell leakage. The STATUS data pool of the AUS system is used to enable the 'unsmearing' of fluxes and nuclide editing with minimal user input. The module distinguishes between neutron and photon groups, and printed reaction rates are formed accordingly. Bilinear weighting may be used to obtain material reactivity worths and to average cross sections. Bilinear weighting is at present restricted to diffusion theory leakage estimates made using mesh-average fluxes

The total neutron cross sections for 14N and 24Mg

International Nuclear Information System (INIS)

Bommer, J.

This report contains tables of the total neutron cross sections of 14 N and 24 Mg as determined in a recent measurement for neutron energies between 1 and 5.3 MeV. Graphic representations and details on the evaluation of the cross sections are included. (orig.) [de

Observation of neutron standing waves at total reflection by precision gamma spectroscopy

International Nuclear Information System (INIS)

Aksenov, V.L.; Gundorin, N.A.; Nikitenko, Yu.V.; Popov, Yu.P.; Cser, L.

1998-01-01

Total reflection of polarized neutrons from the layered structure glass/Fe (1000 A Angstrom)/Gd (50 A Angstrom) is investigated by registering neutrons and gamma-quanta from thermal neutron capture. The polarization ratio of gamma counts of neutron beams polarized in and opposite the direction of the magnetic field is measured. The polarization ratio is larger than unity for the neutron wavelengths λ 2.2 A Angstrom. Such behaviour of the wavelength dependence of the gamma-quanta polarization ratio points to the fact that over the surface of the Fe Layer a neutron standing wave caused by the interference of the incident neutron wave and the wave refracted from the magnetized Fe layer is formed

Expected total counts for the Self-Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Rossa, Riccardo [Belgian nuclear research centre SCK.CEN (Belgium); Universite Libre de Bruxelles (Belgium); Borella, Alessandro; Van der Meer, Klaas [Belgian nuclear research centre SCK.CEN. Boeretang 200, 2400 Mol (Belgium); Labeau, Pierre-Etienne; Pauly, Nicolas [Universite Libre de Bruxelles. Av. F. D. Roosevelt 50, B1050 Brussels (Belgium)

2015-07-01

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in spent fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron counts in the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach in this study consisted in introducing a small neutron detector in the central guide tube of a PWR 17x17 fuel assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types were used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the total neutron counts that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of total neutron counts and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the total neutron counts by increasing the detector size. The study shows that the highest total neutron counts are achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the total neutron counts

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

International Nuclear Information System (INIS)

Syromyatnikov, V.G.; Zaw Lin, Kyaw

2017-01-01

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

Analysis of average radiation widths of neutron resonances

International Nuclear Information System (INIS)

Malezki, H.; Popov, A.B.; Trzeciak, K.

1982-01-01

On the basis of the available data on parameters of neutron resonances average values of radiation widths (GITAsub(γ)) are calculated for a wide range of nuclei in the 50 upto 250 atomic weight range. Experimental values are compared with different variants of theoretical estimates of GITAsub(γ) which are reduced to the GITAsub(γ) dependence upon atomic weight A, excitation energy U and level density parameter a as GITAsub(γ)=CAsup(α)Usup(β)asup(γ). Besides, empirical values C, α, β, γ are selected satisfying the experimental data best of all. It is determined that the use of a=kA hypothesis leads to a sufficiently better agreement between all theoretical estimates of GITAsub(γ) and experimental values. It turned out that the estimations by Weisskopf, Bondarenko-Urin or with empirically chosen parameters give an approximately similar correspondence of calculated values GITAsub(γ)sup(p) to experimental data [ru

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)

Neutron reflection effect on total absorption detector method used in SWINPC neutron multiplication experiment for beryllium

International Nuclear Information System (INIS)

Tian Dongfeng; Ho Yukun; Yang Fujia

2001-01-01

The SWINPC integral experiment on neutron multiplication in bulk beryllium showed that there were marked discrepancies between experimental data and calculated values with the ENDF/B-VI data. The calculated values become higher than experimental ones as the sample thickness increases. Several works had been devoted to find problems existing in the experiment. This paper discusses the neutron reflection effect on the total absorption detector method which was used in the experiment to measure the neutron leakage from samples. One systematic correction is suggested to make the experimental values agree with the calculated ones with the ENDF/B-VI data within experimental errors. (author)

Measurement of total reaction cross sections of exotic neutron rich nuclei

International Nuclear Information System (INIS)

Mittig, W.; Chouvel, J.M.; Wen Long, Z.

1987-01-01

Total reaction cross-sections of neutron rich nuclei from C to Mg in a thick Si-target have been measured using the detection of the associated γ-rays in a 4Π-geometry. This cross-section strongly increases with neutron excess, indicating an increase of as much as 15% of the reduced strong absorption radius with respect to stable nuclei

Evaluations of average level spacings

International Nuclear Information System (INIS)

Liou, H.I.

1980-01-01

The average level spacing for highly excited nuclei is a key parameter in cross section formulas based on statistical nuclear models, and also plays an important role in determining many physics quantities. Various methods to evaluate average level spacings are reviewed. Because of the finite experimental resolution, to detect a complete sequence of levels without mixing other parities is extremely difficult, if not totally impossible. Most methods derive the average level spacings by applying a fit, with different degrees of generality, to the truncated Porter-Thomas distribution for reduced neutron widths. A method that tests both distributions of level widths and positions is discussed extensivey with an example of 168 Er data. 19 figures, 2 tables

New Maxwellian averaged neutron capture cross sections for 35,37Cl

International Nuclear Information System (INIS)

Guber, K.H.; Sayer, R.O.; Valentine, T.E.; Leal, L.C.; Spencer, R.R.; Harvey, J.A.; Koehler, P.E.; Rauscher, T.

2002-01-01

The Oak Ridge Electron Linear Accelerator (ORELA) was used to measure neutron total and capture cross sections of natural chlorine in the energy range from 100 eV to 600 keV. We performed an R-matrix analysis of our new capture and transmission data up to 500 keV. From these resonance parameters new (n,γ) astrophysical reaction rates were determined over the entire energy range needed by the latest stellar models of the s process

Average number of neutrons in π-p, π-n, and π-12C interactions at 4 GeV/c

International Nuclear Information System (INIS)

Bekmirzaev, R.N.; Grishin, V.G.; Muminov, M.M.; Suvanov, I.; Trka, Z.; Trkova, J.

1984-01-01

The average numbers of secondary neutrons in π - p, π - n, and π -12 C interactions at 4 GeV/c have been determined by investigating secondary neutral stars produced by neutrons in a propane bubble chamber. The following values were obtained for the charge-exchange coefficients: α(p→n) = 0.39 +- 0.04 and α(n→p) = 0.37 +- 0.08

Simultaneous measurement of neutrons and fission fragments of thermal neutron fission of U-233

International Nuclear Information System (INIS)

Itsuro Kimura; Katsuhisa Nishio; Yoshihiro Nakagome

2000-01-01

The multiplicity and the energy of prompt neutrons from the fragments for 233 U(n th , f) were measured as functions of fragment mass and total kinetic energy. Average neutron energy against the fragment mass showed a nearly symmetric distribution about the half mass division with two valleys at 98 and 145 u. The slope of the neutron multiplicity with total kinetic energy depended on the fragment mass and showed the minimum at about 130 u. The obtained neutron data were applied to determine the total excitation energy of the system, and the resulting value in the typical asymmetric fission lied between 22 and 25 MeV. The excitation energy agreed with that determined by subtracting the total kinetic energy from the Q-value within 1 MeV, thus satisfied the energy conservation. In the symmetric fission, where the mass yield was drastically suppresses, the total excitation energy is significantly large and reaches to about 40 MeV, suggesting that fragment pairs are preferentially formed in a compact configuration at the scission point [ru

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

Generation of neutron standing waves at total reflection of polarized neutrons

International Nuclear Information System (INIS)

Aksenov, V.L.; Nikitenko, Yu.V.; Kozhevnikov, S.V.; Radu, F.; Kruijs, R.; Rekveldt, M.Th.

1999-01-01

The regime of neutron standing waves at reflection of polarized thermal neutrons from the structure glass/Cu (1000 A Angstrom)/Ti (2000 A Angstrom)/Co (60 A Angstrom)/Ti (300 A Angstrom) in a magnetic field directed at an angle to the sample plane is realized. The intensity of neutrons with a particular spin projection on the external magnetic field direction appears to be a periodic function of the neutron wavelength and the glancing angle of the reflected beam. It is shown that the neutron standing wave regime can be a very sensitive method for the determination of changes in the spatial position of magnetic noncollinear layers. (author)

Neutron total and scattering cross sections of 6Li in the few MeV region

International Nuclear Information System (INIS)

Smith, A.; Guenther, P.; Whalen, J.

1980-02-01

Neutron total cross sections of 6 Li are measured from approx. 0.5 to approx. 4.8 MeV at intervals of approx. 10 scattering angles and at incident-neutron intervals of approx.< 100 keV. Neutron differential inelastic-scattering cross sections are measured in the incident-energy range 3.5 to 4.0 MeV. The experimental results are extended to lower energies using measured neutron total cross sections recently reported elsewhere by the authors. The composite experimental data (total cross sections from 0.1 to 4.8 MeV and scattering cross sections from 0.22 to 4.0 MeV) are interpreted in terms of a simple two-level R-matrix model which describes the observed cross sections and implies the reaction cross section in unobserved channels; notably the (n;α)t reaction (Q = 4.783 MeV). The experimental and calculational results are compared with previously reported results as summarized in the ENDF/B-V evaluated nuclear data file

Average fast neutron flux in three energy ranges in the Quinta assembly irradiated by two types of beams

Directory of Open Access Journals (Sweden)

Strugalska-Gola Elzbieta

2017-01-01

Full Text Available This work was performed within the international project “Energy plus Transmutation of Radioactive Wastes” (E&T - RAW for investigations of energy production and transmutation of radioactive waste of the nuclear power industry. 89Y (Yttrium 89 samples were located in the Quinta assembly in order to measure an average high neutron flux density in three different energy ranges using deuteron and proton beams from Dubna accelerators. Our analysis showed that the neutron density flux for the neutron energy range 20.8 - 32.7 MeV is higher than for the neutron energy range 11.5 - 20.8 MeV both for protons with an energy of 0.66 GeV and deuterons with an energy of 2 GeV, while for deuteron beams of 4 and 6 GeV we did not observe this.

Status report and measurement of total cross-sections at the Pohang Neutron Facility

International Nuclear Information System (INIS)

Kim, G.N.; Meaze, A.K.M.M.H.; Ahmed, H.

2004-01-01

We report the status of the Pohang Neutron Facility which consists of an electron linear accelerator, a water-cooled Ta target, and an 11-m time-of-flight path. It has been equipped with a four-position sample changer controlled remotely by a CAMAC data acquisition system, which allows simultaneous accumulation of the neutron time of flight spectra from 4 different detectors. It is possible to measure the neutron total cross-sections in the neutron energy range from 0.1 eV to 100 eV by using the neutron time of flight method. A 6 LiZnS(Ag) glass scintillator was used as a neutron detector. The neutron flight path from the water-cooled Ta target to the neutron detector was 10.81±0.02 m. The background level was determined by using notch-filters of Co, In, Ta, and Cd sheets. In order to reduce the gamma rays from Bremsstrahlung and those from neutron capture, we employed a neutron-gamma separation system based on their different pulse shapes. The present measurements are in general agreement with the evaluated data in ENDF/B-VI. The resonance parameters were extracted from the transmission data from the SAMMY fitting and compared with the previous ones. (author)

Fast-neutron total and scattering cross sections of elemental palladium

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-06-01

Neutron total cross sections of palladium are measured from approx. = 0.6 to 4.5 MeV with resolutions of approx. = 30 to 70 keV at intervals of less than or equal to 50 keV. Differential neutron elastic- and inelastic-scattering cross sections are measured from 1.4 to 3.85 MeV at intervals of 50 to 100 keV and at 10 to 20 scattering angles distributed between approx. = 20 and 160 0 . The experimental results are compared with respective quantities given in ENDF/B-V and used to deduce an optical potential that provides a good description of the measured values

Fast-neutron total and scattering cross sections of elemental palladium

Energy Technology Data Exchange (ETDEWEB)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-06-01

Neutron total cross sections of palladium are measured from approx. = 0.6 to 4.5 MeV with resolutions of approx. = 30 to 70 keV at intervals of less than or equal to 50 keV. Differential neutron elastic- and inelastic-scattering cross sections are measured from 1.4 to 3.85 MeV at intervals of 50 to 100 keV and at 10 to 20 scattering angles distributed between approx. = 20 and 160/sup 0/. The experimental results are compared with respective quantities given in ENDF/B-V and used to deduce an optical potential that provides a good description of the measured values.

Fast-neutron total and scattering cross sections of 103Rh

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-07-01

Fast-neutron total cross sections of 103 Rh are measured with 30 to 50 keV resolutions from 0.7 to 4.5 MeV. Differential elastic- and inelastic-scattering cross sections are measured from 1.45 to 3.85 MeV. Scattered-neutron groups corresponding to excited levels at 334 +- 13, 536 +- 7, 648 +- 25, 796 +- 20, 864 +- 22, 1120 +- 22, 1279 +- 50, 1481 +- 27, 1683 +- 39, 1840 +- 79, 1991 +- 71 and 2050 (tentative) keV are observed. An optical-statistical model is derived from the elastic-scattering results. The experimental values are compared with comparable quantities given in the ENDF/B-V evaluation

The spin-spin effect in the total neutron cross section of polarized neutrons on polarized 165Ho

International Nuclear Information System (INIS)

Fasoli, U.; Galeazzi, G.; Pavan, P.; Toniolo, D.; Zago, G.; Zannoni, R.

1978-01-01

The spin-spin effect in the total neutron cross section of polarized neutrons on polarized 165 Ho has been measured in the energy interval 0.4 to 2.5 MeV, in perpendicular geometry. The results are consistent with zero effect. The spin-spin cross section sigmasub(ss) has been theoretically evaluated by a non-adiabatic coupled-channel calculation. From the comparison between the experimental and theoretical results a value Vsub(ss) = 9+-77 keV for the strength of the spin-spin potential has been obtained. Compound-nucleus effects do not seem to be relevant. (Auth.)

Fission Fragment Mass Distributions and Total Kinetic Energy Release of 235-Uranium and 238-Uranium in Neutron-Induced Fission at Intermediate and Fast Neutron Energies

Energy Technology Data Exchange (ETDEWEB)

Duke, Dana Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-12

This Ph.D. dissertation describes a measurement of the change in mass distributions and average total kinetic energy (TKE) release with increasing incident neutron energy for fission of ²³⁵U and ²³⁸U. Although fission was discovered over seventy-five years ago, open questions remain about the physics of the fission process. The energy of the incident neutron, En, changes the division of energy release in the resulting fission fragments, however, the details of energy partitioning remain ambiguous because the nucleus is a many-body quantum system. Creating a full theoretical model is difficult and experimental data to validate existing models are lacking. Additional fission measurements will lead to higher-quality models of the fission process, therefore improving applications such as the development of next-generation nuclear reactors and defense. This work also paves the way for precision experiments such as the Time Projection Chamber (TPC) for fission cross section measurements and the Spectrometer for Ion Determination in Fission (SPIDER) for precision mass yields.

Total body-calcium measurements: comparison of two delayed-gamma neutron activation facilities

International Nuclear Information System (INIS)

Ma, R.; Ellis, K.J.; Shypailo, R.J.; Pierson, R.N. Jr.

1999-01-01

This study compares two independently calibrated delayed-gamma neutron activation (DGNA) facilities, one at the Brookhaven National Laboratory (BNL), Upton, New York, and the other at the Children's Nutrition Research Center (CNRC), Houston, Texas that measure total body calcium (TBCa). A set of BNL phantoms was sent to CNRC for neutron activation analysis, and a set of CNRC phantoms was measured at BNL. Both facilities showed high precision (<2%), and the results were in good agreement, within 5%. (author)

The nTOF Total Absorption Calorimeter for neutron capture measurements at CERN

International Nuclear Information System (INIS)

Guerrero, C.; Abbondanno, U.; Aerts, G.; Alvarez, H.; Alvarez-Velarde, F.; Andriamonje, S.; Andrzejewski, J.; Assimakopoulos, P.; Audouin, L.; Badurek, G.; Baumann, P.; Becvar, F.; Berthoumieux, E.; Calvino, F.; Calviani, M.; Cano-Ott, D.; Capote, R.; Carrapico, C.; Cennini, P.; Chepel, V.

2009-01-01

The n T OF Collaboration has built and commissioned a high-performance detector for (n,γ) measurements called the Total Absorption Calorimeter (TAC). The TAC was especially designed for measuring neutron capture cross-sections of low-mass and/or radioactive samples with the accuracy required for nuclear technology and stellar nucleosynthesis. We present a detailed description of the TAC and discuss its overall performance in terms of energy and time resolution, background discrimination, detection efficiency and neutron sensitivity.

Synovectomy by Neutron capture; Sinovectomia por captura de neutrones

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

Fast-neutron total and scattering cross sections of niobium

Energy Technology Data Exchange (ETDEWEB)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-07-01

Neutron total cross sections of niobium were measured from approx. = 0.7 to 4.5 MeV at intervals of less than or equal to 50 keV with broad resolution. Differential-elastic-scattering cross sections were measured from approx. = 1.5 to 4.0 MeV at intervals of 0.1 to 0.2 MeV and at 10 to 20 scattering angles distributed between approx. = 20 and 160 degrees. Inelastically-scattered neutrons, corresponding to the excitation of levels at: 788 +- 23, 982 +- 17, 1088 +- 27, 1335 +- 35, 1504 +- 30, 1697 +- 19, 1971 +- 22, 2176 +- 28, 2456 +- (.), and 2581 +- (.) keV, were observed. An optical-statistical model, giving a good description of the observables, was deduced from the measured differential-elastic-scattering cross sections. The experimental-results were compared with the respective evaluated quantities given in ENDF/B-V.

Fast-neutron total and scattering cross sections of niobium

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-07-01

Neutron total cross sections of niobium were measured from approx. = 0.7 to 4.5 MeV at intervals of less than or equal to 50 keV with broad resolution. Differential-elastic-scattering cross sections were measured from approx. = 1.5 to 4.0 MeV at intervals of 0.1 to 0.2 MeV and at 10 to 20 scattering angles distributed between approx. = 20 and 160 degrees. Inelastically-scattered neutrons, corresponding to the excitation of levels at: 788 +- 23, 982 +- 17, 1088 +- 27, 1335 +- 35, 1504 +- 30, 1697 +- 19, 1971 +- 22, 2176 +- 28, 2456 +- (.), and 2581 +- (.) keV, were observed. An optical-statistical model, giving a good description of the observables, was deduced from the measured differential-elastic-scattering cross sections. The experimental-results were compared with the respective evaluated quantities given in ENDF/B-V

Microscopic description of average level spacing in even-even nuclei

International Nuclear Information System (INIS)

Huong, Le Thi Quynh; Hung, Nguyen Quang; Phuc, Le Tan

2017-01-01

A microscopic theoretical approach to the average level spacing at the neutron binding energy in even-even nuclei is proposed. The approach is derived based on the Bardeen-Cooper-Schrieffer (BCS) theory at finite temperature and projection M of the total angular momentum J , which is often used to describe the superfluid properties of hot rotating nuclei. The exact relation of the J -dependent total level density to the M -dependent state densities, based on which the average level spacing is calculated, was employed. The numerical calculations carried out for several even-even nuclei have shown that in order to reproduce the experimental average level spacing, the M -dependent pairing gaps as well as the exact relation of the J -dependent total level density formula should be simultaneously used. (paper)

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)

Neutrons characterization of the nuclear reactor Ian-R1 of Colombia; Caracterizacion de los neutrones del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez P, L. X.; Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Carretera Central del Norte Km. 1, Via Paipa, 150003 Tunja, Boyaca (Colombia); Vega C, H. R., E-mail: s.agustin.martinez@uptc.edu.co [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

Analysis of the boron pile measurement of the average neutron yield per fission of 252Cf: (AWBA development program)

International Nuclear Information System (INIS)

Ullo, J.J.

1977-08-01

The Harwell Boron Pile measurement of the average number of prompt neutrons emitted per fission, ν-bar/sub p/, of 252 Cf was analyzed in detail by a Monte Carlo method. From the calculated energy dependence of the neutron detection efficiency a value of ν-bar/sub p/ = 3.733 +- 0.022 was obtained. This value is 0.76 percent higher than the original reported value of 3.705 +- 0.015. Possible causes for this increase are discussed. 3 figures, 6 tables

Amino acids analysis by total neutron cross-sections determinations: part V

International Nuclear Information System (INIS)

Voi, Dante L.; Ferreira, Francisco de O.; Rocha, Helio F. da

2013-01-01

Total neutron cross-sections of twenty essential and non-essential amino acids to human were determined using crystal spectrometer installed on the Argonauta reactor of IEN (Instituto de Engenharia Nuclear (CNEN-RJ) and compared with data generated by parceling and grouping methodologies developed at this institution. For each amino acid was calculated the respective neutron cross-section by molecular structure, conformation and chemistry analysis. The results obtained for eighteen of twenty amino acids confirm the specifications and product formulations indicated by manufactures. These initial results allow to build a neutron cross-sections database as part of quality control of the amino supplied to hospitals for production of nutriments for parenteral or enteral formulations used in critical patients dependent on artificial feed, and for application in future studies of structure and dynamics for more complex molecules, including proteins, enzymes, fatty acids, membranes, organelles and other cell components. (author)

Neutron detection technique

International Nuclear Information System (INIS)

Oblath, N.S.; Poon, A.W.P.

2000-01-01

The Sudbury Neutrino Observatory (SNO) has the ability to measure the total flux of all active flavors of neutrinos using the neutral current reaction, whose signature is a neutron. By comparing the rates of the neutral current reaction to the charged current reaction, which only detects electron neutrinos, one can test the neutrino oscillation hypothesis independent of solar models. It is necessary to understand the neutron detection efficiency of the detector to make use of the neutral current reaction. This report demonstrates a coincidence technique to identify neutrons emitted from the 252 Cf neutron calibration source. The source releases on average four neutrons when a 252 Cf nucleus spontaneously fissions. Each neutron is detected as a separate event when the neutron is captured by a deuteron, releasing a gamma ray of approximately 6.25 MeV. This gamma ray is in turn detected by the photomultiplier tube (PMT) array. By investigating the time and spatial separation between neutron-like events, it is possible to obtain a pure sample of neutrons for calibration study. Preliminary results of the technique applied to two calibration runs are presented

Experimental and theoretical total neutron scattering cross-section of water confined in silica microspheres

Energy Technology Data Exchange (ETDEWEB)

Muhrer, G., E-mail: muhrer@lanl.gov [Los Alamos National Laboratory, Los Alamos, 87545 NM (United States); Hartl, M.; Mocko, M.; Tovesson, F.; Daemen, L. [Los Alamos National Laboratory, Los Alamos, 87545 NM (United States)

2012-07-21

In the search for moderator materials encapsulated materials have been discussed, but very little is known regarding the effect of encapsulation on neutron moderation properties. As a first step toward a better understanding, we present the measured total neutron cross-section of water confined in silica microspheres and compare the measured data to the predicted theoretical cross-section.

Measurements of the total neutron cross-sections of U and UO2 below 2 eV at different temperatures

International Nuclear Information System (INIS)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Ashry, A.; Abbas, Y.; Abu-Zahra, A.; Hamouda, I.

1982-11-01

The total neutron cross-sections of natural uranium and its oxide are measured using two time of flight spectrometers, installed in front of two of the ET-RR-1 reactor horizontal channels, and also by a neutron diffraction spectrometer. The measurements were carried out at room temperature in the energy range from 2 eV-0.002 eV and at 210 deg. C, for neutron energies below 0.005 eV. The coherent scattering cross-section of U was deduced both from the Bragg cut-offs observed in the behaviour of the total neutron cross-section of both U and UO 2 at cold neutron energies and the neutron diffraction pattern obtained at room temperature. (author)

Measurement of the polarized neutron---polarized 3He total cross section

International Nuclear Information System (INIS)

Keith, C.D.; Gould, C.R.; Haase, D.G.; Seely, M.L.; Huffman, P.R.; Roberson, N.R.; Tornow, W.; Wilburn, W.S.

1995-01-01

The first measurements of polarized neutron--polarized 3 He scattering in the few MeV energy region are reported. The total cross section difference Δσ T for transversely polarized target and beam has been measured for neutron energies between 1.9 and 7.5 MeV. Comparison is made to predictions of Δσ T using various descriptions of the 4 He continuum. A brute-force polarized target of solid 3 He has been developed for these measurements. The target is 4.3x10 22 atoms/cm 2 thick and is polarized to 38% at 7 Telsa and 12 mK. copyright 1995 American Institute of Physics

In vivo measurement of total body carbon using 238Pu/Be neutron sources

International Nuclear Information System (INIS)

Sutcliffe, J.F.; Mitra, S.; Hill, G.L.

1990-01-01

Total body carbon has been measured by in vivo neutron activation analysis (IVNAA) in 278 surgical gastroenterological patients and 29 normal volunteers. This is based on the inelastic scattering reaction { 12 C(n,n') 12 C*} for neutrons with energy above 4.8MeV, producing 4.43 MeV gamma rays. Since only part of the body is scanned, total body carbon is estimated as the ratio of the gamma ray emission from carbon to the emission from hydrogen, using hydrogen as the internal standard. The precision of the estimate is ±1.6kg for a whole body dose of 0.3mSv. There is a significant difference between the estimates of total body water from IVNAA measurements of carbon and nitrogen and measurements of body water in these subjects by tritium dilution (t=3.1, p < 0.005). (author)

A recent investigation of neutron total cross section of zirconium in the wavelength range (0.1-1.25) Ao

International Nuclear Information System (INIS)

Abu El-Ela, M.A.

1996-01-01

The neutron total cross section of zirconium has been investigated in the neutron wavelength range (0.1 -1.52) A o by using slow neutron time of flight spectrometer, installed in front of the horizontal channel No.6 of the ETRR-1 reactor (2MW). The results have showed that the neutrons with short wavelength (0.1 - 0.76) A o cannot interact with the crystal structure while it can interact with the free bound atom to give the value (6.2 +0.1) barns for the potential scattering cross section or (the scattering length = 6.2 fermi)). The present measured value is in good agreement with the international published values by different technique. The neutrons with longer wavelength (0.76 - 1.52) A o have showed dependence of the total cross section on the neutron wavelength. Such dependence between the total cross section and the neutron wavelength can not be observed in the reported previous measurements, which can be attributed to the limited number of the measured values. 4 figs

Measurement of total body chlorine by prompt gamma in vivo neutron activation analysis

International Nuclear Information System (INIS)

Beddoe, A.H.; Streat, S.J.; Hill, G.L.

1987-01-01

A method of measuring total body chlorine (TBCl) by prompt gamma in vivo neutron activation analysis is described depending on the same NaI(Tl) spectra used for determinations of total body nitrogen. Ratios of chlorine to hydrogen are derived and TBCl determined using a model of body composition depending on measured body weight, total body water (by tritium dilution) and protein (6.25 x nitrogen) as well as estimated body minerals and glycogen. The precision of the method based on scanning an anthropomorphic phantom is approximately 9% (SD), for a patient dose equivalent of less than 0.30 mSv. Spectra collected from 67 normal volunteers (32 male, 35 female) yielded mean values of TBCl of 72 +- 19 (SD) g in males and 53.6 +- 15 g in females, in broad agreement with values reported by workers using delayed gamma methods. Results are presented for two human cadavers analysed by neutron activation and conventional chemical analysis; the ratios of TBCl (neutron activation) to TBCl (chemical) were 0.980 +- 0.028 (SEM) and 0.91 +- 0.09. It is suggested that an improvement in precision will be achieved by increasing the scanning time (thereby increasing the radiation dose equivalent) and by adding two more detectors. (author)

Neutron cross section measurement using the Oak Ridge Electron Linear Accelerator

International Nuclear Information System (INIS)

Winters, R.R.

1991-08-01

This report discusses: argon-40 -- neutron reaction total cross sections from 6.9 kev to 50 kev; The maxwellian averaged neutron capture cross section of oxygen-16; r-matrix parameter analysis of the lead-208 -- neutron reaction cross section measurement; r-matrix parameter analysis of the ORELA neutron transmission zirconium-90 low energy measurement; porting computer codes from the HP9000 to the IBM RISC/6000;and measurements of neutron reactions with strontium-88, zirconium-90, and calcium-40

Neutrons characterization of the nuclear reactor Ian-R1 of Colombia

International Nuclear Information System (INIS)

Gonzalez P, L. X.; Martinez O, S. A.; Vega C, H. R.

2014-08-01

By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

Some implications of batch average burnup calculations on predicted spent fuel compositions

International Nuclear Information System (INIS)

Alexander, C.W.; Croff, A.G.

1984-01-01

The accuracy of using batch-averaged burnups to determine spent fuel characteristics (such as isotopic composition, activity, etc.) was examined for a typical pressurized-water reactor (PWR) fuel discharge batch by comparing characteristics computed by (a) performing a single depletion calculation using the average burnup of the spent fuel and (b) performing separate depletion calculations based on the relative amounts of spent fuel in each of twelve burnup ranges and summing the results. The computations were done using ORIGEN 2. Procedure (b) showed a significant shift toward a greater quantity of the heavier transuranics, which derive from multiple neutron captures, and a corresponding decrease in the amounts of lower transuranics. Those characteristics which derive primarily from fission products, such as total radioactivity and total thermal power, are essentially identical for the two procedures. Those characteristics that derive primarily from the heavier transuranics, such as spontaneous fission neutrons, are underestimated by procedure (a)

Neutron capture studies of {sup 206}Pb at a cold neutron beam

Energy Technology Data Exchange (ETDEWEB)

Schillebeeckx, P.; Kopecky, S.; Quetel, C.R.; Tresl, I.; Wynants, R. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); Belgya, T.; Szentmiklosi, L. [Institute for Energy Security and Environmental Safety, Centre for Energy Research, Budapest (Hungary); Borella, A. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); SCK CEN, Mol (Belgium); Mengoni, A. [Nuclear Data Section, International Atomic Energy Agency (IAEA), Wagramerstrasse 5, PO Box 100, Vienna (Austria); Agenzia Nazionale per le Nuove Tecnologie, l' Energia e lo Sviluppo Economico Sostenibile (ENEA), Bologna (Italy)

2013-11-15

Gamma-ray transitions following neutron capture in {sup 206}Pb have been studied at the cold neutron beam facility of the Budapest Neutron Centre using a metallic sample enriched in {sup 206}Pb and a natural lead nitrate powder pellet. The measurements were performed using a coaxial HPGe detector with Compton suppression. The observed {gamma} -rays have been incorporated into a decay scheme for neutron capture in {sup 206}Pb. Partial capture cross sections for {sup 206}Pb(n, {gamma}) at thermal energy have been derived relative to the cross section for the 1884 keV transition after neutron capture in {sup 14}N. From the average crossing sum a total thermal neutron capture cross section of 29{sup +2}{sub -1} mb was derived for the {sup 206}Pb(n, {gamma}) reaction. The thermal neutron capture cross section for {sup 206}Pb has been compared with contributions due to both direct capture and distant unbound s-wave resonances. From the same measurements a thermal neutron-induced capture cross section of (649 {+-} 14) mb was determined for the {sup 207}Pb(n, {gamma}) reaction. (orig.)

Evaluation of the total gamma-ray production cross-sections for nonelastic interaction of fast neutrons with iron nuclei

International Nuclear Information System (INIS)

Savin, M.V.; Nefedov, Yu.Ya; Livke, A.V.; Zvenigorodskij, A.G.

2001-01-01

Experimental data on the total gamma-ray production cross-sections for inelastic interaction of fast neutrons with iron nuclei were analysed. The total gamma-ray production cross-sections, grouped according to E γ , were evaluated in the neutron energy range 0.5-19 MeV. The statistical spline approximation method was used to evaluate the experimental data. Evaluated data stored in the ENDF, JENDL, BROND, and other libraries on gamma-ray production spectra and cross-sections for inelastic interaction of fast neutrons with iron nuclei, were analysed. (author)

Simultaneous investigation of fission fragments and neutrons in 252Cf(s,f)

International Nuclear Information System (INIS)

Budtz-Joergensen, C.; Knitter, H.H.

1986-01-01

The gridded twin ion chamber developed at CBNM is used to measure the kinetic energy-, mass- and angular distributions of the fission fragments of 252 Cf in an advantageous 4π-geometry. Together with a neutron time-of-flight detector this experimental arrangement permits to measure the correlation between neutron emission, fragment angle, mass and energy in the spontaneous fission of 252 Cf. With the present experimental set-up a mass resolution for fission fragments of 0.5 a.m.u., an angular resolution of Δcosθ = 0.05 and a timing resolution of 0.7 ns FWHM were observed. Preliminary evaluations of the raw experimental data are presented for the fission fragment mass distribution, the average total kinetic energy and their variance as function of mass, the angular distribution between fragments and neutrons, the number of neutrons emitted per fragment as function of fragment mass, the average neutron emission energies as function of mass, and the prompt fission neutron spectrum averaged over all fragments. (author)

Study of the fluctuations of the partial and total radiative widths by neutron capture resonance method; Etude des fluctuations des largeurs radiatives partielles et totales par la capture des neutrons de resonance

Energy Technology Data Exchange (ETDEWEB)

Huynh, V D [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-06-01

Radiative capture experiments by neutron time-of-flight methods have been made for following studies: distribution of partial radiative widths, effects of correlation between different radiative transitions, fluctuations of total radiative widths {gamma}{sub {gamma}} from resonance to resonance, variation of {gamma}{sub {gamma}} with number of mass and the search for the existence of potential capture. Also, some other experiments with the use of neutron capture gamma-rays spectra have been investigated. (author) [French] Par la capture des neutrons de resonance dont les energies sont selectionnees a l'aide de la technique du temps de vol, differents types d'experiences ont ete realisees concernant les etudes des distributions des largeurs radiatives partielles, des effets de correlation entre differentes voies de desexcitation, de la fluctuation des largeurs radiatives totales {gamma}{sub {gamma}} de resonance a resonance, de la variation de la quantite {gamma}{sub {gamma}} en fonction du nombre de masse et de la mise en evidence de l'existence du processus de capture potentielle. Quelques autres applications de l'emploi du spectre de rayons gamma ont egalement ete presentees. (auteur)

FPGA based computation of average neutron flux and e-folding period for start-up range of reactors

International Nuclear Information System (INIS)

Ram, Rajit; Borkar, S.P.; Dixit, M.Y.; Das, Debashis

2013-01-01

Pulse processing instrumentation channels used for reactor applications, play a vital role to ensure nuclear safety in startup range of reactor operation and also during fuel loading and first approach to criticality. These channels are intended for continuous run time computation of equivalent reactor core neutron flux and e-folding period. This paper focuses only the computational part of these instrumentation channels which is implemented in single FPGA using 32-bit floating point arithmetic engine. The computations of average count rate, log of average count rate, log rate and reactor period are done in VHDL using digital circuit realization approach. The computation of average count rate is done using fully adaptive window size moving average method, while Taylor series expansion for logarithms is implemented in FPGA to compute log of count rate, log rate and reactor e-folding period. This paper describes the block diagrams of digital logic realization in FPGA and advantage of fully adaptive window size moving average technique over conventional fixed size moving average technique for pulse processing of reactor instrumentations. (author)

Proposal of a wide-band mirror polarizer of slow neutrons at a pulsed neutron source

International Nuclear Information System (INIS)

Nikitenko, Yu.V.; Ostanevich, Yu.M.

1993-01-01

The new type of wide-band mirror-based neutron polarizer, which is to be operated at a pulsed neutron source, is suggested. The idea is to use a movable polarizing mirror system, which, with the incoming beam monochromatized by the time-of-flight, would allow one to tune glancing angles in time so that the total reflection condition is always fulfilled only for one of the two neutron spin eigenstates. Estimates show that with the pulsed reactor IBR-2 such a polarizer allows one to build a small angle neutron scattering instrument capable of effectively using the wavelength band from 2 A with a rather high luminosity (time-averaged flux at sample position being up to 10 7 n/s/cm -2 ). (orig.)

Synovectomy by Neutron capture

International Nuclear Information System (INIS)

Vega C, H.R.; Torres M, C.

1998-01-01

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

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)

Reevaluation of the average prompt neutron emission multiplicity (nubar) values from fission of uranium and transuranium nuclides

International Nuclear Information System (INIS)

Holden, N.E.; Zucker, M.S.

1984-01-01

In response to a need of the safeguards community, we have begun an evaluation effort to upgrade the recommended values of the prompt neutron emission multiplicity distribution, P/sub nu/ and its average value, nubar. This paper will report on progress achieved thus far. The evaluation of the uranium, plutonium, americium and curium nuclide's nubar values will be presented. The recommended values will be given and discussed. 61 references

Neutron-nucleus interactions and fission. Chapter 1

International Nuclear Information System (INIS)

1998-01-01

The central problem in nuclear-reactor kinetics is to predict the evolution in time of the neutron population in a multiplying medium. Point kinetics allows study of the global behaviour of the neutron population from the average properties of the medium. Before tackling, in the following chapters, the equations governing the time variation of the reactor power (proportional to the total neutron population), the properties of a neutron-multiplying medium shall be discussed briefly. After recalling a number of definitions, a qualitative description shall be given of the principal nuclear reactions at play in a self-sustaining chain reaction, with emphasis on the source of fission neutrons. Since delayed neutrons play a crucial role in reactor kinetics, their production in a reactor shall be described in greater detail. (author)

Multiplicity and energy of neutrons from {sup 233}U(n{sub th},f) fission fragments

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

The correlation between fission fragments and prompt neutrons from the reaction {sup 233}U(n{sub th},f) was measured with improved accuracy. The results determined the neutron multiplicity and emission energy as a function of fragment mass and total kinetic energy. The average energy as a function of fragment mass followed a nearly symmetric distribution centered about the equal mass-split and formed a remarkable contrast with the saw-tooth distribution of the average neutron multiplicity. The neutron multiplicity from the specified fragment decreases linearly with total kinetic energy, and the slope of multiplicity with kinetic energy had the minimum value at about 130 u. The level density parameter versus mass determined from the neutron data showed a saw-tooth structure with the pronounced minimum at about 128 and generally followed the formula by Gilbert and Cameron, suggesting that the neutron emission process was very much affected by the shell-effect of the fission fragment. (author)

Quantitation of the degree of osteoporosis by measure of total-body calcium employing neutron activation

International Nuclear Information System (INIS)

Cohn, S.H.; Zanzi, I.; Vaswani, A.; Wallach, S.; Aloia, J.; Ellis, K.J.

1975-01-01

Two techniques for measuring the amount of Ca in the total skeleton were employed: total-body neutron activation analysis (TBNAA) and the determination of the mineral content of a bone of the appendicular skeleton (absorptiometric measurement of the radius, BMC). (U.S.)

Study of the fluctuations of the partial and total radiative widths by neutron capture resonance method; Etude des fluctuations des largeurs radiatives partielles et totales par la capture des neutrons de resonance

Energy Technology Data Exchange (ETDEWEB)

Huynh, V.D. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-06-01

Radiative capture experiments by neutron time-of-flight methods have been made for following studies: distribution of partial radiative widths, effects of correlation between different radiative transitions, fluctuations of total radiative widths {gamma}{sub {gamma}} from resonance to resonance, variation of {gamma}{sub {gamma}} with number of mass and the search for the existence of potential capture. Also, some other experiments with the use of neutron capture gamma-rays spectra have been investigated. (author) [French] Par la capture des neutrons de resonance dont les energies sont selectionnees a l'aide de la technique du temps de vol, differents types d'experiences ont ete realisees concernant les etudes des distributions des largeurs radiatives partielles, des effets de correlation entre differentes voies de desexcitation, de la fluctuation des largeurs radiatives totales {gamma}{sub {gamma}} de resonance a resonance, de la variation de la quantite {gamma}{sub {gamma}} en fonction du nombre de masse et de la mise en evidence de l'existence du processus de capture potentielle. Quelques autres applications de l'emploi du spectre de rayons gamma ont egalement ete presentees. (auteur)

Fast-neutron total and scattering cross sections of sup 58 Ni and nuclear models

Energy Technology Data Exchange (ETDEWEB)

Smith, A.B.; Guenther, P.T.; Whalen, J.F. (Argonne National Lab., IL (United States)); Chiba, S. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

1991-07-01

The neutron total cross sections of {sup 58}Ni were measured from {approx} 1 to > 10 MeV using white-source techniques. Differential neutron elastic-scattering cross sections were measured from {approx} 4.5 to 10 MeV at {approx} 0.5 MeV intervals with {ge} 75 differential values per distribution. Differential neutron inelastic-scattering cross sections were measured, corresponding to fourteen levels with excitations up to 4.8 MeV. The measured results, combined with relevant values available in the literature, were interpreted in terms of optical-statistical and coupled-channels model using both vibrational and rotational coupling schemes. The physical implications of the experimental results nd their interpretation are discussed in the contexts of optical-statistical, dispersive-optical, and coupled-channels models. 61 refs.

Total reaction cross sections and neutron-removal cross sections of neutron-rich light nuclei measured by the COMBAS fragment-separator

Science.gov (United States)

Hue, B. M.; Isataev, T.; Erdemchimeg, B.; Artukh, A. G.; Aznabaev, D.; Davaa, S.; Klygin, S. A.; Kononenko, G. A.; Khuukhenkhuu, G.; Kuterbekov, K.; Lukyanov, S. M.; Mikhailova, T. I.; Maslov, V. A.; Mendibaev, K.; Sereda, Yu M.; Penionzhkevich, Yu E.; Vorontsov, A. N.

2017-12-01

Preliminary results of measurements of the total reaction cross sections σR and neutron removal cross section σ-xn for weakly bound 6He, 8Li, 9Be and 10Be nuclei at energy range (20-35) A MeV with 28Si target is presented. The secondary beams of light nuclei were produced by bombardment of the 22Ne (35 A MeV) primary beam on Be target and separated by COMBAS fragment-separator. In dispersive focal plane a horizontal slit defined the momentum acceptance as 1% and a wedge degrader of 200 μm Al was installed. The Bρ of the second section of the fragment-separator was adjusted for measurements in energy range (20-35) A MeV. Two-neutron removal cross sections for 6He and 10Be and one -neutron removal cross sections 8Li and 9Be were measured.

Total cross section measurement of radioactive isotopes with a thin beam neutron spectrometer

International Nuclear Information System (INIS)

Razbudej, V.F.; Vertebnyj, V.P.; Padun, G.S.; Muravitskij, A.V.

1975-01-01

The method for measuring the neutron total cross sections of radioactive isotopes by a time-of-flight spectrometer with a narrow (0.17 mm in diameter) beam of thermal neutrons is described. The distinguishing feature of this method is the use of capillary samples with a small amount of substance (0.05-1.0 mg). The energy range is 0.01-0.3 eV. The total cross sections of irradiated samples of sub(153)Eu and sub(151)Eu are measured. From them are obtained the cross sections of sub(152)Eu (Tsub(1/2)=12.4 g) and of sub(154)E (Tsub(1/2)=8.6 yr); they equal 11400+-1400 and 1530+-190 barn at E=0.0253 eV. The cross section of the sub(152)Eu absorption for the thermal spectrum (T=333 K) is determined by the activation method; it is 8900+-1200 barn

Evaluation of neutron nuclear data for 249Cf

International Nuclear Information System (INIS)

Zhou Delin; Yu Baosheng; Yuan Hanrong; Liu Tong; Zhang Jin; Su Zongdi; Yan Shiwei; Wang Cuilan; Zhang Jingshang

1992-01-01

A complete set of neutron nuclear data from 10 -5 eV to 20 MeV in ENDF/B-6 format has been evaluated for 249 Cf based on measured data, systematics predications and model theory calculations. The evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) reaction cross sections, the resolved and unresolved resonance parameters, the angular and energy distributions of emitted neutrons and the average number of neutrons emitted per fission. The numerical data are available in ENDF/B-6 format

Neutron multiplicity of fission fragments

Energy Technology Data Exchange (ETDEWEB)

Abdelrahman, Y S [Physics department, mu` rah university Al-Karak, (Jordan)

1995-10-01

The total average neutron multiplicity of the fission fragments produced by the spontaneous fission of {sup 248} Cm has been measured. This measurement has been done by using a new experimental technique. This technique mainly depends on {gamma}-{gamma} coincidence using a very high resolution high purity germanium (HPGe) detector. 2 figs.

Ultracold neutrons

International Nuclear Information System (INIS)

Steenstrup, S.

Briefly surveys recent developments in research work with ultracold neutrons (neutrons of very low velocity, up to 10 m/s at up to 10 -7 eV and 10 -3 K). Slow neutrons can be detected in an ionisation chamber filled with B 10 F 3 . Very slow neutrons can be used for investigations into the dipole moment of neutrons. Neutrons of large wave length have properties similar to those of light. The limit angle for total reflection is governed by the wave length and by the material. Total reflection can be used to filter ultracold neutrons out of the moderator material of a reactor. Total reflection can also be used to store ultracold neutrons but certain problems with storage have not yet been clarified. Slow neutrons can be made to lose speed in a neutron turbine, and come out as ultracold neutrons. A beam of ultracold neutrons could be used in a neutron microscope. (J.S.)

The neutron total cross-section measurement of 56Fe and 57Fe by using Japan Proton Accelerator Research Complex facility

International Nuclear Information System (INIS)

Kim, Eun Ae; Shvetsov, Valery; Cho, Moo Hyun; Won, Nam Kung; Kim, Kwang Soo; Yang, Sung Chul; Lee, Man Woo; Kim, Guin Yun; Yi, Kyoung Rak; Choi, Hong Yub; Ro, Tae Ik; Mizumoto, Motoharu; Katabuchi, Tatsuya; Igashira, Masayuki

2012-01-01

The measurement of neutron cross section using Time-Of-Flight (TOF) method gives significant information for the nuclear data research. In the present work, the neutron total cross section of 56 Fe and 57 Fe has been measured in the energy range between 10 eV and 100 keV by using the neutron beam produced from 3-GeV proton synchrotron accelerator. The 3-GeV proton synchrotron accelerator is located at Japan Proton Accelerator Research Complex (J-PARC) facility in Tokai village. In this study, the neutron total cross section data measured by 6 Li glass scintillator detector was compared with the evaluated values of ENDF/B-VII.0

Contribution to the study of the unresolved resonance range of the neutrons cross sections

International Nuclear Information System (INIS)

Noguere, Gilles

2014-01-01

This document presents the statistical description of neutron cross sections in the unresolved resonance range. The modeling of the total cross section and of the 'shape - elastic' cross section is based on the 'average R-Matrix' formalism. The partial cross sections describing the radiative capture, elastic scattering, inelastic scattering and fission process are calculated using the Hauser-Feshbach formalism with width fluctuation corrections. In the unresolved resonance range, these models depend on the average resonance parameters (neutron strength function Sc, mean level spacing D c , average partial reaction widths Γ c , channel radius a c , effective radius R' and distant level parameter R-bar c ∞ ). The codes (NJOY, CALENDF...) dedicated to the processing of nuclear data libraries (JEFF, ENDF/B, JENDL, CENDL, BROND... ) use the average parameters to take into account the self-shielding phenomenon for the simulation of the neutron transport in Monte-Carlo (MCNP, TRIPOLI... ) and deterministic (APOLLO, ERANOS...) codes. The evaluation work consists in establishing a consistent set of average parameters as a function of the total angular momentum J of the system and of the orbital moment of the incident neutron l. The work presented in this paper aims to describe the links between the S-Matrix and the 'average R-Matrix' formalism for the calculation of Sc, R-bar c ∞ , ac and R'. (author) [fr

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

Progress on FP13 Total Cross Section Measurements Capability

Energy Technology Data Exchange (ETDEWEB)

Ullmann, John Leonard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Couture, Aaron Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Koehler, Paul E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mocko, Michal [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wender, Stephen Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-26

An accurate knowledge of the neutron capture cross section is important for many applications. Experimental measurements are important since theoretical calculations of capture have been notoriously difficult, with the ratio of measured to calculated cross sections often a factor of 2 or more in the 10 keV to 1 MeV region. However, a direct measurement of capture cannot be made on many interesting radioactive nuclides because of their short half-life or backgrounds caused by their nuclear decay. On the other hand, neutron transmission measurements of the total cross section are feasible for a wide range of radioactive nuclides since the detectors are far from the sample, and often are less sensitive to decay radiation. The parameters extracted from a total cross section measurement, which include the average resonance spacing, the neutron strength function, and the average total radiation width, (Γ_γ), provide tight constraints on the calculation of the capture cross section, and when applied produce much more accurate results. These measurements can be made using the intense epithermal neutron flux at the Lujan Center on relatively small quantities of target material. It was the purpose of this project to investigate and develop the capability to make these measurements. A great deal of progress was made towards establishing this capability during 2016, including setting up the flight path and obtaining preliminary results, but more work remains to be done.

The EUROBALL neutron wall - design and performance tests of neutron detectors

CERN Document Server

Skeppstedt, Ö; Lindström, L; Wadsworth, R; Hibbert, I; Kelsall, N; Jenkins, D; Grawe, H; aGórska, M; Moszynski, M; Sujkowski, Z; Wolski, D; Kapusta, M; Hellström, M; Kalogeropoulos, S; Oner, D; Johnson, A; Cederkäll, J; Klamra, W; Nyberg, J; Weiszflog, M; Kay, J; Griffiths, R; Garces-Narro, J; Pearson, C; Eberth, J

1999-01-01

The mechanical design of the EUROBALL neutron wall and neutron detectors, and their performance measured with a sup 2 sup 4 sup 6 sup , sup 2 sup 4 sup 8 Cm fission source are described. The array consists of 15 pseudohexaconical detector units subdivided into three, 149 mm high, hermetically separated segments and a smaller central pentagonal unit subdivided into five segments. The detectors are filled with Bicron BC501A liquid scintillator. Each section of the hexaconical detectors is viewed by a 130 mm diameter Philips XP4512PA photomultiplier while the sections of pentagonal detectors are viewed by Philips XP4312B PMTs. The tests of n-gamma discrimination performed by zero-crossing and time-of-flight methods show a full separation of gamma- and neutron events down to 50 keV recoil electron energy. These tests demonstrate the excellent timing properties of the detectors and an average time resolution of 1.56 ns. The factors determining the efficiency of neutron detectors are discussed. The total efficiency...

Measurements of Neutron Induced Cross Sections at the Oak Ridge Electron Linear Accelerator

International Nuclear Information System (INIS)

Guber, K.H.; Harvey, J.A.; Hill, N.W.; Koehler, P.E.; Leal, L.C.; Sayer, R.O.; Spencer, R.R.

1999-01-01

We have used the Oak Ridge Electron Linear Accelerator (ORELA) to measure neutron total and the fission cross sections of 233 U in the energy range from 0.36 eV to 700 keV. We report average fission and total cross sections. Also, we measured the neutron total cross sections of 27 Al and Natural chlorine as well as the capture cross section of Al over an energy range from 100 eV up to about 400 keV

Preliminary results of total kinetic energy modelling for neutron-induced fission

International Nuclear Information System (INIS)

Visan, I.; Giubega, G.; Tudora, A.

2015-01-01

The total kinetic energy as a function of fission fragments mass TKE(A) is an important quantity entering in prompt emission calculations. The experimentally distributions of TKE(A) are referring to a limited number of fission systems and incident energies. In the present paper, a preliminary model for TKE calculation in neutron induced fission system is presented. The range of fission fragments is chosen as in the Point by Point treatment. The model needs as input only mass excesses and deformation parameters taken from available nuclear databases being based on the following approximations: total excitation energy of fully accelerated fission fragments TXE is calculated from energy balance of neutron-induced fission systems as sum of the total excitation energy at scission E*sciss and deformation energy Edef. The deformation energy at scission is given by minimizing the potential energy at the scission configuration. At the scission point, the fission system is described by two spheroidal fragments nearly touching by a pre-scission distance or neck caused by the nuclear forces between fragments. Therefore, the Columbian repulsion depending on neck and, consequently, on the fragments deformation at scission, is essentially in TKE determination. An approximation is made based on the fission modes. For the very symmetric fission, the dominant super long channel is characterized by long distance between fragments leading to low TKE values. Due to magic and double-magic shells closure, the dominant S1 fission mode for pairs with heavy fragment mass AH around 130-134 is characterized by spherical heavy fragment shape and easily deformed light fragment. The nearly spherical shape of the complementary fragments are characterized by minimum distance, and consequently to maximum TKE values. The results obtained for TKE(A) are in good agreement with existing experimental data for many neutron induced fission systems, e.g. ''2''3''3&apos

Neutron total, scattering and inelastic gamma-ray cross sections of yttrium at few MeV energies

International Nuclear Information System (INIS)

Budtz-Joergensen, C.; Guenther, P.; Smith, A.; Whalen, J.; McMurray, W.R.; Renan, M.J.; Heerden, I.J. van

1984-01-01

Neutron total, scattering and (n; n', γ) cross sections of elemental yttrium ( 89 Y) were measured in the few-MeV region. The neutron total-cross-section measurements were made with broad resolutions from approx.=0.5 to 4.2 MeV in steps of < or approx.0.1 MeV. Neutron elastic- and inelastic-scattering cross sections were measured from approx.=1.5 to 4.0 MeV, at incident-neutron energy intervals of approx.=50 keV and at ten or more scattering angles distributed between 20 and 160 degrees using neutron detection. Inelastic-scattering cross sections were also determined using the (n; n', γ) reaction at incident energies from 1.6 to 3.8 MeV at intervals of 0.1 MeV. Gamma-rays and/or inelastically-scattered neutrons were observed corresponding to the excitation of levels at: 909.0+-0.5, 1,507.4+-0.3, 1,744.5+-0.3, 2,222.6+-0.5, 2,530+-0.8, 2,566.4+-1.0, 2,622.5+-1.0, 2,871.9+-1.5, 2,880.6+-2.0, 3,067.0+-2.0, 3,107.0+-2.0, 3,140.0+-2.0, 3,410.0+-2.0, 3,450.0+-2.0, 3,504.0+-1.5, 3,514.0+-2.0, 3,556.0+-2.0, 3,619.0+-3.0, 3,629.0+-3.0 and 3,715.0+-3.0 keV. The experimental results are discussed in terms of the spherical-optical-statistical, coupled-channels, and core-coupling models, and in the context of previously reported excited-level structure. (orig.)

Fast-neutron total and elastic-scattering cross sections of elemental indium

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-11-01

Broad-resolution neutron total cross sections of elemental indium were measured from 0.8 to 4.5 MeV. Differential-elastic-scattering cross sections were measured from approx. = 1.5 to 3.8 MeV at intervals of approx. = 50 to 200 keV and at scattering angles in the range 20 to 160 degrees. The experimental results are interpreted in terms of the optical-statistical model and are compared with respective values given in ENDF/B-V

Recent performance of the Intense Pulsed Neutron Source accelerator system

International Nuclear Information System (INIS)

Potts, C.; Brumwell, F.; Rauchas, A.; Stipp, V.; Volk, G.; Donley, L.

1987-03-01

The Intense Pulsed Neutron Source (IPNS) accelerator system has now been in operation as part of a national user program for over five years. During that period steady progress has been made in both beam intensity and reliability. Almost 1.8 billion pulses totaling 4 x 10 21 protons have now been delivered to the spallation neutron target. Recent weekly average currents have reached 15 μA (3.2 x 10 12 protons per pulse, 30 pulses per second) and short-term peaks of almost 17 μA have been reached. In fact, the average current for the last two years is up 31% over the average for the first three years of operation

The use of averages and other summation quantities in the testing of evaluated fission product yield and decay data. Applications to ENDF/B(IV)

International Nuclear Information System (INIS)

Walker, W.H.

1976-01-01

Averages of some fission product properties can be obtained by multiplying the fission product yield for each fission product by the value of the property (e.g. mass, atomic number, mass defect) for that fission product and summing all significant contributions. These averages can be used to test the reliability of the yield set or provide useful data for reactor calculations. The report gives the derivation of these averages and discusses their application using the ENDF/B(IV) fission product library. The following quantities are treated here: the number of fission products per fission ΣYsub(i); the average mass number and the average number of neutrons per fission; the average atomic number of the stable fission products and the average number of β-decays per fission; the average mass defect of the stable fission products and the total energy release per fission; the average decay energy per fission (beta, gamma and anti-neutrino); the average β-decay energy per fission; individual and group-averaged delayed neutron emission; the total yield for each fission product element. Wherever it is meaningful to do so, a sum is subdivided into its light and heavy mass components. The most significant differences between calculated values based on ENDF/B(IV) and measurements are the β and γ decay energies for 235 U thermal fission and delayed neutron yields for other fissile nuclides, most notably 238 U. (author)

The total neutron cross-sections of 151Eu, 153Eu and Eu below 1 eV

International Nuclear Information System (INIS)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Ashry, A.; Hamouda, I.

1981-01-01

Total neutron cross-section measurements have been carried out for natural Eu and its stable isotopes in the energy range from 3 meV to 1 eV. The measurements were performed using two time-of-flight spectrometers installed in front of two of the horizontal channels of the ET-RR-1 reactor. The following results have been obtained: sigmasub(a) ( 151 Eu) = (9180 +- 150) b at 0.0253 eV, sigmasub(s) ( 153 Eu) = (375 +- 20) b at 0.0253 eV, sigmasub(d) (Eu) = (4600 +- 120) b at 0.0253 eV. The contribution of the resonance to the total neutron cross-sections, in the thermal region, was calculated using the single-level Breit-Wigner formula. (orig.)

Nuclear data for neutron emission in the fission process

International Nuclear Information System (INIS)

Ganesan, S.

1991-11-01

This document contains the proceedings of the IAEA Consultants' Meeting on Nuclear Data for Neutron Emission in the Fission Process, Vienna, 22 - 24 October 1990. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers provide a review of the status of experimental and theoretical data on neutron emission in spontaneous and neutron induced fission with reference to the data needs for reactor applications oriented towards actinide burner studies. The specific topics covered are the following: experimental measurements and theoretical predictions and evaluations of fission neutron energy spectra, average prompt fission neutron multiplicity, correlation in neutron emission from complementary fragments, neutron emission during acceleration of fission fragments, statistical properties of neutron rich nuclei by study of emission spectra of neutrons from the excited fission fragments, integral qualification of nu-bar for the major fissile isotopes, nu-bar total of 239 Pu and 235 U, and related problems. Refs figs and tabs

P-wave assignment of 232Th neutron resonances

International Nuclear Information System (INIS)

Corvi, F.; Pasquariello, G.; Veen, T. van der

1978-01-01

A method of p-wave assignment which exploits the parity dependence of the primary capture γ-ray spectrum was applied to the 232 Th resonance. The yield of capture γ-rays above 4.4 MeV from a 6 mm thick metallic thorium disk was measured in the neutron energy range 20-2200 eV and compared to a similar run with γ-rays in the range 3.7 - 4.4 MeV. A total of 58 resonances showing an enhancement of the high energy γ-ray yield were assigned as p-waves. Assuming that their reduced neutron widths follow a Porter-Thomas distribution, their average value and then the p-wave strength function S 1 were estimated with a maximum likelihood method. The results are: average neutron width=3.4(+0.8 or -0.6)meV; S 1 = 2.0 (+0.5 or -0.4).10 -4

Evaluation of neutron nuclear data for 241Am and 243Am

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki

1982-08-01

Neutron nuclear data of 241 Am and 243 Am were evaluated for JENDL-2. Evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) and (n,4n) reaction cross sections, the resolved and unresolved resonance parameters, the angular or energy distribution of the emitted neutrons, and the average number of neutrons emitted per fission. The fission cross section was evaluated on the basis of newly measured data, and lower values than JENDL-1 were given in the subthreshold energy region. The reliability of the calculation parameters are also much improved, because experimental data became available for the total and capture cross sections of 241 Am in the high energy region. (author)

The effect of surface texture on total reflection of neutrons and X-rays from modified interfaces

DEFF Research Database (Denmark)

Goldar, A.; Roser, S.J.; Hughes, A.

2002-01-01

X-ray and neutron scattering from macroscopically rough surfaces and interfaces is considered and a new method of analysis based on the variation of the shape of the total reflection edge in the reflectivity profile is proposed. It was shown that in the limit that the correlation length and the h......X-ray and neutron scattering from macroscopically rough surfaces and interfaces is considered and a new method of analysis based on the variation of the shape of the total reflection edge in the reflectivity profile is proposed. It was shown that in the limit that the correlation length...... and the height of the surface roughness are larger than the wavelength (at least 100 times bigger) of the incoming beam, the total reflection edge in the reflection profile becomes rounded. This technique allows direct analysis of the variation of the reflectivity pro le in terms of the structure of the surface...

Normal levels of total body sodium and chlorine by neutron activation analysis

International Nuclear Information System (INIS)

Kennedy, N.S.J.; Eastell, R.; Smith, M.A.; Tothill, P.

1983-01-01

In vivo neutron activation analysis was used to measure total body sodium and chlorine in 18 male and 18 female normal adults. Corrections for body size were developed. Normalisation factors were derived which enable the prediction of the normal levels of sodium and chlorine in a subject. The coefficient of variation of normalised sodium was 5.9% in men and 6.9% in women, and of normalised chlorine 9.3% in men and 5.5% in women. In the range examined (40-70 years) no significant age dependence was observed for either element. Total body sodium was correlated with total body chlorine and total body calcium. Sodium excess, defined as the amount of body sodium in excess of that associated with chlorine, also correlated well with total body calcium. In females there was a mean annual loss of sodium excess of 1.2% after the menopause, similar to the loss of calcium. (author)

Delayed neutron yield from fast neutron induced fission of 238U

International Nuclear Information System (INIS)

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

2002-01-01

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

Neutron stochastic transport theory with delayed neutrons

International Nuclear Information System (INIS)

Munoz-Cobo, J.L.; Verdu, G.

1987-01-01

From the stochastic transport theory with delayed neutrons, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation. From this theory a general expression is obtained for the Feynman Y-function when delayed neutrons are included. The single mode approximation for the particular case of a subcritical assembly is developed, and it is shown that Y-function reduces to the familiar expression quoted in many books, when delayed neutrons are not considered, and spatial and source effects are not included. (author)

The neutron total cross-section measurement of {sup 56}Fe and {sup 57}Fe by using Japan Proton Accelerator Research Complex facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Eun Ae; Shvetsov, Valery; Cho, Moo Hyun [Pohang University of Science and Technology, Pohang (Korea, Republic of); Won, Nam Kung [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Kim, Kwang Soo; Yang, Sung Chul; Lee, Man Woo; Kim, Guin Yun [Kyungpook National University, Daegu (Korea, Republic of); Yi, Kyoung Rak; Choi, Hong Yub; Ro, Tae Ik [Dong-A University, Pusan (Korea, Republic of); Mizumoto, Motoharu; Katabuchi, Tatsuya; Igashira, Masayuki [Tokyo Institute of Technology, Tokyo (Japan)

2012-05-15

The measurement of neutron cross section using Time-Of-Flight (TOF) method gives significant information for the nuclear data research. In the present work, the neutron total cross section of {sup 56}Fe and {sup 57}Fe has been measured in the energy range between 10 eV and 100 keV by using the neutron beam produced from 3-GeV proton synchrotron accelerator. The 3-GeV proton synchrotron accelerator is located at Japan Proton Accelerator Research Complex (J-PARC) facility in Tokai village. In this study, the neutron total cross section data measured by {sup 6}Li glass scintillator detector was compared with the evaluated values of ENDF/B-VII.0

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

Monte Carlo efficiency calibration of a neutron generator-based total-body irradiator

International Nuclear Information System (INIS)

Shypailo, R.J.; Ellis, K.J.

2009-01-01

Many body composition measurement systems are calibrated against a single-sized reference phantom. Prompt-gamma neutron activation (PGNA) provides the only direct measure of total body nitrogen (TBN), an index of the body's lean tissue mass. In PGNA systems, body size influences neutron flux attenuation, induced gamma signal distribution, and counting efficiency. Thus, calibration based on a single-sized phantom could result in inaccurate TBN values. We used Monte Carlo simulations (MCNP-5; Los Alamos National Laboratory) in order to map a system's response to the range of body weights (65-160 kg) and body fat distributions (25-60%) in obese humans. Calibration curves were constructed to derive body-size correction factors relative to a standard reference phantom, providing customized adjustments to account for differences in body habitus of obese adults. The use of MCNP-generated calibration curves should allow for a better estimate of the true changes in lean tissue mass that many occur during intervention programs focused only on weight loss. (author)

Total cross-section measurements on aluminium, carbon, fluorine and hydrogen for d, d reaction neutrons using the coincidence method (1963); Mesure de la section efficace totale de l'aluminium, du carbone, du fluor et de l'hydrogene pour des neutrons de la reaction d, d par la methode des coincidences (1963)

Energy Technology Data Exchange (ETDEWEB)

Ambrosino, G; Sorriaux, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The experiment described consists in the measurement of the total cross-section of various materials: aluminium, carbon, fluorine and hydrogen, for mono-energetic 2.77 MeV neutrons obtained from the d,d reaction. The measurement is carried out by transmission. The neutrons are detected by means of a plastic scintillator mounted on a 56 AVP photomultiplier, and are isolated from all secondary phenomena (background noise, scattered neutrons) by coincidence with helium 3. which particles are associated to the neutrons from the reaction {sup 2}{sub 1}D ({sup 2}{sub 1}D, n) {sup 3}{sub 2}H The helium 3 particles are detected by a PN junction diode used with inverted polarisation. An absorption exponential has been traced out using measurements made on seven aluminium bars. The accuracy of the total cross-section measurements is about 10{sup -2}. (authors) [French] L'experience exposee dans ce rapport consiste en la mesure des sections efficaces totales de differents materiaux: aluminium, carbone, fluor et hydrogene, pour des neutrons monoenergetiques de 2,77 MeV, obtenus par la reaction d,d. La mesure est faite par transmission. Les neutrons sont detectes par un scintillateur plastique monte sur un photomultiplicateur 56 AVP, et sont separes de tout phenomene secondaire (bruit de fond, neutrons diffuses) par coincidence avec les helium 3, particules associees aux neutrons de la reaction {sup 2}{sub 1}D ({sup 2}{sub 1}D, n) {sup 3}{sub 2}H Les helium 3 sont detectes par une diode a jonction PN utilisee en polarisation inverse. Une exponentielle d'absorption a ete tracee a partir de mesures faites sur sept barreaux d'aluminium. La precision des mesures des sections efficaces totales est de l'ordre de 10{sup -2}. (auteurs)

Mass number dependence of total neutron cross section; a discussion based on the semi-classical optical model

International Nuclear Information System (INIS)

Angeli, Istvan

1990-01-01

The dependence of total neutron cross section on mass number can be calculated by the black nucleus formula, according to the optical model. The fine structure of mass number dependence is studied, and a correction factor formula is given on the basis of a semi-classical optical model. Yielding results in good agreement with experimental data. In addition to the mass number dependence, the neutron-energy dependence can also be calculated using this model. (K.A.)

Curves and tables of neutron cross sections

International Nuclear Information System (INIS)

Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi

1990-07-01

Neutron cross-section curves from the Japanese Evaluated Nuclear Data Library version 3, JENDL-3, are presented in both graphical and tabular form for users in a wide range of application areas in the nuclear energy field. The contents cover cross sections for all the main reactions induced by neutrons with an energy below 20 MeV including; total, elastic scattering, capture, and fission, (n,n'), (n,2n), (n,3n), (n,α), (n,p) reactions. The 2200 m/s cross-section values, resonance integrals, and Maxwellian- and fission-spectrum averaged cross sections are also tabulated. (author)

Total neutron cross-sections of /sup 151/Eu, /sup 153/Eu and Eu below 1 eV

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Ashry, A.; Hamouda, I.

1981-01-01

Total neutron cross-section measurements have been carried out for natural Eu and its stable isotopes in the energy range from 3 meV to 1 eV. The measurements were performed using two time-of-flight spectrometers installed in front of two of the horizontal channels of the ET-RR-1 reactor. The following results have been obtained: sigmasub(a) (/sup 151/Eu) = (9180 +- 150) b at 0.0253 eV, sigmasub(s) (/sup 153/Eu) = (375 +- 20) b at 0.0253 eV, sigmasub(d) (Eu) = (4600 +- 120) b at 0.0253 eV. The contribution of the resonance to the total neutron cross-sections, in the thermal region, was calculated using the single-level Breit-Wigner formula.

Neutron total cross section measurements in the energy region from 47 keV to 20 MeV

International Nuclear Information System (INIS)

Poenitz, W.P.; Whalen, J.F.

1983-05-01

Neutron total cross sections were measured for 26 elements. Data were obtained in the energy range from 47 keV to 20 MeV for 11 elements in the range of light-mass fission products. Previously reported measurements for eight heavy and actinide isotopes were extended to 20 MeV. Data were also obtained for Cu (47 keV to 1.4 MeV) and for Sc, Zn, Nd, Hf, and Pt (1.8 to 20 MeV). The present work is part of a continuing effort to provide accurate neutron total cross sections for evaluations and for optical-model parameteriztions. The latter are required for the derivation of other nuclear-data information of importance to applied programs. 37 references

Neutron-induced displacement damage analysis (with particular reference to zirconium)

International Nuclear Information System (INIS)

Woo, C.H.

1978-10-01

Neutron irradiation produces damage in a solid, initially in the form of atomic displacements. As a first step to understanding the effects of irradiation damage in reactor structural materials, information on the initial atomic displacements is necessary. The computer program DISPKAN, based on an extension and generalization of the program RICE, written at ORNL for such calculations, has been developed and installed on the CDC system at CRNL. Using neutron scattering data from ENDF/B files, DISPKAN performs a displacement and PKO analysis on the initial damage caused by neutrons from a given spectrum. The following quantities are calculated: the displacement rate per unit neutron flux, the PKO spectrum, the displacement spectrum, the fraction of PKO's with energy above T, the fraction of displacements produced by PKO's with energy above T, the average PKO energy, the average number of displacements produced per PKO, and the total number of PKO's produced per atom of the solid per unit fluence. The input and output formats of the program are explained. Sample runs are demonstrated. Results for zirconium, exposed to five neutron spectra typically available to experimentalist, are given to illustrate the spectral dependence of the initial displacement events. (author)

Evaluation of apoptosis and micronucleation induced by reactor neutron beams with two different cadmium ratios in total and quiescent cell populations within solid tumors

International Nuclear Information System (INIS)

Masunaga, Shin-ichiro; Ono, Koji; Sakurai, Yoshinori; Takagaki, Masao; Kobayashi, Tooru; Kinashi, Yuko; Suzuki, Minoru

2001-01-01

Purpose: Response of quiescent (Q) and total tumor cells in solid tumors to reactor neutron beam irradiation with two different cadmium (Cd) ratios was examined in terms of micronucleus (MN) frequency and apoptosis frequency, using four different tumor cell lines. Methods and Materials: C57BL mice bearing EL4 tumors, C3H/He mice bearing SCC VII or FM3A tumors, and Balb/c mice bearing EMT6/KU tumors received 5-bromo-2'-deoxyuridine (BrdU) continuously for 5 days via implanted mini-osmotic pumps to label all proliferating (P) cells. Thirty min after i.p. injection of sodium borocaptate- 10 B (BSH), or 3 h after oral administration of p-boronophenylalanine- 10 B (BPA), the tumors were irradiated with neutron beams. The tumors without 10 B-compound administration were irradiated with neutron beams or γ-rays. This neutron beam irradiation was performed using neutrons with two different Cd ratios. The tumors were then excised, minced, and trypsinized. The tumor cell suspensions thus obtained were incubated with cytochalasin-B (a cytokinesis blocker), and the MN frequency in cells without BrdU labeling (=Q cells) was determined using immunofluorescence staining for BrdU. Meanwhile, for apoptosis assay, 6 h after irradiation, tumor cell suspensions obtained in the same manner were fixed, and the apoptosis frequency in Q cells was also determined with immunofluorescence staining for BrdU. The MN and apoptosis frequencies in total (P+Q) tumor cells were determined from the tumors that were not pretreated with BrdU. Results: Without 10 B-compounds, the sensitivity difference between total and Q cells was reduced by neutron beam irradiation. Under our particular neutron beam irradiation condition, relative biological effectiveness (RBE) of neutrons was larger in Q cells than in total cells, and the RBE values were larger for low Cd-ratio than high Cd-ratio neutrons. With 10 B-compounds, both frequencies were increased for each cell population, especially for total cells. BPA

A Study on the Neutron Dose Distribution in Case of 10 MV X-rays Radiotherapy

International Nuclear Information System (INIS)

Park, Cheol Soo; Shin, Seong Soo; Lim, Cheong Hwan; Jung, Hong Ryang

2008-01-01

This study is to measure the radiation dose of neutrons generated by the particle accelerator during X-ray (photon) treatment with a neutron detection method by using CR-39, and to research how the generation of neutrons may incur problems associated with radiation doses for patient treatment when using high energy photons for cancer treatment as a clinical application. The findings are summarized as follows : The results showed that average 0.35 mSv was measured with exposure of 1 Gy photon in case of fast neutron, 0.65 mSv with exposure of 2 Gy photon, 1.82 mSv exposure of 5 Gy, 0.26 mSv with exposure of 1 Gy photon in case of thermal neutron, 0.56 mSv with exposure of 2 Gy photon, and 1.23 mSv with exposure of 5 Gy of photon. By measuring the occurrence of neutron by using Wedge Filter, it has been confirmed that the occurrence of neutrons increased when using Wedge Filter. The results also showed that more neutrons were detected over the existing experiments when using an SRS Cone requiring high doses of radiation. Total 2.85 mSv neutrons were found on the average with exposure of 5 Gy photon in case of fast neutron and 1.37 mSv neutrons were found on the average with exposure of 5 Gy photon in case of thermal neutron. During the general treatment, about 1.6 times more neutrons over 5 Gy photon were found in case of fast neutron and about 1.12 time more neutrons over 5 Gy photon were found in case of thermal neutron.

Total body calcium by neutron activation analysis. Reference data for children

International Nuclear Information System (INIS)

Ellis, K.J.; Shypailo, R.J.

2001-01-01

There is a paucity of data on the chemical composition of the human body during growth. Total body calcium (TBCa) has been reported for only one male child, aged 41/2 yr. TBCa values for 25 children and 27 young women using in vivo neutron activation analysis have been obtained. TBCa results were lower than those reported for the one male cadaver, as well as the estimates derived for the 'Reference Man' model. It was concluded that the reference values for TBCa may need to be adjusted to appropriately describe skeletal mineralization of contemporary children. (author)

In vivo Prompt Gamma Neutron Activation Analysis Facility for Total Body Nitrogen and Cd

International Nuclear Information System (INIS)

Munive, Marco; Revilla, Angel; Solis, Jose L.

2007-01-01

A Prompt Gamma Neutron Activation Analysis (PGNAA) system has been designed and constructed to measure the total body nitrogen and Cd for in vivo studies. An aqueous solution of KNO 3 was used as phantom for system calibration. The facility has been used to monitor total body nitrogen (TBN) of mice and found that is related to their diet. Some mice swallowed diluted water with Cl 2 Cd, and the presence of Cd was detected in the animals. The minimum Cd concentration that the system can detect was 20 ppm

Atlas of neutron resonances

CERN Document Server

Mughabghab, Said

2018-01-01

Atlas of Neutron Resonances: Resonance Properties and Thermal Cross Sections Z= 1-60, Sixth Edition, contains an extensive list of detailed individual neutron resonance parameters for Z=1-60, as well as thermal cross sections, capture resonance integrals, average resonance parameters and a short survey of the physics of thermal and resonance neutrons. The long introduction contains: nuclear physics formulas aimed at neutron physicists; topics of special interest such as valence neutron capture, nuclear level density parameters, and s-, p-, and d-wave neutron strength functions; and various comparisons of measured quantities with the predictions of nuclear models, such as the optical model. As in the last edition, additional features have been added to appeal to a wider spectrum of users. These include: spin-dependent scattering lengths that are of interest to solid-state physicists, nuclear physicists and neutron evaluators; calculated and measured Maxwellian average 5-keV and 30-keV capture cross sections o...

Total number albedo and average cosine of the polar angle of low-energy photons reflected from water

Directory of Open Access Journals (Sweden)

Marković Srpko

2007-01-01

Full Text Available The total number albedo and average cosine of the polar angle for water and initial photon energy range from 20 keV to 100 keV are presented in this pa per. A water shield in the form of a thick, homogenous plate and per pendicular incidence of the monoenergetic photon beam are assumed. The results were obtained through Monte Carlo simulations of photon reflection by means of the MCNP computer code. Calculated values for the total number albedo were compared with data previously published and good agreement was confirmed. The dependence of the average cosine of the polar angle on energy is studied in detail. It has been found that the total average cosine of the polar angle has values in the narrow interval of 0.66-0.67, approximately corresponding to the reflection angle of 48°, and that it does not depend on the initial photon energy.

Nanocrystalline alloys of Fe-Cu-Nb-Si-B after neutron irradiation

International Nuclear Information System (INIS)

Sitek, J.; Toth, I.; Degmova, J.; Uvacik, P.

1997-01-01

Transmission Moessbauer spectroscopy was used to study changes induced by irradiation of amorphous and nanocrystalline samples. In an as-cast sample, neutrons mostly affect the orientation of the net magnetic moment. The average hyperfine field decreases with increasing neutron fluencies. In the case of the nanocrystalline samples a new disordered structure is created in the amorphous remainder corresponding to boride phases as it is shown in the samples isothermally heated from 1 to 8 hours. The structural changes of the amorphous remainder depend on the stage of crystallization and total neutron fluencies. (author). 1 tab., 3 figs., 7 refs

Surrogate 239Pu(n, fxn) and 241Pu(n, fxn) average fission-neutron-multiplicity measurements

Energy Technology Data Exchange (ETDEWEB)

Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Alan, B. S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Akindele, O. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Casperson, R. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hughes, R. O. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fisher, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-09-26

We have constructed a new neutron-charged-particle detector array called NeutronSTARS. It has been described extensively in LLNL-TR-703909 [1] and Akindele et al [2]. We have used this new neutron-charged-particle array to measure the 241Pu and 239Pu fissionneutron multiplicity as a function of equivalent incident-neutron energy from 100 keV to 20 MeV. The experimental approach, detector array, data analysis, and results are summarized in the following sections.

Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries

Science.gov (United States)

Pritychenko, B.; Mughabghab, S. F.

2012-12-01

We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-process Maxwellian-averaged cross sections and astrophysical reaction rates, systematically calculate uncertainties, and provide additional insights on currently available neutron-induced reaction data. Nuclear reaction calculations are discussed and new results are presented. Due to space limitations, the present paper contains only calculated Maxwellian-averaged cross sections and their uncertainties. The complete data sets for all results are published in the Brookhaven National Laboratory report.

Nuclear structure studies via neutron interactions. Progress report, 1 July 1994--30 June 1995

International Nuclear Information System (INIS)

Carlton, R.F.

1995-01-01

Neutron total cross section measurements of 122 Sn has been analyzed over the energy range 0.013 to 0.310 MeV. The R - matrix analysis has resulted in resonance and average parameters which provide for a complete representation of the neutron entrance the s 1/2 , p 1/2 , and p 3/2 contributions. The s - and p - neutron strength functions have been determined to be 0.123 ± 0.023 and 2.0 ± 0.200, respectively (in units of 10 -4 ). Limits are placed on level spacings and strength functions for the individual partial wave components. The s - wave potential scattering radius has been determined to be 6.3 ± 0.1 fm. Average scattering functions, deduced from the average parameters, have been used to determine the real well depth of an optical model potential which reproduces these functions. We find, as have others, that the real well depth is parity dependent. Two computer codes were developed: (1) a minimization program for finding parameters of a dispersive optical model which reproduce the high resolution data of the ORELA facility, and (2) an X-windows graphical user interface has been developed to permit interactive resonance analysis of the neutron total cross section data using analysis-and-minimization codes developed previously. These will contribute to the more routine analysis of ORELA data

Alteration of sensitivity of intratumor quiescent and total cells to γ-rays following thermal neutron irradiation with or without 10B-compound

International Nuclear Information System (INIS)

Masunaga, Shin-ichiro; Ono, Koji; Suzuki, Minoru; Sakurai, Yoshinori; Kobayashi, Tooru; Takagaki, Masao; Kinashi, Yuko; Akaboshi, Mitsuhiko

2000-01-01

Purpose: Changes in the sensitivity of intratumor quiescent (Q) and total cells to γ-rays following thermal neutron irradiation with or without 10 B-compound were examined. Methods and Materials: 5-Bromo-2'-deoxyuridine (BrdU) was injected to SCC VII tumor-bearing mice intraperitoneally 10 times to label all the proliferating (P) tumor cells. As priming irradiation, thermal neutrons alone or thermal neutrons with 10 B-labeled sodium borocaptate (BSH) or dl-p-boronophenylalanine (BPA) were administered. The tumor-bearing mice then received a series of γ-ray radiation doses, 0 through 24 h after the priming irradiation. During this period, no BrdU was administered. Immediately after the second irradiation, the tumors were excised, minced, and trypsinized. Following incubation of tumor cells with cytokinesis blocker, the micronucleus (MN) frequency in cells without BrdU labeling (= Q cells at the time of priming irradiation) was determined using immunofluorescence staining for BrdU. The MN frequency in the total (P + Q) tumor cells was determined from the tumors that were not pretreated with BrdU before the priming irradiation. To determine the BrdU-labeled cell ratios in the tumors at the time of the second irradiation, each group also included mice that were continuously administered BrdU until just before the second irradiation using mini-osmotic pumps which had been implanted subcutaneously 5 days before the priming irradiation. Results: In total cells, during the interval between the two irradiations, the tumor sensitivity to γ-rays relative to that immediately after priming irradiation decreased with the priming irradiation ranking in the following order: thermal neutrons only > thermal neutrons with BSH > thermal neutrons with BPA. In contrast, in Q cells, during that time the sensitivity increased in the following order: thermal neutrons only 10 B-compound, especially BPA, in thermal neutron irradiation causes the recruitment from the Q to P population

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

Experimental results on RPC neutron sensitivity

Energy Technology Data Exchange (ETDEWEB)

Abbrescia, M.; Altieri, S.; Baratti, V.; Barnaba, O.; Belli, G.; Bruno, G.; Colaleo, A.; De Vecchi, C.; Guida, R. E-mail: roberto.guida@pv.infn.it; Iaselli, G.; Imbres, E.; Loddo, F.; Maggi, M.; Marangelli, B.; Musitelli, G.; Nardo, R.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Ratti, S.; Riccardi, C.; Romano, F.; Torre, P.; Vicini, A.; Vitulo, P

2003-08-01

RPC neutron sensitivity has been studied during two tests done with different neutrons energies. In the first test, neutrons from spontaneous fission events of {sup 252}Cf were used (average energy 2 MeV); while in the second test neutrons were produced using a 50 MeV deuteron beam on a 1 cm thick beryllium target (average energy 20 MeV). Preliminary results show that the neutron sensitivity in double gap mode is (0.52{+-}0.03)x10{sup -3} at about 2 MeV and (5.3{+-}0.5)x10{sup -3} at about 20 MeV.

Prompt fission neutron spectra from fission induced by 1 to 8 MeV neutrons on 235U and 239Pu using the double time-of-flight technique

International Nuclear Information System (INIS)

Noda, S.; Haight, R. C.; Nelson, R. O.; Devlin, M.; O'Donnell, J. M.; Chatillon, A.; Granier, T.; Belier, G.; Taieb, J.; Kawano, T.; Talou, P.

2011-01-01

Prompt fission neutron spectra from 235 U and 239 Pu were measured for incident neutron energies from 1 to 200 MeV at the Weapons Neutron Research facility (WNR) of the Los Alamos Neutron Science Center, and the experimental data were analyzed with the Los Alamos model for the incident neutron energies of 1-8 MeV. A CEA multiple-foil fission chamber containing deposits of 100 mg 235 U and 90 mg 239 Pu detected fission events. Outgoing neutrons were detected by the Fast Neutron-Induced γ-Ray Observer array of 20 liquid organic scintillators. A double time-of-flight technique was used to deduce the neutron incident energies from the spallation target and the outgoing energies from the fission chamber. These data were used for testing the Los Alamos model, and the total kinetic energy parameters were optimized to obtain a best fit to the data. The prompt fission neutron spectra were also compared with the Evaluated Nuclear Data File (ENDF/B-VII.0). We calculate average energies from both experimental and calculated fission neutron spectra.

Quantitative neutron radiography using neutron absorbing honeycomb

International Nuclear Information System (INIS)

Tamaki, Masayoshi; Oda, Masahiro; Takahashi, Kenji; Ohkubo, Kohei; Tasaka, Kanji; Tsuruno, Akira; Matsubayashi, Masahito.

1993-01-01

This investigation concerns quantitative neutron radiography and computed tomography by using a neutron absorbing honeycomb collimator. By setting the neutron absorbing honeycomb collimator between object and imaging system, neutrons scattered in the object were absorbed by the honeycomb material and eliminated before coming to the imaging system, but the neutrons which were transmitted the object without interaction could reach the imaging system. The image by purely transmitted neutrons gives the quantitative information. Two honeycombs were prepared with coating of boron nitride and gadolinium oxide and evaluated for the quantitative application. The relation between the neutron total cross section and the attenuation coefficient confirmed that they were in a fairly good agreement. Application to quantitative computed tomography was also successfully conducted. The new neutron radiography method using the neutron-absorbing honeycomb collimator for the elimination of the scattered neutrons improved remarkably the quantitativeness of the neutron radiography and computed tomography. (author)

Neutron reflectometry

International Nuclear Information System (INIS)

Van Well, A.A.

1999-01-01

Neutron research where reflection, refraction, and interference play an essential role is generally referred to as 'neutron optics'. The neutron wavelength, the scattering length density and the magnetic properties of the material determine the critical angle for total reflection. The theoretical background of neutron reflection, experimental methods and the interpretation of reflection data are presented. (K.A.)

Measurement of neutron sensitivity of self powered neutron detectors

International Nuclear Information System (INIS)

Mahant, A.K.; Yeshuraja, V.; Ghodke, Shobha

2005-01-01

Self powered neutron detectors (SPNDs ) will form the part of Reactor Instrumentation in the upcoming 500 MWe power reactors. ECIL has developed Vanadium and Cobalt SPNDs for NPCIL to be used in regulation and protection channels. Experimental determination of neutron sensitivity of the vanadium and cobalt Self Powered Neutron Detectors (SPNDs) was carried out in A-l location of Apsara reactor at BARC. The measurements involved determination of total detector signal, its various components and the thermal neutron flux at the detector location. The paper describes the experimental techniques used to measure various parameters required to evaluate the neutron sensitivity of the SPNDs and also the parameters required to ascertain the integrity of SPNDs. Neutron flux measurement was done by gold foil irradiation technique. The predominant signal component from the vanadium SPND is Ib the current due to activation of the vanadium emitter, it forms about 85% of the total signal. The other components I n,γ due to the capture gamma rays of 52 V and I externalγ produced by the external reactor gamma rays contribute about 10% and 5% respectively to the total signal. Whereas in the cobalt SPND the main signal component is due to the capture gamma rays of 60 Co and accounts for about the 95% of the total signal. Remaining 5% signal is due to external reactor gamma rays. (author)

Measurements of fission product yield in the neutron-induced fission of {sup 238}U with average energies of 9.35 MeV and 12.52 MeV

Energy Technology Data Exchange (ETDEWEB)

Mukerji, Sadhana; Krishnani, Pritam Das; Suryanarayana, Saraswatula Venkat; Naik, Haladhara; Goswami, Ashok [Bhabha Atomic Research Centre, Mumbai (India); Shivashankar, Byrapura Siddaramaiah [Manipal University, Manipal (India); Mulik, Vikas Kaluram [University of Pune, Pune (India)

2014-07-15

The yields of various fission products in the neutron-induced fission of {sup 238}U with the flux-weighted averaged neutron energies of 9.35 MeV and 12.52 MeV were determined by using an off-line gamma ray spectroscopic technique. The neutrons were generated using the {sup 7}Li(p, n) reaction at Bhabha Atomic Research Centre-Tata Institute of Fundamental Research Pelletron facility, Mumbai. The gamma- ray activities of the fission products were counted in a highly-shielded HPGe detector over a period of several weeks to identify the decaying fission products. At both the neutron energies, the fission-yield values are reported for twelve fission product. The results obtained from the present work have been compared with the similar data for mono-energetic neutrons of comparable energy from the literature and are found to be in good agreement. The peak-to-valley (P/V) ratios were calculated from the fission-yield data and were found to decreases for neutron energy from 9.35 to 12.52 MeV, which indicates the role of excitation energy. The effect of the nuclear structure on the fission product-yield is discussed.

Neutron capture cross section of $^{90}$Zr Bottleneck in the s-process reaction flow

CERN Document Server

Tagliente, G; Milazzo, P M; Moreau, C; Aerts, G; Abbondanno, U; Alvarez, H; Alvarez-Velarde, F; Andriamonje, Samuel A; Andrzejewski, J; Assimakopoulos, Panayiotis; Audouin, L; Badurek, G; Baumann, P; Bečvář, F; Berthoumieux, E; Bisterzo, S; Calviño, F; Calviani, M; Cano-Ott, D; Capote, R; Carrapiço, C; Cennini, P; Chepel, V; Chiaveri, Enrico; Colonna, N; Cortés, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillman, I; Domingo-Pardo, C; Dridi, W; Durán, I; Eleftheriadis, C; Embid-Segura, M; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Furman, W; Gallino, R; Gonçalves, I; Gonzalez-Romero, E; Gramegna, F; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martínez, A; Igashira, M; Jericha, E; Käppeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Kerveno, M; Köhler, P; Kossionides, E; Krtička, M; Lamboudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marrone, S; Martínez, T; Massimi, C; Mastinu, P; Mengoni, A; Mosconi, M; Neves, F; Oberhummer, Heinz; O'Brien, S; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M T; Plag, R; Plompen, A; Plukis, A; Poch, A; Praena, J; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rubbia, Carlo; Rudolf, G; Rullhusen, P; Salgado, J; Santos, J; Sarchiapone, L; Savvidis, I; Stéphan, C; Taín, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarín, D; Vincente, M, C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescher, M; Wisshak, K

2008-01-01

The neutron capture cross sections of the Zr isotopes have important implications in nuclear astrophysics and for reactor design. The small cross section of the neutron magic nucleus 90Zr, which accounts for more than 50% of natural zirconium represents one of the key isotopes for the stellar s-process, because it acts as a bottleneck in the neutron capture chain between the Fe seed and the heavier isotopes. The same element, Zr, also is an important component of the structural materials used in traditional and advanced nuclear reactors. The (n,γ) cross section has been measured at CERN, using the n_TOF spallation neutron source. In total, 45 resonances could be resolved in the neutron energy range below 70 keV, 10 being observed for the first time thanks to the high resolution and low backgrounds at n_TOF. On average, the Γγ widths obtained in resonance analyses with the R-matrix code SAMMY were 15% smaller than reported previously. By these results, the accuracy of the Maxwellian averaged cross section f...

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

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2001-01-01

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

Resistive plate chamber neutron and gamma sensitivity measurement with a {sup 252}Cf source

Energy Technology Data Exchange (ETDEWEB)

Abbrescia, M.; Altieri, S.; Baratti, V.; Barnaba, O.; Belli, G.; Bruno, G.; Colaleo, A.; DeVecchi, C.; Guida, R. E-mail: roberto.guida@pv.infn.it; Iaselli, G.; Imbres, E.; Loddo, F.; Maggi, M.; Marangelli, B.; Musitelli, G.; Nardo, R.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Ratti, S.; Riccardi, C.; Romano, F.; Torre, P.; Vicini, A.; Vitulo, P.; Volpe, F

2003-06-21

A bakelite double gap Resistive Plate Chamber (RPC), operating in avalanche mode, has been exposed to the radiation emitted from a {sup 252}Cf source to measure its neutron and gamma sensitivity. One of the two gaps underwent the traditional electrodes surface coating with linseed oil. RPC signals were triggered by fission events detected using BaF{sub 2} scintillators. A Monte Carlo code, inside the GEANT 3.21 framework with MICAP interface, has been used to identify the gamma and neutron contributions to the total number of collected RPC signals. A neutron sensitivity of (0.63{+-}0.02)x10{sup -3} (average energy 2 MeV) and a gamma sensitivity of (14.0{+-}0.5)x10{sup -3} (average energy 1.5 MeV) have been measured in double gap mode. Measurements done in single gap mode have shown that both neutron and gamma sensitivity are independent of the oiling treatment.

SASSI, Total and Differential Elastic and Inelastic Neutron Cross-Sections by Hauser-Feshbach

International Nuclear Information System (INIS)

Benzi, V.; Fabbri, F.; Zuffi, L.

2001-01-01

1 - Nature of physical problem solved: Neutron total and differential elastic and inelastic cross-section evaluation by means of the statistical model of Hauser-Feshbach (1) as modified by D. Goldman (2) (3). The Goldman modification includes the effect of spin-orbit coupling on transmission coefficients. 2 - Method of solution: For numerical integration the Fox-Goodwins method is used. 3 - Restrictions on the complexity of the problem: Angular momentum I less than or equal to 50. Number of excited levels less than or equal to 30

Neutron transmission measurements of poly and pyrolytic graphite crystals

Science.gov (United States)

Adib, M.; Abbas, Y.; Abdel-Kawy, A.; Ashry, A.; Kilany, M.; Kenawy, M. A.

The total neutron cross-section measurements of polycrystalline graphite have been carried out in a neutron wavelength from 0.04 to 0.78 nm. This work also presents the neutron transmission measurements of pyrolytic graphite (PG) crystal in a neutron wavelength band from 0.03 to 0.50 nm, at different orientations of the PG crystal with regard to the beam direction. The measurements were performed using three time-of-flight (TOF) spectrometers installed in front of three of the ET-RR-1 reactor horizontal channels. The average value of the coherent scattering amplitude for polycrystalline graphite was calculated and found to be bcoh = (6.61 ± 0.07) fm. The behaviour of neutron transmission through the PG crystal, while oriented at different angles with regard to the beam direction, shows dips at neutron wavelengths corresponding to the reflections from (hkl) planes of hexagonal graphite structure. The positions of the observed dips are found to be in good agreement with the calculated ones. It was also found that a 40 mm thick PG crystal is quite enough to reduce the second-order contamination of the neutron beam from 2.81 to 0.04, assuming that the incident neutrons have a Maxwell distribution with neutron gas temperature 330 K.

Neutron transmission measurements of poly and pyrolytic graphite crystals

International Nuclear Information System (INIS)

Adib, M.; Abdel-Kawy, A.; Kilany, M.

1989-01-01

The total neutron cross-section measurements of polycrystalline graphite have been carried out in a neutron wavelength from 0.04 to 0.78 nm. This work also presents the neutron transmission measurements of pyrolytic graphite (PG) crystal in a neutron wavelength band from 0.03 to 0.50 nm, at different orientations of the PG crystal with regard to the beam direction. The measurements were performed using three time-of-flight (TOF) spectrometers installed in front of three of the ET-RR-1 reactor horizontal channels. The average value of the coherent scattering amplitude for polycrystalline graphite was calculated and found to be b coh = (6.61 ± 0.07) fm. The behaviour of neutron transmission through the PG crystal, while orientated at different angles with regard to the beam direction, shows dips at neutron wavelengths corresponding to the reflections from (hk1) planes of hexagonal graphite structure. The positions of the observed dips are found to be in good agreement with the calculated ones. It was also found that a 40 mm thick PG crystal is quite enough to reduce the second-order contamination of the neutron beam from 2.81 to 0.04, assuming that the incident neutrons have a Maxwell distribution with neutron gas temperature 330 K. (author)

Multi-fold correlations between 252Cf (sf) fragments and fission neutrons/γ-rays

International Nuclear Information System (INIS)

Duering, I.; Jahnke, U.

1993-01-01

Direction-sensitive spectroscopy of fission fragments (twin ionization chamber with Frisch grids) was combined with the measurement of neutron multiplicity distribution (P(ν), average total γ-ray energy (2x2 π Gd-loaded scintillator) as well as energy and angular distribution of neutrons and γ-rays. Based on the careful account for necessary corrections, scission configurations given by mass asymmetry, elongation (total kinetic energy of fragments), and shape asymmetry (ν 1 /ν 2 ) can be studied exclusively in correlation with differential distributions of emission products. The scheme for correcting the neutron multiplicity distribution including its separation into the contributions from the complementary fragments is presented in detail. The mass yield for extreme anti ν 1 / anti ν 2 ratios show fine structures indicating the cold shape-asymmetric fission. (orig.)

Constitutive laws for the neutron density current

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Morales-Sandoval, Jaime B.; Vazquez-Rodriguez, Rodolfo; Espinosa-Martinez, Erick-G.

2008-01-01

In this technical note, a fractional wave equation for the average neutron motion in nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fickian and non-Fickian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional exponent that can be manipulated to obtain the best representation of the neutron transport phenomena. The detrended fluctuation analysis (DFA) method is presented in this paper to estimate the fractional exponent

Note on neutron scattering and the optical model near A = 208

International Nuclear Information System (INIS)

Guenther, P.; Havel, D.; Smith, A.

1976-09-01

Elastic neutron scattering cross sections of 206 Pb, 207 Pb, 208 Pb and 209 Bi are measured at incident neutron energy intervals of approx. 25 keV from 0.6 to 1.0 MeV with resolutions of approx. 25 keV. Optical model parameters are obtained from the energy-averaged experimental results for each of the isotopes. The observed elastic-neutron-scattering distributions and derived parameters for the lead isotopes (doubly magic or neutron holes in the closed shell) tend to differ from those of 209 Bi (doubly closed shell plus a proton). These potentials, derived in the approx. spherical region of A approximately 208, are extrapolated for the analysis of total and scattering cross sections of 238 U introducing only a small N-Z/A dependence and the known deformation of 238 U. Good descriptions of 238 U total cross sections are obtained from a few hundred keV to 10.0 MeV and the prediction of measured scattering distributions in the low MeV region are as suitable as frequently reported with other specially developed potentials

New neutron capture and total cross section measurements on 88Sr and their impact on s-process nucleosynthesis

International Nuclear Information System (INIS)

Koehler, P.E.; Spencer, R.R.; Guber, K.H.

1998-01-01

The authors have made new and improved measurements of the neutron capture and total cross sections of 88 Sr at the Oak Ridge Electron Linear Accelerator (ORELA). Improvements over previous measurements include a wider incident neutron energy range, the use of metallic rather than carbonate samples, better background subtraction, reduced sensitivity to sample-dependent backgrounds, and better pulse-height weighting functions. Because of its small cross section, the 88 Sr(n,γ) reaction is an important bottleneck during the s-process nucleosynthesis. Hence, an accurate determination of this rate is needed to better constrain the neutron exposure in s-process models and to more fully exploit the recently discovered isotopic anomalies in certain meteorites. They describe the experimental procedures, compare the results to previous data, and discuss their astrophysical impact

Examination of total cross section resonance structure of niobium and silicon in neutron transmission experiments

Science.gov (United States)

Andrianova, Olga; Lomakov, Gleb; Manturov, Gennady

2017-09-01

The neutron transmission experiments are one of the main sources of information about the neutron cross section resonance structure and effect in the self-shielding. Such kind of data for niobium and silicon nuclides in energy range 7 keV to 3 MeV can be obtained from low-resolution transmission measurements performed earlier in Russia (with samples of 0.027 to 0.871 atom/barn for niobium and 0.076 to 1.803 atom/barn for silicon). A significant calculation-to-experiment discrepancy in energy range 100 to 600 keV and 300 to 800 keV for niobium and silicon, respectively, obtained using the evaluated nuclear data library ROSFOND, were found. The EVPAR code was used for estimation the average resonance parameters in energy range 7 to 600 keV for niobium. For silicon a stochastic optimization method was used to modify the resolved resonance parameters in energy range 300 to 800 keV. The improved ROSFOND evaluated nuclear data files were tested in calculation of ICSBEP integral benchmark experiments.

Evaluation of neutron nuclear data of 241pu for JENDL-2

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki; Sekine, Nobuo

1984-06-01

Neutron nuclear data of 241 Pu were newly evaluated for JENDL-2. Evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) and (n,4n) reaction cross sections, the resolved and unresolved resonance parameters, the angular and energy distributions of emitted neutrons, and the average number of neutrons emitted per fission. The simultaneous evaluation method was adopted for the fission cross section so as to keep the consistency among the main fissile and fertile material nuclides. The theoretical calculations based on the spherical optical model and the statistical model were also used, when the experimental data were not sufficient. Discussion is given on the evaluation method. (author)

Pulsed neutron generator for use with pulsed neutron activation techniques

International Nuclear Information System (INIS)

Rochau, G.E.

1980-01-01

A high-output, transportable, pulsed neutron generator has been developed by Sandia National Laboratories for use with Pulsed Neutron Activation (PNA) techniques. The PNA neutron generator generates > 10 10 14 MeV D-T neutrons in a 1.2 millisecond pulse. Each operation of the unit will produce a nominal total neutron output of 1.2 x 10 10 neutrons. The generator has been designed to be easily repaired and modified. The unit requires no additional equipment for operation or measurement of output

Neutron resonance spectroscopy on 113Cd: The p-wave levels

International Nuclear Information System (INIS)

Frankle, C.M.; Bowman, C.D.; Bowman, J.D.; Seestrom, S.J.; Sharapov, E.I.; Popov, Y.P.; Roberson, N.R.

1992-01-01

Weak levels in the compound nucleus 114 Cd were located by neutron time-of-flight spectroscopy techniques. Neutron capture measurements were performed with both a natural cadmium target and a highly enriched 113 Cd target. A total of 22 new resonances were located in the neutron energy interval 20-500 eV and were assumed to be p-wave. Resonance parameters, E 0 and gΓ n , are given for the newly identified levels. The p-wave strength function was determined to be 10 4 S 1 =2.8±0.8 and the average level spacing left-angle D 1 right-angle=14 eV. Comparison of the reduced widths with a Porter-Thomas distribution is consistent with having missed 15% of the p-wave levels

Thermal neutron diffusion parameters in homogeneous mixtures

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

Thermal neutron diffusion parameters in homogeneous mixtures

International Nuclear Information System (INIS)

Drozdowicz, K.; Krynicka, E.

1995-01-01

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

Slow neutron total cross-section, transmission and reflection calculation for poly- and mono-NaCl and PbF2 crystals

Science.gov (United States)

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

2016-10-01

A detailed study about the calculation of total neutron cross-section, transmission and reflection from crystalline materials was performed. The developed computer code is approved to be sufficient for the required calculations, also an excellent agreement has been shown when comparing the code results with the other calculated and measured values. The optimal monochromator and filter parameters were discussed in terms of crystal orientation, mosaic spread, and thickness. Calculations show that 30 cm thick of PbF2 poly-crystal is an excellent cold neutron filter producing neutron wavelengths longer than 0.66 nm needed for the investigation of magnetic structure experiments. While mono-crystal filter PbF2 cut along its (1 1 1), having mosaic spread (η = 0.5°) and thickness 10 cm can only transmit thermal neutrons of the desired wavelengths and suppress epithermal and γ-rays forming unwanted background, when it is cooled to liquid nitrogen temperature. NaCl (2 0 0) and PbF2 (1 1 1) monochromator crystals having mosaic spread (η = 0.5°) and thickness 10 mm shows high neutron reflectivity for neutron wavelengths (λ = 0.114 nm and λ = 0.43 nm) when they used as a thermal and cold neutron monochromators respectively with very low contamination from higher order reflections.

The averaged cross sections of natural carbon in the energy region 90-160 keV

International Nuclear Information System (INIS)

Gritzay, O.O.; Volkovetskyi, S.P.; Libman, V.A.

2012-01-01

This paper presents the results of measurements of the total cross-section of carbon in the energy region 90 - 160 keV. These results were obtained using a method of the modified filtered beams, developed in the Neutron Physics Department and implemented in the horizontal experimental channel HEC-9 at the Kyiv research reactor. The experiment was carried out using 5 modified filters. Ten values of the averaged cross-sections of carbon were obtained

Ramsauer effect in triplet neutron-neutron scattering

International Nuclear Information System (INIS)

Pupyshev, V.V.; Solovtsova, O.P.

1995-01-01

As we show, due to interplay of pure nuclear and magnetic moment interactions, the total cross section of triplet neutron-neutron scattering should possess a non-zero limit at E cm = 0 and a local minimum at ∼ 20 keV. 17 refs., 1 fig

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

Total average diastolic longitudinal displacement by colour tissue doppler imaging as an assessment of diastolic function

DEFF Research Database (Denmark)

de Knegt, Martina Chantal; Biering-Sørensen, Tor; Søgaard, Peter

2016-01-01

BACKGROUND: The current method for a non-invasive assessment of diastolic dysfunction is complex with the use of algorithms of many different echocardiographic parameters. Total average diastolic longitudinal displacement (LD), determined by colour tissue Doppler imaging (TDI) via the measurement...

Non-Fick ian law for the neutron density current

International Nuclear Information System (INIS)

Espinosa P, G.; Vazquez R, R.; Morales S, J.

2008-01-01

In this paper, a fractional wave equation for the average neutron motion in a nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fick ian and non-Fick ian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion. The relaxation time associated with a rapid variation in the neutron flux contains an adjustable parameter, which can be manipulated to obtain the best representation of the neutron transport phenomena. (Author)

Fission neutron spectrum averaged cross sections of some threshold reactions on cadmium: production feasibility of no-carrier-added 103Pd in a nuclear reactor

International Nuclear Information System (INIS)

Abbasi, I.A.; Subhani, M.S.; Zaidi, J.H.; Arif, M.

2006-01-01

Systematic studies on fission neutron spectrum averaged cross sections of some threshold reactions like (n, p) and (n, α) on cadmium were carried out using the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy. Special attention was paid to the formation of 103 Pd via the 106 Cd(n,a α) 103 Pd reaction since it is an important therapeutic radionuclide. At a fast flux neutron density of 7.5 x 10 13 cm 2 s -1 and an irradiation time of 120 h, using 100% enriched 106 Cd target 340 MBq of no-carrier-added 103 Pd per batch could be produced. The method is thus suitable for medium-scale production of this radionuclide. (orig.)

Fission fragment yields and total kinetic energy release in neutron-induced fission of235,238U,and239Pu

Science.gov (United States)

Tovesson, F.; Duke, D.; Geppert-Kleinrath, V.; Manning, B.; Mayorov, D.; Mosby, S.; Schmitt, K.

2018-03-01

Different aspects of the nuclear fission process have been studied at Los Alamos Neutron Science Center (LANSCE) using various instruments and experimental techniques. Properties of the fragments emitted in fission have been investigated using Frisch-grid ionization chambers, a Time Projection Chamber (TPC), and the SPIDER instrument which employs the 2v-2E method. These instruments and experimental techniques have been used to determine fission product mass yields, the energy dependent total kinetic energy (TKE) release, and anisotropy in neutron-induced fission of U-235, U-238 and Pu-239.

Average nuclear surface properties

International Nuclear Information System (INIS)

Groote, H. von.

1979-01-01

The definition of the nuclear surface energy is discussed for semi-infinite matter. This definition is extended also for the case that there is a neutron gas instead of vacuum on the one side of the plane surface. The calculations were performed with the Thomas-Fermi Model of Syler and Blanchard. The parameters of the interaction of this model were determined by a least squares fit to experimental masses. The quality of this fit is discussed with respect to nuclear masses and density distributions. The average surface properties were calculated for different particle asymmetry of the nucleon-matter ranging from symmetry beyond the neutron-drip line until the system no longer can maintain the surface boundary and becomes homogeneous. The results of the calculations are incorporated in the nuclear Droplet Model which then was fitted to experimental masses. (orig.)

Self-powered neutron detector

International Nuclear Information System (INIS)

Goldstein, N.P.; Todt, W.H.

1976-01-01

A self-powered neutron detector is detailed wherein a thin conductive layer of low neutron cross section, high density material is disposed about an emitter core of material which spontaneously emits radiation on neutron capture. The high density material is absorptive of beta radiation emitted by decay of the emitter core activation product, but is substantially transmissive to the high average energy prompt electrons emitted by the emitter core material. (author)

Total body neutron activation analysis of calcium: calibration and normalisation

Energy Technology Data Exchange (ETDEWEB)

Kennedy, N S.J.; Eastell, R; Ferrington, C M; Simpson, J D; Strong, J A [Western General Hospital, Edinburgh (UK); Smith, M A; Tothill, P [Royal Infirmary, Edinburgh (UK)

1982-05-01

An irradiation system has been designed, using a neutron beam from a cyclotron, which optimises the uniformity of activation of calcium. Induced activity is measured in a scanning, shadow-shield whole-body counter. Calibration has been effected and reproducibility assessed with three different types of phantom. Corrections were derived for variations in body height, depth and fat thickness. The coefficient of variation for repeated measurements of an anthropomorphic phantom was 1.8% for an absorbed dose equivalent of 13 mSv (1.3 rem). Measurements of total body calcium in 40 normal adults were used to derive normalisation factors which predict the normal calcium in a subject of given size and age. The coefficient of variation of normalised calcium was 6.2% in men and 6.6% in women, with the demonstration of an annual loss of 1.5% after the menopause. The narrow range should make single measurements useful for diagnostic purposes.

Neutron resonance parameters of CM isotopes

International Nuclear Information System (INIS)

Belanova, T.S.; Kolesov, A.G.; Poruchikov, V.A.

1977-01-01

The total neutron cross sections of isotopes 244, 245, 246, 248 Curium have been measured on reactor CM-2 using the time-of-flight method. Single-level Breit-Wigner resonance parameters: energy E 0 , neutron width 2g GITAn, total width GITA, total neutron cross section in resonance sigma 0 have been obtained by the shape and area methods

Neutron multiplicity for neutron induced fission of 235U, 238U, and 239Pu as a function of neutron energy

International Nuclear Information System (INIS)

Zucker, M.S.; Holden, N.E.

1986-01-01

Recent development in the theory and practice of neutron correlation (''coincidence'') counting require knowledge of the higher factorial moments of the P/sub ν/ distribution (the probability that (ν) neutrons are emitted in a fission) for the case where the fission is induced by bombarding neutrons of more than thermal energies. In contrast to the situation with spontaneous and thermal neutron induced fission, where with a few exceptions the P/sub ν/ is reasonably well known, in the fast neutron energy region, almost no information is available concerning the multiplicity beyond the average value, [ν], even for the most important nuclides. The reason for this is the difficulty of such experiments, with consequent statistically poor and physically inconsistent results

Fission-fragment angular distributions and total kinetic energies for 235U(n,f) from .18 to 8.83 MeV

International Nuclear Information System (INIS)

Meadows, J.W.; Budtz-Joergensen, C.

1982-01-01

A gridded ion chamber was used to measure the fission fragment angular distribution and total kinetic energy for the 235 U(n,f) reaction from 0.18 to 8.81 MeV neutron energy. The anisotropies are in generally good agreement with earlier measurements. The average total kinetic energy is approx. 0.2 MeV greater than the thermal value at neutron energies < 2 MeV and shows a sudden decrease of approx. 0.8 MeV between 4 and 5 MeV neutron energy, well below the (n, n'f) threshold. Possible causes of this decrease are a change in the mass distribution or decreased shell effects in the heavy fragment

Development of a sealed-accelerator-tube neutron generator

Science.gov (United States)

Verbeke; Leung; Vujic

2000-10-01

Sealed-accelerator-tube neutron generators are being developed in Lawrence Berkeley National Laboratory (LBNL) for applications ranging from neutron radiography to boron neutron capture therapy and neutron activation analysis. The new generation of high-output neutron generators is based on the D-T fusion reaction, producing 14.1-MeV neutrons. The main components of the neutron tube--the ion source, the accelerator and the target--are all housed in a sealed metal container without external pumping. Thick-target neutron yield computations are performed in this paper to estimate the neutron yield of titanium and scandium targets. With an average deuteron beam current of 1 A and an energy of 120 keV, a time-averaged neutron production of approximately 10(14) n/s can be estimated for a tritiated target, for both pulsed and cw operations. In mixed deuteron/triton beam operation, a beam current of 2 A at 150 keV is required for the same neutron output. Recent experimental results on ion sources and accelerator columns are presented and discussed.

Variation estimation of the averaged cross sections in the direct and adjoint fluxes

International Nuclear Information System (INIS)

Cardoso, Carlos Eduardo Santos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

1995-01-01

There are several applications of the perturbation theory to specifics problems of reactor physics, such as nonuniform fuel burnup, nonuniform poison accumulation and evaluations of Doppler effects on reactivity. The neutron fluxes obtained from the solutions of direct and adjoint diffusion equations, are used in these applications. In the adjoint diffusion equation has been used the group constants averaged in the energy-dependent direct neutron flux, that it is not theoretically consistent. In this paper it is presented a method to calculate the energy-dependent adjoint neutron flux, to obtain the average group-constant that will be used in the adjoint diffusion equation. The method is based on the solution of the adjoint neutron balance equations, that were derived for a two regions cell. (author). 5 refs, 2 figs, 1 tab

Apparatus and method for quantitatively evaluating total fissile and total fertile nuclide content in samples

International Nuclear Information System (INIS)

Caldwell, J.T.; Cates, M.R.; Franks, L.A.; Kunz, W.E.

1985-01-01

Simultaneous photon and neutron interrogation of samples for the quantitative determination of total fissile nuclide and total fertile nuclide material present is made possible by the use of an electron accelerator. Prompt and delayed neutrons produced from resulting induced fissions are counted using a single detection system and allow the resolution of the contributions from each interrogating flux leading in turn to the quantitative determination sought. Detection limits for 239 Pu are estimated to be about 3 mg using prompt fission neutrons and about 6 mg using delayed delayed neutrons

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

SPECTER-ANL, Neutron Damage for Material Irradiation

International Nuclear Information System (INIS)

1989-01-01

1 - Description of program or function: SPECTER calculates spectral- averaged displacements, recoil spectra, gas production, and total damage energy (Kerma) for 41 pure elements using ENDF/B-V derived cross sections. The user need only specify a neutron energy spectrum. Because SPECTER does not handle compounds, SPECOMP was developed to determine displacement damage for alloys, insulators, and breeder materials. 2 - Method of solution: In SPECTER elastic scattering is treated exactly including angular distributions from ENDF/B-V. Inelastic scattering calculations consider both discrete and continuous nuclear level distributions. Multiple (n,xn) reactions use a Monte Carlo technique to derive the recoil distributions. The (n,d) and (n,t) reactions are treated as (n,p) and (n, 3 He) as (n, 4 He). The neutron-gamma reaction and subsequent beta-decay are also included, using a new treatment of gamma-gamma coincidences, angular correlations, beta-neutrino correlations and the incident neutron energy. The Lindhard model was used to compute the energy available for nuclear displacement at each recoil energy. SPECOMP reads the required files from SPECTER, computes secondary displacement functions for each combination of recoil and matrix atom, and then integrates over recoil energy to find the net displacement cross section at each neutron energy. Damage due to neutron, gamma-ray and beta decay events is then added in and the results are summed to obtain the total dpa cross section. 3 - Restrictions on the complexity of the problem: The DISCS computer code was used to process ENDF/B-V data for 41 pure elements for use with SPECTER-ANL. SPECOMP can use any combination of four elements in a single run

A Long-Pulse Spallation Source at Los Alamos: Facility description and preliminary neutronic performance for cold neutrons

International Nuclear Information System (INIS)

Russell, G.J.; Weinacht, D.J.; Pitcher, E.J.; Ferguson, P.D.

1998-03-01

The Los Alamos National Laboratory has discussed installing a new 1-MW spallation neutron target station in an existing building at the end of its 800-MeV proton linear accelerator. Because the accelerator provides pulses of protons each about 1 msec in duration, the new source would be a Long Pulse Spallation Source (LPSS). The facility would employ vertical extraction of moderators and reflectors, and horizontal extraction of the spallation target. An LPSS uses coupled moderators rather than decoupled ones. There are potential gains of about a factor of 6 to 7 in the time-averaged neutron brightness for cold-neutron production from a coupled liquid H 2 moderator compared to a decoupled one. However, these gains come at the expense of putting ''tails'' on the neutron pulses. The particulars of the neutron pulses from a moderator (e.g., energy-dependent rise times, peak intensities, pulse widths, and decay constant(s) of the tails) are crucial parameters for designing instruments and estimating their performance at an LPSS. Tungsten is the reference target material. Inconel 718 is the reference target canister and proton beam window material, with Al-6061 being the choice for the liquid H 2 moderator canister and vacuum container. A 1-MW LPSS would have world-class neutronic performance. The authors describe the proposed Los Alamos LPSS facility, and show that, for cold neutrons, the calculated time-averaged neutronic performance of a liquid H 2 moderator at the 1-MW LPSS is equivalent to about 1/4th the calculated neutronic performance of the best liquid D 2 moderator at the Institute Laue-Langevin reactor. They show that the time-averaged moderator neutronic brightness increases as the size of the moderator gets smaller

Yield of Prompt Gamma Radiation in Slow-Neutron Induced Fission of 235U as a Function of the Total Fragment Kinetic Energy

Energy Technology Data Exchange (ETDEWEB)

Albinsson, H [Chalmers Univ. of Technology, Goeteborg (SE)

1971-07-01

Fission gamma radiation yields as functions of the total fragment kinetic energy were obtained for 235U thermal-neutron induced fission. The fragments were detected with silicon surface-barrier detectors and the gamma radiation with a Nal(Tl) scintillator. In some of the measurements mass selection was used so that the gamma radiation could also be measured as a function of fragment mass. Time discrimination between the fission gammas and the prompt neutrons released in the fission process was employed to reduce the background. The gamma radiation emitted during different time intervals after the fission event was studied with the help of a collimator, the position of which was changed along the path of the fission fragments. Fission-neutron and gamma-ray data of previous experiments were used for comparisons of the yields, and estimates were made of the variation of the prompt gamma-ray energy with the total fragment kinetic energy

Remarks on some rock neutron parameters

International Nuclear Information System (INIS)

Czubek, J.A.

1983-01-01

A method to calculate the thermal neutron parameters (absorption cross-section, diffusion coefficient and diffusion length) of rocks is given. It is based on a proper energy averaging of cross-sections for all rock matrix and rock saturating liquid constituents. Special emphasis is given to the presence of hydrogen. The diffusion lengths in different lithologies in the function of the variable rock porosity have been calculated. An influence of the thermal neutron spectrum on the shape of the porosity calibration curves for the dual spacing neutron method is shown. This influence has been estimated on two porosity units, on average. (author)

Measurements of the total neutron cross-section of cerium and thulium in the energy range from 1.8 MeV to 1.8 eV

International Nuclear Information System (INIS)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Abu-Elnour, F.; Hamouda, I.

1979-01-01

Total neutron cross-section measurements have been carried out for cerium and thulium in the energy range from 1.8 meV to 1.8 eV. The measurements were performed using the time-of-flight spectrometer installed in front of one of the horizontal channels of the ET-RR-1 reactor. The obtained total neutron cross-sections were analyzed using the single level Breit-Wigner formula and the magnetic form factor. The potential scattering cross-section of Ce was found to be (3.14 +- 0.3) barns. Its coherent scattering amplitude was determined from the Bragg reflections observed in the total neutron cross-section of CeO 2 and found to be (4,8 +- 0.2) fm. The potential scattering and absorption cross-sections of Tm, at E = 0.025 eV, were found to be (7.5 +- 0.7) barns and (89.1 +- 4.1) barns respectively. (orig.) [de

Measurement of fluorine total concentration in dental enamel using fast neutron activation

International Nuclear Information System (INIS)

Mouadili, A.; Vernais, J.; Isabelle, D.B.

1988-01-01

Fluorine which is present in dental enamel, at the level of a few tens to a few hundred ppm, plays an important role in the behaviour of this tissue. Therefore quantitative determination is of interest for particular studies of the dental system. We present a nuclear nondestructive method to determine the total fluorine content in dental enamel by cyclotron-produced fast-neutron activation. The 19 F(n,2n) reaction leads to 18 F which is a β + emitter with a 109.8 min half-life. The irradiated sample activity is measured by detecting in coincidence the annihilation photons. A fluorine standard is used for calibration. The detection limit is of the order of 1 ppm, while the reproducibility is better than 95% [pt

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

Slow neutron total cross-section, transmission and reflection calculation for poly- and mono-NaCl and PbF{sub 2} crystals

Energy Technology Data Exchange (ETDEWEB)

Mansy, Muhammad S., E-mail: mmansy88@asrt.sci.eg [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); Radioactive Waste Management Unit, Hot Labs Centre, Atomic Energy Authority, Cairo (Egypt); Adib, M.; Habib, N. [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); Bashter, I.I. [Physics Department, Faculty of Science, Zagazig University (Egypt); Morcos, H.N.; El-Mesiry, M.S. [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt)

2016-10-01

Highlights: • Slow neutron cross-section calculation for poly- and mono-crystalline materials. • Monochromatic features of PbF{sub 2} and NaCl mono-crystals. • Characterization of poly- and mono-crystal filters used in neutron diffraction. • Computer code developed calculates neutron cross-section, transmission & reflection. - Abstract: A detailed study about the calculation of total neutron cross-section, transmission and reflection from crystalline materials was performed. The developed computer code is approved to be sufficient for the required calculations, also an excellent agreement has been shown when comparing the code results with the other calculated and measured values. The optimal monochromator and filter parameters were discussed in terms of crystal orientation, mosaic spread, and thickness. Calculations show that 30 cm thick of PbF{sub 2} poly-crystal is an excellent cold neutron filter producing neutron wavelengths longer than 0.66 nm needed for the investigation of magnetic structure experiments. While mono-crystal filter PbF{sub 2} cut along its (1 1 1), having mosaic spread (η = 0.5°) and thickness 10 cm can only transmit thermal neutrons of the desired wavelengths and suppress epithermal and γ-rays forming unwanted background, when it is cooled to liquid nitrogen temperature. NaCl (2 0 0) and PbF{sub 2} (1 1 1) monochromator crystals having mosaic spread (η = 0.5°) and thickness 10 mm shows high neutron reflectivity for neutron wavelengths (λ = 0.114 nm and λ = 0.43 nm) when they used as a thermal and cold neutron monochromators respectively with very low contamination from higher order reflections.

Changes in body chemical composition with age measured by total-body neutron activation

International Nuclear Information System (INIS)

Cohn, S.H.; Vaswani, A.; Zanzi, I.; Aloia, J.F.; Roginsky, M.S.; Ellis, K.J.

1976-01-01

Total-body levels of calcium and phosphorus (reflecting skeletal mass) and total-body levels of potassium (reflecting muscle mass) were measured by neutron activation analysis in 39 men and 40 women ages 30 to 90 yr. In order to intercompare the total body calcium (TBCa) values in a heterogeneous population, such as this, it was necessary to normalize the data for skeletal size. The normalization consisted of dividing the absolute calcium level by the predicted calcium level for each individual matched to a set of critical parameters. The parameter used in the computation of normal values were age, sex, muscle mass, i.e., total body potassium (TBK) and height. For the calcium data of the women, it was necessary to add an age correction factor after the age of 55 yr. The calcium ratio (mean ratio of the predicted to measured TBCa) in men was 1.000 +- 7.8 percent and in women 0.996 +- 7.1 percent. The TBCa of normal males and females can thus be predicted to +-13 percent (at the 90 percent confidence level). An exception to this was found in males (70 to 90 yr) who exhibited a mean calcium ratio greater than 1.13

Diffuse scattering of neutrons

International Nuclear Information System (INIS)

Novion, C.H. de.

1981-02-01

The use of neutron scattering to study atomic disorder in metals and alloys is described. The diffuse elastic scattering of neutrons by a perfect crystal lattice leads to a diffraction spectrum with only Bragg spreads. the existence of disorder in the crystal results in intensity and position modifications to these spreads, and above all, to the appearance of a low intensity scatter between Bragg peaks. The elastic scattering of neutrons is treated in this text, i.e. by measuring the number of scattered neutrons having the same energy as the incident neutrons. Such measurements yield information on the static disorder in the crystal and time average fluctuations in composition and atomic displacements [fr

Study of neutron fields around an intense neutron generator.

Science.gov (United States)

Kicka, L; Machrafi, R; Miller, A

2017-12-01

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

Neutron data evaluation of 232U

International Nuclear Information System (INIS)

Maslov, V.M.; Porodzinskij, Yu.V.; Tetereva, N.A.; Kagalenko, A.B.; Kornilov, N.V.; Baba, M.; Hasegawa, A.

2003-03-01

Consistent evaluation of 232 U measured data base is performed. Hauser-Feshbach- Moldauer theory, coupled channel model and double-humped fission barrier model are employed. Total, differential scattering, fission and (n,xn) data are consistently reproduced as a major constraint for inelastic scattering cross section estimate. The direct excitation of ground state and higher band levels is calculated within rigid rotator and soft (deformable) rotator model, respectively. Prompt fission neutron spectra data are described. Average resonance parameters are provided, which reproduce evaluated cross sections in the range of 10-150 keV. (author)

Neutron measuring device

International Nuclear Information System (INIS)

Hatayama, Akiyoshi; Seki, Eiji; Kita, Yoshio; Nishitani, Takeo.

1993-01-01

The device of the present invention concerns measurement for neutrons in a tokamak type thermonuclear device and it can measure total amount of generated neutrons accurately throughout the operation period even if an error is caused in counted values by plasma disruption. That is, the device comprises (1) a means for detecting presence or absence of occurrence of plasma disruption and the time for the initiation of the occurrence, (2) a first data processing means for processing detection signals, (3) a means for detecting neutrons generated in plasmas and (4) a second data processing means for calculating integrated values for the number of neutrons generated from the start to the completion of electric discharge when no disruption occurs and calculating integrated values for the number of generated neutrons from the start of electric discharge to the time at the initiation of occurrence of the disruption when disruption is present. In the thus constituted device, even if an error is caused by frequent occurrence of plasma disruption, total time integrated amount of neutrons generated in the plasmas can be measured accurately. (I.S.)

Neutron production with mixture of deuterium and krypton in Sahand Filippov type plasma focus facility

International Nuclear Information System (INIS)

Mohammadi, M.A.; Sobhanian, S.; Rawat, R.S.

2011-01-01

This Letter reports the order of magnitude enhancement in neutron yield from Sahand plasma focus device with krypton seeded deuterium operation. The highest average neutron yield of 2.2x10 9 neutrons per shot was achieved at 1.00 Torr deuterium with 3% krypton which is higher than the best average neutron yield of 3.18x10 8 neutrons per shot for pure deuterium operation. Estimation of average neutron energy showed that the maximum and minimum average energies are 2.98±0.6 MeV at 16 kV in 0.25 Torr deuterium with 3% Kr and 2.07±0.2 MeV at 18 kV operation in 0.5 Torr deuterium with 3% Kr, respectively. The anisotropy of neutron emission from Sahand DPF showed that the neutrons are produced mainly by beam-target mechanisms. -- Highlights: → The highest average neutron yield of 2.2x10 9 neutrons per shot was achieved at mixture of deuterium and krypton. → In the krypton seeding of deuterium also anisotropy of neutron emission deuterium is found. → The krypton seeding of deuterium made the neutron emission more reliable over wider operating pressure ranges.

Deformation effect in the fast neutron total cross section of alligned 59Co

International Nuclear Information System (INIS)

Fasoli, U.; Pavan, P.; Toniolo, D.; Zago, G.; Zannoni, R.; Galeazzi, G.

1982-01-01

The variation of the total neutron cross section, DELTAsigmasub(align), on 59 Co due to nuclear alignement of the target in the beam direction, has been measured over the energy range from 0.8 to 20 MeV, emploing a cobalt single-crystal with a 34% nuclear alignment degree. The results show that DELTAsigmasub(align) oscillates from a minimum of -5% at about 2.5 MeV to a maximum of +1% at about 10 MeV. The data were succesfully fitted by optical model coupled-channel calculations. The coupling terms were deduced from a model representing the 59 Co nucleus as a vibrational 60 Ni core coupled to a proton hole in a (lf7/2) shell, without free parameters. The optical model parameters were determined by fitting the total cross section independently measured. The theoretical calculations show that, at lower energies, DELTAsigmasub(align) depends appreciably on the coupling with the low-liyng levels

Deformation effect in the fast neutron total cross section of aligned 59Co

International Nuclear Information System (INIS)

Fasoli, U.; Pavan, P.; Toniolo, D.; Zago, G.; Zannoni, R.; Galeazzi, G.

1983-01-01

The variation of the total neutron cross section, Δsigma/sub align/, on 59 Co due to nuclear alignment of the target has been measured over the energy range from 0.8 to 20 MeV employing a cobalt single crystal with a 34% nuclear alignment. The results show that Δsigma/sub align/ oscillates from a minimum of -5% at about 2.5 MeV to a maximum of +1% at about 10 MeV. The data were successfully fitted by optical model coupled-channel calculations. The coupling terms were deduced from a model representing the 59 Co nucleus as a vibrational 60 Ni core coupled to a proton hole in a (1f/sub 7/2/) shell, without free parameters. The optical model parameters were determined by fitting the total cross section, which was independently measured. The theoretical calculations show that, at lower energies, Δsigma/sub align/ depends appreciably on the coupling with the low-lying levels

A long neutron optical horn for the ILL neutron-antineutron oscillation experiment

International Nuclear Information System (INIS)

Bitter, T.; Eisert, F.; El-Muzeini, P.; Kessler, M.; Klemt, E.; Lippert, W.; Meienburg, W.; Dubbers, D.

1992-01-01

In the neutron-antineutron oscillation experiment at ILL the divergence of the free flying cold neutron beam was strongly reduced without loss of intensity by the use of a 34 m long neutron-optical horn system. The divergence reduction was accurately studied in order to maintain the total width of the neutron beam below 1.1 m after a neutron free flight distance of about 80 m. The fabrication and performance of this system are described. (orig.)

Electrostatic levitation facility optimized for neutron diffraction studies of high temperature liquids at a spallation neutron source

Energy Technology Data Exchange (ETDEWEB)

Mauro, N. A., E-mail: namauro@noctrl.edu [Department of Physics, North Central College, Naperville, Illinois 60540 (United States); Vogt, A. J. [Instrument and Source Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Derendorf, K. S. [Mechanical Engineering and Materials Science, Washington University, St. Louis, Missouri 63130 (United States); Johnson, M. L.; Kelton, K. F. [Department of Physics and Institute of Materials Science and Engineering, Washington University, 1 Brookings Drive, St. Louis, Missouri 63130 (United States); Rustan, G. E.; Quirinale, D. G.; Goldman, A. I. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Kreyssig, A. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Division of Materials Sciences and Engineering, Ames Laboratory, Ames, Iowa 50011 (United States); Lokshin, K. A. [Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee 37996 (United States); Quantum Condensed Matter Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Neuefeind, J. C.; An, Ke [Chemical and Engineering Materials Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Wang, Xun-Li [Department of Physics and Materials Science, City University of Hong Kong, 83 Tat Chee Ave., Kowloon (Hong Kong); Egami, T. [Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee 37996 (United States); Department of Physics and Astronomy, Joint Institute for Neutron Sciences, University of Tennessee, Knoxville, Tennessee 37996 (United States)

2016-01-15

Neutron diffraction studies of metallic liquids provide valuable information about inherent topological and chemical ordering on multiple length scales as well as insight into dynamical processes at the level of a few atoms. However, there exist very few facilities in the world that allow such studies to be made of reactive metallic liquids in a containerless environment, and these are designed for use at reactor-based neutron sources. We present an electrostatic levitation facility, NESL (for Neutron ElectroStatic Levitator), which takes advantage of the enhanced capabilities and increased neutron flux available at spallation neutron sources (SNSs). NESL enables high quality elastic and inelastic neutron scattering experiments to be made of reactive metallic and other liquids in the equilibrium and supercooled temperature regime. The apparatus is comprised of a high vacuum chamber, external and internal neutron collimation optics, and a sample exchange mechanism that allows up to 30 samples to be processed between chamber openings. Two heating lasers allow excellent sample temperature homogeneity, even for samples approaching 500 mg, and an automated temperature control system allows isothermal measurements to be conducted for times approaching 2 h in the liquid state, with variations in the average sample temperature of less than 0.5%. To demonstrate the capabilities of the facility for elastic scattering studies of liquids, a high quality total structure factor for Zr{sub 64}Ni{sub 36} measured slightly above the liquidus temperature is presented from experiments conducted on the nanoscale-ordered materials diffractometer (NOMAD) beam line at the SNS after only 30 min of acquisition time for a small sample (∼100 mg)

Pulsed neutron activation calibration technique

International Nuclear Information System (INIS)

Kehler, P.

1979-01-01

A pulsed neutron activation (PNA) for measurement of two-phase flow consists of a pulsed source of fast neutron to activate the oxygen in a steam-water mixture. Flow is measured downstream by an NaI detector. Measured counts are sorted by a multiscaler into different time channels. A counts vs. time distribution typical for two-phase flow with slip between the two phases is obtained. Proper evaluation for the counts/time distribution leads to flow-regime independent equations for the average of the inverse transil time and the average density. After calculation of the average mass flow velocity, the true mass flow is derived

Gamma/neutron competition above the neutron separation energy in delayed neutron emitters

Directory of Open Access Journals (Sweden)

Valencia E.

2014-03-01

Full Text Available To study the β-decay properties of some well known delayed neutron emitters an experiment was performed in 2009 at the IGISOL facility (University of Jyväskylä in Finland using Total Absorption γ-ray Spectroscopy (TAGS technique. The aim of these measurements is to obtain the full β-strength distribution below the neutron separation energy (Sn and the γ/neutron competition above. This information is a key parameter in nuclear technology applications as well as in nuclear astrophysics and nuclear structure. Preliminary results of the analysis show a significant γ-branching ratio above Sn.

Study of neutron focusing at the Texas Cold Neutron Source. Final report

International Nuclear Information System (INIS)

Wehring, B.W.; Uenlue, K.

1995-01-01

Funds were received for the first year of a three year DOE Nuclear Engineering Research Grant, ''Study of Neutron Focusing at the Texas Cold Neutron Source'' (FGO2-92ER75711). The purpose of this three year study was to develop a neutron focusing system to be used with the Texas Cold Neutron Source (TCNS) to produce an intense beam of neutrons. A prompt gamma activation analysis (PGAA) facility was also to be designed, setup, and tested under the three year project. During the first year of the DOE grant, a new procedure was developed and used to design a focusing converging guide consisting of truncated rectangular cone sections. Detailed calculations were performed using a 3-D Monte Carlo code which we wrote to trace neutrons through the curved guide of the TCNS into the proposed converging guide. Using realistic reflectivities for Ni-Ti supermirrors, we obtained gains of 3 to 5 for the neutron flux averaged over an area of 1 x 1 cm

Observations on total-body calcium in humans with bone disease

International Nuclear Information System (INIS)

Spinks, T.J.; Bewley, D.K.; Ranicar, A.S.O.; Joplin, G.F.; Evans, I.M.A.; Vlotides, J.; Paolillo, M.

1979-01-01

Total-body calcium was measured in-vivo by neutron activation in a number of patients suffering from metabolic abnormalities which affect the skeleton. In general, less than 2% of total calcium resides in tissue other than bone allowing calcium mass to be directly related to skeletal mass. The conditions studied were (i) Paget's disease, treated with synthetic human calcitonin, (ii) osteoporosis, treated variously with calcium and phosphate supplements and 1,25 hydroxycholecalciferol, and (iii) Cushing's disease treated by pituitary implant of 198 Au or 90 Y seeds. The neutron beam used in these studies was produced by bombarding a beryllium target with deuterons accelerated in a cyclotron. The mean neutron energy was 7.5 MeV and patients received a total dose of 1 rem in about 30 s, a bilateral irradiation being employed. Measurements were made at approximately yearly intervals, the maximum period of study being about four and a half years. The precision of the method was estimated to be +-3% (SE) and a correction was applied for changes in body weight. In most patients, total calcium remained stable. However, in the Paget's patients, there was an indication of a slow upward trend while the osteoporotics (both treated and untreated) showed on average no change. Most of the patients with Cushing's disease showed no recovery of skeletal mass. Absolute calibration indicated that mean total body calcium in the Paget's patients was close to a predicted normal while that for the osteoporotic and Cushing's patients was 20-25% below this. (author)

Neutron flux measuring system for nuclear reactor

International Nuclear Information System (INIS)

Aoki, Kazuo.

1977-01-01

Purpose: To avoid the generation of an undesired scram signal due to abrupt changes in the neutron level given to the detectors disposed near the boundary between the moderator and the atmosphere. Constitution: In a nuclear reactor adapted to conduct power control by the change of the level in the moderator such as heavy water, the outputs from the neutron detectors disposed vertically are averaged and the nuclear reactor is scramed corresponding to the averaged value. In this system, moderator level detectors are additionally provided to the nuclear reactor and their outputs, moderator level signal, are sent to a power averaging device where the output signals of the neutron detectors are judged if they are delivered from neutrons in the moderator or not depending on the magnitude of the level signal and the outputs of the detectors out of the moderator are substantially excluded. The reactor interlock signal from the device is utilized as a scram signal. (Seki, T.)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Determination of human and Sprague-Dawley rat trimethylseleonium ion and total selenium urine concentrations from endogenous body selenium pool by neutron activation analysis

International Nuclear Information System (INIS)

Blotcky, A.J.; Claassen, J.P.; Rack, E.P.

1992-01-01

This study determined trimethylselenonium ion [TMSe,(CH 3 ) 3 Se + ] and total organic selenium cationic species urinary excretion values for healthy human subjects and Sprague-Dawley rats fed regular diets. The only source of TMSe was from the endogenous selenium body pool. Total selenium concentration in urine was determined by instrumental neutron activation analysis. TMSe and total selenium cationic species concentrations and percent of total selenium urine excretion were determined by chemical neutron activation analysis and coupled anion-cation exchange chromatography and anion-exchange chromatography, respectively. Within experimental error, mean values for TMSe and cationic species as percent selenium were comparable for both human subjects and Sprague-Dawley rats. This study suggested that TMSe excreated in urine by healthy human subjects and Sprague-Dawley rats fed a normal diet is not a minor but a general metabolite of selenium ingested in a normal diet. (author) 27 refs.; 1 fig.; 2 tabs

Compact neutron generator

Science.gov (United States)

Leung, Ka-Ngo; Lou, Tak Pui

2005-03-22

A compact neutron generator has at its outer circumference a toroidal shaped plasma chamber in which a tritium (or other) plasma is generated. A RF antenna is wrapped around the plasma chamber. A plurality of tritium ion beamlets are extracted through spaced extraction apertures of a plasma electrode on the inner surface of the toroidal plasma chamber and directed inwardly toward the center of neutron generator. The beamlets pass through spaced acceleration and focusing electrodes to a neutron generating target at the center of neutron generator. The target is typically made of titanium tubing. Water is flowed through the tubing for cooling. The beam can be pulsed rapidly to achieve ultrashort neutron bursts. The target may be moved rapidly up and down so that the average power deposited on the surface of the target may be kept at a reasonable level. The neutron generator can produce fast neutrons from a T-T reaction which can be used for luggage and cargo interrogation applications. A luggage or cargo inspection system has a pulsed T-T neutron generator or source at the center, surrounded by associated gamma detectors and other components for identifying explosives or other contraband.

The neutron production rate measurement of an indigenously developed compact D-D neutron generator

Directory of Open Access Journals (Sweden)

Das Basanta Kumar

2013-01-01

Full Text Available One electrostatic accelerator based compact neutron generator was developed. The deuterium ions generated by the ion source were accelerated by one accelerating gap after the extraction from the ion source and bombarded to a target. Two different types of targets, the drive - in titanium target and the deuteriated titanium target were used. The neutron generator was operated at the ion source discharge potential at +Ve 1 kV that generates the deuterium ion current of 200 mA at the target while accelerated through a negative potential of 80 kV in the vacuum at 1.3×10-2 Pa filled with deuterium gas. A comparative study for the neutron yield with both the targets was carried out. The neutron flux measurement was done by the bubble detectors purchased from Bubble Technology Industries. The number of bubbles formed in the detector is the direct measurement of the total energy deposited in the detector. By counting the number of bubbles the total dose was estimated. With the help of the ICRP-74 neutron flux to dose equivalent rate conversion factors and the solid angle covered by the detector, the total neutron flux was calculated. In this presentation the operation of the generator, neutron detection by bubble detector and estimation of neutron flux has been discussed.

Self-shielding coefficient and thermal flux depression factor of voluminous sample in neutron activation analysis

International Nuclear Information System (INIS)

Noorddin Ibrahim; Rosnie Akang

2009-01-01

Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)

Advanced neutron source project

International Nuclear Information System (INIS)

Gorynina, L.V.; Proskuryakov, S.F.; Tishchenko, V.A.; Uzhanova, V.V.

1991-01-01

The project of the ANS improved neutron source intended for fundamental researches in nuclear physics and materials testing is considered. New superhigh-flux heavy-water 350 MW reactor is used for the source creation. The standard fuel is uranium silicide (U 3 Si 2 ). Reactor core volume equals 67.4 l and average power density is 4.9 MW/l. Neutron flux density is 10 16 neutron/(cm 2 xs). The facility construction begin is planned for 1996. The first experiments should be accomplished in 2000

A neutron well logging system

International Nuclear Information System (INIS)

1980-01-01

A pulsed neutron well logging system using a sealed off neutron generator tube is provided with a programmable digital neutron output control system. The control system monitors the target beam current and compares a function of this current with a pre-programmed control function to develop a control signal for the neutron generator. The control signal is used in a series regulator to control the average replenisher current of the neutron generator tube. The programmable digital control system of the invention also provides digital control signals as a function of time to provide ion source voltages. This arrangement may be utilized to control neutron pulses durations and repetition rates or to produce other modulated wave forms for intensity modulating the output of the neutron generator as a function of time. (Auth.)

Approximative Krieger-Nelkin orientation averaging and anisotropy of water molecules vibrations

International Nuclear Information System (INIS)

Markovic, M.I.

1974-01-01

Quantum-mechanics approach of water molecules dynamics should be taken into account for precise theoretical calculation of differential scattering cross sections of neutrons. Krieger and Nelkin have proposed an approximate method for averaging orientation of molecules regarding directions of incoming and scattered neutron. This paper shows that this approach can be successfully applied for general shape of water molecule vibration anisotropy

Measurements and analysis of the 127I and 129I neutron capture and total cross sections

International Nuclear Information System (INIS)

Noguere, G.

2005-01-01

Most of the experimental work on the interaction of neutrons with matter has focused on materials important to reactor physics and reactor structures. By comparison, the corresponding data for minor actinides or long-lived fission products are poor. A significant demand has developed for improved neutron cross-section data of these little-studied nuclides due to the surge of interest in the transmutation of nuclear waste. With 400 kg of 129 I produced yearly in the reactors of the EU countries and a very long β - half-life of 1.57 x 10 7 years, iodine requires disposal strategies that will isolate this isotope from the environment for long periods of time. Therefore, 129 I is potentially a key long-lived fission product for transmutation applications, since 129 I transmutes in 130 I after a single neutron capture and decays to 130 Xe with a 12.36 h half-life. Accurate capture cross sections would help to reduce uncertainties in waste management concepts. For that purpose, Time-Of-Flight measurements covering the [0.5 eV-100 keV] energy range have been carried out at the 150 MeV pulsed neutron source GELINA of the Institute for Reference Materials and Measurements (IRMM). Two types of experiments have been performed at the IRMM, namely capture and transmission experiments. They are respectively related to the neutron capture and total cross sections. Since the PbI 2 samples used in this work contain natural and radioactive iodine, extensive measurements of 129 I have been carried out under the same experimental conditions as for the 129 I. The data reduction process was performed with the AGS system, and the resonance parameters were extracted with the SAMMY and REFIT shape analysis codes. In a last step, the parameters have been converted into ENDF-6 format and processed with the NJOY code to produce point-wise and multigroup cross sections, as well as MCNP and ERANOS libraries. (author)

Neutron-Star Radius from a Population of Binary Neutron Star Mergers.

Science.gov (United States)

Bose, Sukanta; Chakravarti, Kabir; Rezzolla, Luciano; Sathyaprakash, B S; Takami, Kentaro

2018-01-19

We show how gravitational-wave observations with advanced detectors of tens to several tens of neutron-star binaries can measure the neutron-star radius with an accuracy of several to a few percent, for mass and spatial distributions that are realistic, and with none of the sources located within 100 Mpc. We achieve such an accuracy by combining measurements of the total mass from the inspiral phase with those of the compactness from the postmerger oscillation frequencies. For estimating the measurement errors of these frequencies, we utilize analytical fits to postmerger numerical relativity waveforms in the time domain, obtained here for the first time, for four nuclear-physics equations of state and a couple of values for the mass. We further exploit quasiuniversal relations to derive errors in compactness from those frequencies. Measuring the average radius to well within 10% is possible for a sample of 100 binaries distributed uniformly in volume between 100 and 300 Mpc, so long as the equation of state is not too soft or the binaries are not too heavy. We also give error estimates for the Einstein Telescope.

Importance of the elemental composition in brachytherapy with neutrons; Importancia de la composicion elemental en braquiterapia con neutrones

Energy Technology Data Exchange (ETDEWEB)

Paredes G, L.; Balcazar G, M. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico); Azorin N, J. [UAM-I, 09340 Mexico D.F. (Mexico); Francois L, J.L. [ICN-UNAM, 04510 Mexico D.F. (Mexico)

2004-07-01

An analysis is presented of as the small differences that exist in the elementary composition of the wicked tumors, healthy fabrics and some material substitutes of fabric employees in dosimetry, they generate variations in the value of the kerma coefficient and consequently in the absorbed dose of neutrons in the interval 11 eV to 29 MeV. These differences make that the coefficient of kerma of neutrons average for the considered wicked tumors, be between 6% and 7% smaller that the coefficient of kerma of neutrons average for soft fabric, in the interval of interest in therapy with quick neutrons. These results have a special importance during the process of planning of brachytherapy treatments with sources of {sup 252} Cf, to optimize and to individualize the treatments to the patients. (Author)

Determination of Total Arsenic in Seaweed Products by Neutron Activation Analysis

International Nuclear Information System (INIS)

Salim, N.; Santoso, M.; Yanuar, A.; Damayanti; Kartawinata, T.G.

2013-01-01

Seaweed products are widely consumed as food nowadays. Seaweeds are known to contain arsenic due to their capability to accumulate arsenic from the environment. Arsenic is a known toxic element which naturally occurs in the environment. Ingestion of high levels of arsenic will cause several adverse health effects. Arsenic in food occurs at trace concentrations which require sensitive and selective analysis methods to perform elemental analysis on. Validated neutron activation analysis was used to determine the arsenic contents in seaweed products namely catoni from domestic product and nori from foreign products. The total arsenic concentration in the samples analyzed ranges from 0.79 mg/kg to 30.14 mg/kg with mean concentration 14.39 mg/kg. The estimated exposure to arsenic contributed by the analyzed products is from 0.07% up to 8.54% of the established provisional tolerable daily intake (PTDI) which is still far below the maximum tolerable level. (author)

Determination of Total Arsenic in Seaweed Products by Neutron Activation Analysis

Directory of Open Access Journals (Sweden)

N. Salim

2013-04-01

Full Text Available Seaweed products are widely consumed as food nowadays. Seaweeds are known to contain arsenic due to their capability to accumulate arsenic from the environment. Arsenic is a known toxic element which naturally occurs in the environment. Ingestion of high levels of arsenic will cause several adverse health effects. Arsenic in food occurs at trace concentrations which require sensitive and selective analysis methods to perform elemental analysis on. Validated neutron activation analysis was used to determine the arsenic contents in seaweed products namely catoni from domestic product and nori from foreign products. The total arsenic concentration in the samples analyzed ranges from 0.79 mg/kg to 30.14 mg/kg with mean concentration 14.39 mg/kg. The estimated exposure to arsenic contributed by the analyzed products is from 0.07% up to 8.54% of the established provisional tolerable daily intake (PTDI which is still far below the maximum tolerable level

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.)

THE SPALLATION NEUTRON SOURCE PROJECT - PHYSICAL CHALLENGES.

Energy Technology Data Exchange (ETDEWEB)

WEI,J.

2002-06-03

The Spallation Neutron Source (SNS) is designed to reach an average proton beam power of 1.4 MW for pulsed neutron production. This paper summarizes design aspects and physical challenges to the project.

Estimation of delayed neutron emission probability by using the gross theory of nuclear β-decay

International Nuclear Information System (INIS)

Tachibana, Takahiro

1999-01-01

The delayed neutron emission probabilities (P n -values) of fission products are necessary in the study of reactor physics; e.g. in the calculation of total delayed neutron yields and in the summation calculation of decay heat. In this report, the P n -values estimated by the gross theory for some fission products are compared with experiment, and it is found that, on the average, the semi-gross theory somewhat underestimates the experimental P n -values. A modification of the β-decay strength function is briefly discussed to get more reasonable P n -values. (author)

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.

Intense pulsed neutron source accelerator status

International Nuclear Information System (INIS)

Potts, C.W.; Brumwell, F.R.; Stipp, V.F.

1983-01-01

The Intense Pulsed Neutron Source (IPNS) facility has been in operation since November 1, 1981. From that date through August 1, 1983, the accelerator system was scheduled for 7191 hours of operation. During this period, 627 million pulses totaling about 1.1 x 10 21 protons were delivered to the spallation target. The accelerator has exceeded goals set in 1981 by averaging 8.65 μA over this two year period. This average beam current, while modest by the standards of proposed machines, makes the IPNS synchrotron (Rapid Cycling Synchrotron [RCS]) the highest intensity proton synchrotron in the world today. Detailed data on accelerator operation are presented. Weekly average currents of 12 μA have been achieved along with peaks of 13.9 μA. A great deal has been learned about the required operating constraints during high beam current operation. It should be possible to increase the average beam current during this next year to 12 μA while observing these restraints. Improvement plans have been formulated to increase the beam current to 16 μA over the next three years

Neutron Capture Cross Section of Unstable Ni63: Implications for Stellar Nucleosynthesis

Science.gov (United States)

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

2013-01-01

The Ni63(n,γ) cross section has been measured for the first time at the neutron time-of-flight facility n_TOF at CERN from thermal neutron energies up to 200 keV. In total, capture kernels of 12 (new) resonances were determined. Maxwellian averaged cross sections were calculated for thermal energies from kT=5-100keV with uncertainties around 20%. Stellar model calculations for a 25M⊙ star show that the new data have a significant effect on the s-process production of Cu63, Ni64, and Zn64 in massive stars, allowing stronger constraints on the Cu yields from explosive nucleosynthesis in the subsequent supernova.

Neutron reflectivity

Directory of Open Access Journals (Sweden)

Cousin Fabrice

2015-01-01

Full Text Available The specular neutron reflectivity is a technique enabling the measurement of neutron scattering length density profile perpendicular to the plane of a surface or an interface, and thereby the profile of chemical composition. The characteristic sizes that are probed range from around 5 Å up 5000 Å. It is a scattering technique that averages information on the entire surface and it is therefore not possible to obtain information within the plane of the interface. The specific properties of neutrons (possibility of tuning the contrast by isotopic substitution, sensitivity to magnetism, negligible absorption, low energy of the incident neutrons makes it particularly interesting in the fields of soft matter, biophysics and magnetic thin films. This course is a basic introduction to the technique and does not address the magnetic reflectivity. It is composed of three parts describing respectively its principle and its formalism, the experimental aspects of the method (spectrometers, samples and two examples related to the materials for energy.

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

International Nuclear Information System (INIS)

Brugger, R.M.

1975-12-01

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

Layered magnets: polarized neutron reflection studies

Energy Technology Data Exchange (ETDEWEB)

Zabel, H; Schreyer, A [Ruhr-Univ. Bochum, Lehrstuhl fuer Experimentalphysik/Festkoerperphysik, Bochum (Germany)

1996-11-01

Neutron reflectivity measurements from extended surfaces, thin films and superlattices provide information on the chemical profile parallel to the film normal, including film thicknesses, average composition and interfacial roughness parameters. Reflectivity measurements with polarized neutrons are particularly powerful for analyzing the magnetic density profiles in thin films and superlattices in addition to chemical profiles. The basic theory of polarized neutron reflectivity is provided, followed by some examples and more recent applications concerning polarized neutron reflectivity studies from exchange coupled Fe/Cr superlattices. (author) 5 figs., 13 refs.

Comparison of power pulses from homogeneous and time-average-equivalent models

International Nuclear Information System (INIS)

De, T.K.; Rouben, B.

1995-01-01

The time-average-equivalent model is an 'instantaneous' core model designed to reproduce the same three dimensional power distribution as that generated by a time-average model. However it has been found that the time-average-equivalent model gives a full-core static void reactivity about 8% smaller than the time-average or homogeneous models. To investigate the consequences of this difference in static void reactivity in time dependent calculations, simulations of the power pulse following a hypothetical large-loss-of-coolant accident were performed with a homogeneous model and compared with the power pulse from the time-average-equivalent model. The results show that there is a much smaller difference in peak dynamic reactivity than in static void reactivity between the two models. This is attributed to the fact that voiding is not complete, but also to the retardation effect of the delayed-neutron precursors on the dynamic flux shape. The difference in peak reactivity between the models is 0.06 milli-k. The power pulses are essentially the same in the two models, because the delayed-neutron fraction in the time-average-equivalent model is lower than in the homogeneous model, which compensates for the lower void reactivity in the time-average-equivalent model. (author). 1 ref., 5 tabs., 9 figs

Total Diet Study. Mg and Mn content estimation of a Market Basket of Sao Paulo state (Brazil) by Instrumental Neutron Activation

International Nuclear Information System (INIS)

Roseane Pagliaro Avegliano; Vera Akiko Maihara

2014-01-01

Total Diet Studies (TDS) have been carried out to estimate dietary intakes of the essential and toxic elements for a large-scale population over a specific period of time. In this study, the TDS was based on the evaluation of food representing a Market Basket (MB), which reflected the dietary habits of the Sao Paulo State population, corresponding to 72 % of the average food consumption for the state of Sao Paulo. In the present Total Diet Study, magnesium and manganese concentrations were determined in 30 of the most consumed food groups of a MB of Sao Paulo State, Brazil. Instrumental Neutron Activation Analysis (INAA) has been successfully used on a regularly basis in several areas of nutrition and foodstuffs. Element concentrations were determined by INAA in freeze-dried samples and ranged in mg kg -1 . Mg 41.4 (fats)-5287 (coffee) and Mn 0.12 (prime grade beef)-32.9 (coffee). The average daily Mg and Mn intake was calculated by multiplying the concentration of each element in each table-ready food group by the respective weight (g day -1 ) of the food group in the MB and adding the products from all food groups. The results of daily dietary intakes in this study were: Mg 174.8 and Mn 1.34 mg day -1 . Theses values were lower than the adequate intake (AI) proposed by the Food and Nutrition Board of the Institute of Medicine (USA National Academy) for adults. The low levels of Mg and Mn intakes presented in this TDS are probably due to the fact that MB of this study represented only 72 % of the weight of the most consumed household foods of Sao Paulo State. (author)

The total neutron cross section of 58Fe in the energy range 7 to 325 keV

International Nuclear Information System (INIS)

Hong, L.D.; Beer, H.; Kaeppeler, F.

1976-08-01

The total neutron cross section of 58 Fe has been determined in the energy range 7-325 keV by a transmission measurement using enriched 58 Fe samples. The data have been shape fitted by means of an R-matrix multi-level formalism to extract resonance parameters for s- and l > 0 wave resonances. The s-wave strength function was determined to S 0 = (4.3 +- 1.9) c 10 -4 . (orig.) [de

A measurement of ΔσL(np), the difference between neutron-proton total cross sections in pure longitudinal spin states

International Nuclear Information System (INIS)

Beddo, M.E.

1990-10-01

A measurement off Δσ L (np), the difference between neutron-proton total cross sections in pure longitudinal spin states, is described. The results will help determine the isospin-zero (I = 0) scattering amplitudes, which are not well known above laboratory energies of 500 MeV, whereas the isospin-one (I = 1) amplitudes are fairly well-determined to 1 GeV. Data points were taken at the Los Alamos Meson Physics Facility (LAMPF) at Los Alamos, New Mexico, for five neutron beam energies: 484, 568, 634,720 and 788 MeV; they are the first in this energy range. Polarized neutrons were produced by charge-exchange of polarized protons on a liquid deuterium target (LD 2 ). Large-volume neutron counters detected the neutrons that passed through a polarized proton target. The counters subtended a range of solid angles large enough to allow extrapolation of the scattered neutrons to 0 degree. Two modifications to the LAMPF accelerator system which were made for this work are described. They included a ''beam buncher,'' which modified the normal rf-time structure of the proton beam and allowed for the selection of peak-energy neutrons by time-of-flight means, and a computerized beam steering program, which reduced systematic effects due to beam motion at the LD 2 target. The experimental values of Δσ L (np) are found to be consistent with other np data, including preliminary data from SIN and Saclay, but not with some results from Argonne which used a polarized proton beam and a polarized deuteron target. The I = 0 component was extracted from Δσ L (np) using existing pp data (I = 1), with the unexpected result that Δσ L (I = 0) was found to be essentially identical in shape to Δσ L (I = 1). The significance of this is not yet understood

Thermal neutrons streaming in straight duct

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Deformation effect in the fast neutron total cross section of aligned /sup 59/Co

Energy Technology Data Exchange (ETDEWEB)

Fasoli, U.; Pavan, P.; Toniolo, D.; Zago, G.; Zannoni, R.; Galeazzi, G.

1982-07-19

The variation of the total neutron cross section, ..delta.. sigma/sub align/, on /sup 59/Co due to nuclear alignment of the target in the beam direction, has been measured over the energy range from 0.8 to 20 MeV, employing a cobalt single-crystal with a 34% nuclear alignment degree. The results show that ..delta.. sigma/sub align/ oscillates from a minimum of -5% at about 2.5 MeV to a maximum of +1% at about 10 MeV. The data were successfully fitted by optical model coupled-channel calculations. The coupling terms were deduced from a model representing the /sup 59/Co nucleus as a virbrational /sup 60/Ni core coupled to a proton-hole in a (lf 7/2) shell, without free parameters. The optical model parameters were determined by fitting the total cross-section independently measured. The theoretical calculations show that, at lower energies, ..delta.. sigma/sub align/ depends appreciably on the coupling with the low-lying levels.

Fast Neutrons - LET Distributions and the Response of Mammalian Cells

Energy Technology Data Exchange (ETDEWEB)

Bewley, D. K. [Medical Research Council Cyclotron Unit, Hammersmith Hospital, London (United Kingdom)

1968-03-15

Distributions of stopping power (LET) are given for four beams of fast neutrons, namely, neutrons of 14.6 MeV, neutrons produced by bombarding a thick beryllium target with 15 MeV deuterons, neutrons of 3 MeV, and fast neutrons produced by bombardment of a {sup 235}U converter plate with thermal neutrons. The track average LET is correlated with mean neutron energy, but the dose average is approximately constant. However, neither of these types of average is expected to have much relevance to radiobiology. Further, specification of a ''biologically effective LET'' depends on the biological test used, and is not solely a function of the radiation quality. An attempt has been made to calculate the response of T.I. kidney cells in tissue culture to these four beams of neutrons, based on their response to charged particles using the track-segment method. The calculated RBE's of the neutron beams are lower than the observed values and the calculated values of the oxygen enhancement ratio are higher. These differences seem too great to be explained by errors in dosimetry and in the calculated LET spectra. The suggestion is made that LET is not an adequate criterion of radiation quality, and that the discrepancies may be explained by more detailed consideration of the part played by delta rays and by heavy recoil tracks of short range. (author)

Neutron resonance parameters of 96Zr below 100 keV

International Nuclear Information System (INIS)

Musgrove, A.R.D.

1977-08-01

Transmission data taken at the 80 m station of the Oak Ridge Electron Linear Accelerator have provided resonance parameters for 96 Zr to 100 keV. The average level spacing and neutron strength function for s-wave neutrons were as follows: = 8 +- 2 keV and S 0 = (0.21 +- 0.10) x 10 -4 . The average p-wave neutron strength function was S 1 = (7.4 +- 2.0) x 10 -4 . (Author)

Calibration of the JET neutron yield monitors using the delayed neutron counting technique

International Nuclear Information System (INIS)

van Belle, P.; Jarvis, O.N.; Sadler, G.; de Leeuw, S.; D'Hondt, P.; Pillon, M.

1990-01-01

The time-resolved neutron yield is routinely measured on the JET tokamak using a set of fission chambers. At present, the preferred technique is to employ activation reactions to determine the neutron fluence at a well-chosen position and to relate the measured fluence to the total neutron emission by means of neutron transport calculations. The delayed neutron counting method is a particularly convenient method of performing the activation measurement and the fission cross sections are accurately known. This paper outlines the measurement technique as used on JET

Neutron and gamma-ray emission in the proton induced fission of {sup 238}U and {sup 242}Pu

Energy Technology Data Exchange (ETDEWEB)

Kniajeva, G.N.; Krupa, L.; Bogachev, A.A.; Chubarian, G.G.; Dorvaux, O.; Itkis, I.M.; Itkis, M.G.; Kliman, J.; Khlebnikov, S.; Kondratiev, N.A.; Kozulin, E.M.; Lyapin, V.; Materna, T.; Pokrovsky, I.V.; Rubchenya, V.A.; Trzaska, W.H.; Vakhtin, D.; Voskressenski, V.M

2004-04-05

Average prescission M{sup pre}{sub n} and postscission M{sup post}{sub n} neutron multiplicities as well as average {gamma}-ray multiplicity , average energy emitted by {gamma}-rays as a function of mass and total kinetic energy (TKE) of fission fragments were measured in proton induced reactions p+{sup 242}Pu{yields}{sup 243}Am, p+{sup 238}U{yields}{sup 239}Np at proton energy E{sub p}=13, 20 and 55 MeV.

Non-relativistic Free–Free Emission due to the n -distribution of Electrons—Radiative Cooling and Thermally Averaged and Total Gaunt Factors

Energy Technology Data Exchange (ETDEWEB)

De Avillez, Miguel A. [Department of Mathematics, University of Évora, R. Romão Ramalho 59, 7000 Évora (Portugal); Breitschwerdt, Dieter, E-mail: mavillez@galaxy.lca.uevora.pt [Zentrum für Astronomie und Astrophysik, Technische Universität Berlin, Hardenbergstrasse 36, D-10623 Berlin (Germany)

2017-09-01

Tracking the thermal evolution of plasmas, characterized by an n -distribution, using numerical simulations, requires the determination of the emission spectra and of the radiative losses due to free–free emission from the corresponding temperature-averaged and total Gaunt factors. Detailed calculations of the latter are presented and associated with n -distributed electrons with the parameter n ranging from 1 (corresponding to the Maxwell–Boltzmann distribution) to 100. The temperature-averaged and total Gaunt factors with decreasing n tend toward those obtained with the Maxwell–Boltzmann distribution. Radiative losses due to free–free emission in a plasma evolving under collisional ionization equilibrium conditions and composed by H, He, C, N, O, Ne, Mg, Si, S, and Fe ions, are presented. These losses decrease with a decrease in the parameter n , reaching a minimum when n  = 1, and thus converge with the loss of thermal plasma. Tables of the thermal-averaged and total Gaunt factors calculated for n -distributions, and a wide range of electron and photon energies, are presented.

Non-relativistic Free–Free Emission due to the n -distribution of Electrons—Radiative Cooling and Thermally Averaged and Total Gaunt Factors

International Nuclear Information System (INIS)

De Avillez, Miguel A.; Breitschwerdt, Dieter

2017-01-01

Tracking the thermal evolution of plasmas, characterized by an n -distribution, using numerical simulations, requires the determination of the emission spectra and of the radiative losses due to free–free emission from the corresponding temperature-averaged and total Gaunt factors. Detailed calculations of the latter are presented and associated with n -distributed electrons with the parameter n ranging from 1 (corresponding to the Maxwell–Boltzmann distribution) to 100. The temperature-averaged and total Gaunt factors with decreasing n tend toward those obtained with the Maxwell–Boltzmann distribution. Radiative losses due to free–free emission in a plasma evolving under collisional ionization equilibrium conditions and composed by H, He, C, N, O, Ne, Mg, Si, S, and Fe ions, are presented. These losses decrease with a decrease in the parameter n , reaching a minimum when n  = 1, and thus converge with the loss of thermal plasma. Tables of the thermal-averaged and total Gaunt factors calculated for n -distributions, and a wide range of electron and photon energies, are presented.

Neutron flux stabilization in the NG-150 neutron generators

International Nuclear Information System (INIS)

Kuz'min, L.E.; Makarov, S.A.; Pronman, I.M.

1986-01-01

Problem of metal tritium target lifetime increase and neutron flux stabilization in the NG-150 neutron generators is studied. Possibility on neutron flux stabilization using the mass analyzer for low-angle (4 deg and 41 deg) mass separation of a beam in thre components, which fall on a target simultaneously, is confirmed experimentally. Basic generator parameters are: accelerating voltage of 150 kV, total beam current on a target of 1.5 mA, beam current density of 0.3-1.6 mA/cm 2 , beam diameter of 8 mm. The initial neutron flux on the targets of 0.73 mg/cm 2 thick constituted 1.1x10 11 ssup(-1). The neutron flux monitoring was accomplished from recoil proton recording by a plastic scintillator. Flux decrease by more than 5% served as a signel for measuring mass analyzer magnetic field providing beam displacement on a target and restoration of the given flux. The NG-150 generator neutron flux stabilization was attained during 2h

Accelerator-based pulsed cold neutron source

International Nuclear Information System (INIS)

Inoue, Kazuhiko; Iwasa, Hirokatsu; Kiyanagi, Yoshiaki

1979-01-01

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

Measurement of the neutron total cross section of sodium

International Nuclear Information System (INIS)

Larson, D.C.; Harvey, J.A.; Hill, M.W.

1976-01-01

The transmission of neutrons through a sample of pure sodium was measured in the energy range 40 keV to 20 MeV. The measurement points out several areas for improvement in the sodium evaluation for ENDF/B-V, the most important being the broadening of the minimum at 300 keV

SB3. Experiment on secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for each of 13 different elements plus a stainless steel

International Nuclear Information System (INIS)

Maerker, R.E.

1976-01-01

The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for iron, oxygen, sodium, aluminum, copper, titanium, calcium, potassium, silicon, nickel, zinc, barium, sulfur, and a type 321 stainless steel. The gamma-ray production cross sections were binned into 0.5-MeV wide gamma-ray energy intervals. 29 tables, 1 figure

Neutron diffraction on pulsed sources

International Nuclear Information System (INIS)

Aksenov, V.L.; Balagurov, A.M.

2016-01-01

The possibilities currently offered and major scientific problems solved by time-of-flight neutron diffraction are reviewed. The reasons for the rapid development of the method over the last two decades has been mainly the emergence of third generation pulsed sources with a MW time-averaged power and advances in neutron-optical devices and detector systems. The paper discusses some historical aspects of time-of-flight neutron diffraction and examines the contribution to this method by F.L.Shapiro whose 100th birth anniversary was celebrated in 2015. The state of the art with respect to neutron sources for studies on output beams is reviewed in a special section. [ru

Measured Neutron Spectra and Dose Equivalents From a Mevion Single-Room, Passively Scattered Proton System Used for Craniospinal Irradiation

Energy Technology Data Exchange (ETDEWEB)

Howell, Rebecca M., E-mail: rhowell@mdanderson.org [Department of Radiation Physics, The University of Texas M. D. Anderson Cancer Center, Houston, Texas (United States); Burgett, Eric A.; Isaacs, Daniel [Department of Nuclear Engineering, Idaho State University, Pocatello, Idaho (United States); Price Hedrick, Samantha G.; Reilly, Michael P.; Rankine, Leith J.; Grantham, Kevin K.; Perkins, Stephanie; Klein, Eric E. [Department of Radiation Oncology, Washington University, St. Louis, Missouri (United States)

2016-05-01

Purpose: To measure, in the setting of typical passively scattered proton craniospinal irradiation (CSI) treatment, the secondary neutron spectra, and use these spectra to calculate dose equivalents for both internal and external neutrons delivered via a Mevion single-room compact proton system. Methods and Materials: Secondary neutron spectra were measured using extended-range Bonner spheres for whole brain, upper spine, and lower spine proton fields. The detector used can discriminate neutrons over the entire range of the energy spectrum encountered in proton therapy. To separately assess internally and externally generated neutrons, each of the fields was delivered with and without a phantom. Average neutron energy, total neutron fluence, and ambient dose equivalent [H* (10)] were calculated for each spectrum. Neutron dose equivalents as a function of depth were estimated by applying published neutron depth–dose data to in-air H* (10) values. Results: For CSI fields, neutron spectra were similar, with a high-energy direct neutron peak, an evaporation peak, a thermal peak, and an intermediate continuum between the evaporation and thermal peaks. Neutrons in the evaporation peak made the largest contribution to dose equivalent. Internal neutrons had a very low to negligible contribution to dose equivalent compared with external neutrons, largely attributed to the measurement location being far outside the primary proton beam. Average energies ranged from 8.6 to 14.5 MeV, whereas fluences ranged from 6.91 × 10{sup 6} to 1.04 × 10{sup 7} n/cm{sup 2}/Gy, and H* (10) ranged from 2.27 to 3.92 mSv/Gy. Conclusions: For CSI treatments delivered with a Mevion single-gantry proton therapy system, we found measured neutron dose was consistent with dose equivalents reported for CSI with other proton beamlines.

Total cross-sections assessment of neutron reaction with stainless steel SUS-310 contained in various nuclear data files

International Nuclear Information System (INIS)

Suwoto

2002-01-01

The integral testing of neutron cross-sections for Stainless Steel SUS-310 contained in various nuclear data files have been performed. The shielding benchmark calculations for Stainless Steel SUS-310 has been analysed through ORNL-Broomstick Experiment calculation which performed by MAERKER, R.E. at ORNL - USA ( 1) . Assessment with JENDL-3.1, JENDL-3.2, ENDF/B-IV, ENDF/B-VI nuclear data files and data from GEEL have also been carried out. The overall calculation results SUS-310 show in a good agreement with the experimental data, although, underestimate results appear below 3 MeV for all nuclear data files. These underestimation tendencies clearly caused by presented of iron nuclide which more than half in Stainless Steel compound. The total neutron cross-sections of iron nuclide contained in various nuclear data files relatively lower on that energy ranges

Neutronic study of spherical cold-neutron sources composed of liquid hydrogen and liquid deuterium

CERN Document Server

Matsuo, Y; Nagaya, Y

2003-01-01

Using the cross-section model for neutron scattering in liquid H sub 2 and D sub 2 , a neutron transport analysis is performed for spherical cold-neutron sources composed of either para H sub 2 , normal H sub 2 or normal D sub 2. A special effort is made to generate a set of energy-averaged cross-sections (80 group constants between 0.1 mu eV and 10 eV) for liquid H sub 2 and D sub 2 at melting and boiling points. A number of conclusions on the spherical cold-neutron source configurations are drawn. It is especially shown that the highest cold-neutron flux is obtainable from the normal D sub 2 source with a radius of about 50 cm, while the normal- and para-H sub 2 sources with radii around 3-4 cm produce maximum cold-neutron fluxes at the center.

Delayed neutrons in liquid metal spallation targets

International Nuclear Information System (INIS)

Ridikas, D.; Bokov, P.; David, J.C.; Dore, D.; Giacri, M.L.; Van Lauwe, A.; Plukiene, R.; Plukis, A.; Ignatiev, S.; Pankratov, D.

2003-01-01

The next generation spallation neutron sources, neutrino factories or RIB production facilities currently being designed and constructed around the world will increase the average proton beam power on target by a few orders of magnitude. Increased proton beam power results in target thermal hydraulic issues leading to new target designs, very often based on flowing liquid metal targets such as Hg, Pb, Pb-Bi. Radioactive nuclides produced in liquid metal targets are transported into hot cells, past electronics, into pumps with radiation sensitive components, etc. Besides the considerable amount of photon activity in the irradiated liquid metal, a significant amount of the delayed neutron precursor activity can be accumulated in the target fluid. The transit time from the front of a liquid metal target into areas, where delayed neutrons may be important, can be as short as a few seconds, well within one half-life of many delayed neutron precursors. Therefore, it is necessary to evaluate the total neutron flux (including delayed neutrons) as a function of time and determine if delayed neutrons contribute significantly to the dose rate. In this study the multi-particle transport code MCNPX combined with the material evolution program CINDER'90 will be used to evaluate the delayed neutron flux and spectra. The following scientific issues will be addressed in this paper: - Modeling of a typical geometry of the liquid metal spallation target; - Predictions of the prompt neutron fluxes, fission fragment and spallation product distributions; - Comparison of the above parameters with existing experimental data; - Time-dependent calculations of delayed neutron precursors; - Neutron flux estimates due to the prompt and delayed neutron emission; - Proposal of an experimental program to measure delayed neutron spectra from high energy spallation-fission reactions. The results of this study should be directly applicable in the design study of the European MegaPie (1 MW

Determination of the Neutron Flux in the Reactor Zones with the Strong Neutron Absorption and Leakage

International Nuclear Information System (INIS)

Ljubenov, V.; Milosevic, M.

2004-11-01

The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114. (author)

Neutron filters for producing monoenergetic neutron beams

International Nuclear Information System (INIS)

Harvey, J.A.; Hill, N.W.; Harvey, J.R.

1982-01-01

Neutron transmission measurements have been made on high-purity, highly-enriched samples of 58 Ni (99.9%), 60 Ni (99.7%), 64 Zn (97.9%) and 184 W (94.5%) to measure their neutron windows and to assess their potential usefulness for producing monoenergetic beams of intermediate energies from a reactor. Transmission measurements on the Los Alamos Sc filter (44.26 cm Sc and 1.0 cm Ti) have been made to determine the characteristics of the transmitted neutron beam and to measure the total cross section of Sc at the 2.0 keV minimum. When corrected for the Ti and impurities, a value of 0.35 +- 0.03 b was obtained for this minimum

The total neutron cross-sections of 151Eu, 153Eu and Eu below 1 eV

International Nuclear Information System (INIS)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Ashry, A.; Hamouda, I.

1980-12-01

Total neutron cross-section measurements have been carried out for natural Eu and its stable isotopes in the energy range from 3 meV-1 eV. The measurements were performed using two time-of-flight spectrometers installed in front of two of the horizontal channels of the ET-RR-1 reactor. The following results have been obtained: sigmasub(α)( 151 Eu) = 9180+-150 b at 0.0253 eV; sigmasub(α)( 153 Eu) = 375+-20 b at 0.0253 eV; sigmasub(α)(Eu) = 4600+-120 b at 0.0253 eV

The use of Total Body In Vivo Neutron Activation Analysis (TBIVNAA) in balance studies in rodents

International Nuclear Information System (INIS)

Smith, D.A.; Lindsay, R.L.; Anderson, J.

1976-01-01

In the investigation of animals subject to alteration in diet or other metabolic experiments, the measurements of change in body calcium, phosphorus, sodium and nitrogen are of considerable interest. However, conventional balance studies are tedious and subject to both random and cumulative error, necessitating as they do accurate estimates of dietary intake and faecal and urinary output. The object of the present study was to determine the usefulness of total body in vivo neutron activation analysis, used at the beginning and end of the experimental period, as an alternative to conventional balance techniques. (orig.) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-10-01

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

Measurement of natural background neutron

CERN Document Server

Li Jain, Ping; Tang Jin Hua; Tang, E S; Xie Yan Fong

1982-01-01

A high sensitive neutron monitor is described. It has an approximate counting rate of 20 cpm for natural background neutrons. The pulse amplitude resolution, sensitivity and direction dependence of the monitor were determined. This monitor has been used for natural background measurement in Beijing area. The yearly average dose is given and compared with the results of KEK and CERN.

Dose Evaluation of Neutron within Containment Building of a CE type Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Wook; Han, Jae Mun; Kim, Kyung Doek; Yun, Cheol Whan; Suh, Jang Soo; Kim, Young Jae [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

2005-03-15

From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutron radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

In vivo neutron activation facility at Brookhaven National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Ma, R.; Yasumura, Seiichi; Dilmanian, F.A.

1997-11-01

Seven important body elements, C, N, Ca, P, K, Na, and Cl, can be measured with great precision and accuracy in the in vivo neutron activation facilities at Brookhaven National Laboratory. The facilities include the delayed-gamma neutron activation, the prompt-gamma neutron activation, and the inelastic neutron scattering systems. In conjunction with measurements of total body water by the tritiated-water dilution method several body compartments can be defined from the contents of these elements, also with high precision. In particular, body fat mass is derived from total body carbon together with total body calcium and nitrogen; body protein mass is derived from total body nitrogen; extracellular fluid volume is derived from total body sodium and chlorine; lean body mass and body cell mass are derived from total body potassium; and, skeletal mass is derived from total body calcium. Thus, we suggest that neutron activation analysis may be valuable for calibrating some of the instruments routinely used in clinical studies of body composition. The instruments that would benefit from absolute calibration against neutron activation analysis are bioelectric impedance analysis, infrared interactance, transmission ultrasound, and dual energy x-ray/photon absorptiometry.

Neutron cross sections for fusion

International Nuclear Information System (INIS)

Haight, R.C.

1979-10-01

First generation fusion reactors will most likely be based on the 3 H(d,n) 4 He reaction, which produces 14-MeV neutrons. In these reactors, both the number of neutrons and the average neutron energy will be significantly higher than for fission reactors of the same power. Accurate neutron cross section data are therefore of great importance. They are needed in present conceptual designs to calculate neutron transport, energy deposition, nuclear transmutation including tritium breeding and activation, and radiation damage. They are also needed for the interpretation of radiation damage experiments, some of which use neutrons up to 40 MeV. In addition, certain diagnostic measurements of plasma experiments require nuclear cross sections. The quality of currently available data for these applications will be reviewed and current experimental programs will be outlined. The utility of nuclear models to provide these data also will be discussed. 65 references

Application of Bayesian approach to estimate average level spacing

International Nuclear Information System (INIS)

Huang Zhongfu; Zhao Zhixiang

1991-01-01

A method to estimate average level spacing from a set of resolved resonance parameters by using Bayesian approach is given. Using the information given in the distributions of both levels spacing and neutron width, the level missing in measured sample can be corrected more precisely so that better estimate for average level spacing can be obtained by this method. The calculation of s-wave resonance has been done and comparison with other work was carried out

Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Tsuneo [ed.

1992-06-15

Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10{sup {minus}5} eV to 20 MeV. Almost of the cross section data reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in order tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum.

Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

International Nuclear Information System (INIS)

Nakagawa, Tsuneo

1992-06-01

Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10 -5 eV to 20 MeV. Almost all the cross section data are reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in other tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum. (author)

Kerma factors for neutrons of 14 MeV to 60 MeV in elemental H, C, N and O

Energy Technology Data Exchange (ETDEWEB)

Behrooz, M.A.; Watt, D.E. (Dundee Univ. (UK). Dept. of Medical Biophysics)

1981-01-01

Total kerma factors, and partial kerma factors for production of specified charged H and He particles and heavier recoils, have been computed using basic theoretical and experimental nuclear data for neutrons at energies between 14 MeV and 60 MeV in the main tissue elements and in ICRU muscle tissue. All the more recent computations of total kerma factors, along with those determined from direct experimental measurements of partial kerma factors, now form a reasonably consistent set of data enabling average total kerma factors with coefficients of better than 3% for hydrogen, 16% for carbon, 23% for nitrogen and 9% for oxygen to be recommended for application to medical dosimetry and radiation protection. Total kerma factors for ICRU muscle tissue have a precision of better than 2.5% over the neutron energy range considered. Although there is adequate precision for total kerma factors for soft tissue, nevertheless analysis of the partial kerma factors indicates that caution must be exercised in use of the information for quality specification, e.g. in microdose spectra, and that more detailed basic reaction data is required for fast neutrons.

Neutron resonance parameters for 238U

International Nuclear Information System (INIS)

Poortmans, F.; Mewissen, L.; Cornelis, E.; Vanpraet, G.; Rohr, G.; Shelley, R.; Veen, T. van der; Weigmann, H.

1977-01-01

A series of total, capture and scattering cross section measurements using the neutron time-of-flight facility at the CBNM linear electron accelerator were performed. The neutron widths have been obtained for more than 400 resonances below 4.3 keV and the total capture width for 73 resonances

Studies on fission with ALADIN. Precise and simultaneous measurement of fission yields, total kinetic energy and total prompt neutron multiplicity at GSI

International Nuclear Information System (INIS)

Martin, Julie-Fiona; Taieb, Julien; Chatillon, Audrey; Belier, Gilbert; Boutoux, Guillaume; Ebran, Adeline; Gorbinet, Thomas; Grente, Lucie; Laurent, Benoit; Pellereau, Eric; Alvarez-Pol, Hector; Ayyad, Yassid; Benlliure, Jose; Cortina Gil, Dolores; Caamano, Manuel; Fernandez Dominguez, Beatriz; Paradela, Carlos; Ramos, Diego; Rodriguez-Sanchez, Jose-Luis; Vargas, Jossitt; Audouin, Laurent; Tassan-Got, Laurent; Aumann, Thomas; Casarejos, Enrique; Farget, Fanny; Rodriguez-Tajes, Carme; Heinz, Andreas; Jurado, Beatriz; Kelic-Heil, Aleksandra; Kurz, Nikolaus; Nociforo, Chiara; Pietri, Stephane; Rossi, Dominic; Schmidt, Karl-Heinz; Simon, Haik; Voss, Bernd; Weick, Helmut

2015-01-01

A novel technique for fission studies, based on the inverse kinematics approach, is presented. Following pioneering work in the nineties, the SOFIA Collaboration has designed and built an experimental set-up dedicated to the simultaneous measurement of isotopic yields, total kinetic energies and total prompt neutron multiplicities, by fully identifying both fission fragments in coincidence, for the very first time. This experiment, performed at GSI, permits to study the fission of a wide variety of fissioning systems, ranging from mercury to neptunium, possibly far from the valley of stability. A first experiment, performed in 2012, has provided a large array of unprecedented data regarding the nuclear fission process. An excerpt of the results is presented. With this solid starter, further improvements of the experimental set-up are considered, which are consistent with the expected developments at the GSI facility, in order to measure more fission observables in coincidence. The completeness reached in the SOFIA data, permits to scrutinize the correlations between the interesting features of fission, offering a very detailed insight in this still unraveled mechanism. (orig.)

Studies on fission with ALADIN. Precise and simultaneous measurement of fission yields, total kinetic energy and total prompt neutron multiplicity at GSI

Energy Technology Data Exchange (ETDEWEB)

Martin, Julie-Fiona; Taieb, Julien; Chatillon, Audrey; Belier, Gilbert; Boutoux, Guillaume; Ebran, Adeline; Gorbinet, Thomas; Grente, Lucie; Laurent, Benoit; Pellereau, Eric [CEA DAM Bruyeres-le-Chatel, Arpajon (France); Alvarez-Pol, Hector; Ayyad, Yassid; Benlliure, Jose; Cortina Gil, Dolores; Caamano, Manuel; Fernandez Dominguez, Beatriz; Paradela, Carlos; Ramos, Diego; Rodriguez-Sanchez, Jose-Luis; Vargas, Jossitt [Universidad de Santiago de Compostela, Santiago de Compostela (Spain); Audouin, Laurent; Tassan-Got, Laurent [CNRS/IN2P3, IPNO, Orsay (France); Aumann, Thomas [Technische Universitaet Darmstadt, Darmstadt (Germany); Casarejos, Enrique [Universidad de Vigo, Vigo (Spain); Farget, Fanny; Rodriguez-Tajes, Carme [CNRS/IN2P3, GANIL, Caen (France); Heinz, Andreas [Chalmers University of Technology, Gothenburg (Sweden); Jurado, Beatriz [CNRS/IN2P3, CENBG, Gradignan (France); Kelic-Heil, Aleksandra; Kurz, Nikolaus; Nociforo, Chiara; Pietri, Stephane; Rossi, Dominic; Schmidt, Karl-Heinz; Simon, Haik; Voss, Bernd; Weick, Helmut [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany)

2015-12-15

A novel technique for fission studies, based on the inverse kinematics approach, is presented. Following pioneering work in the nineties, the SOFIA Collaboration has designed and built an experimental set-up dedicated to the simultaneous measurement of isotopic yields, total kinetic energies and total prompt neutron multiplicities, by fully identifying both fission fragments in coincidence, for the very first time. This experiment, performed at GSI, permits to study the fission of a wide variety of fissioning systems, ranging from mercury to neptunium, possibly far from the valley of stability. A first experiment, performed in 2012, has provided a large array of unprecedented data regarding the nuclear fission process. An excerpt of the results is presented. With this solid starter, further improvements of the experimental set-up are considered, which are consistent with the expected developments at the GSI facility, in order to measure more fission observables in coincidence. The completeness reached in the SOFIA data, permits to scrutinize the correlations between the interesting features of fission, offering a very detailed insight in this still unraveled mechanism. (orig.)

Cold neutron radiographic apparatus and method

International Nuclear Information System (INIS)

Larsen, J.E.

1980-01-01

Cold neutron radiography may be improved by matching neutron temperature to the specific material to be analyzed. It is possible to bombard the material with neutrons having the precise average temperature necessary to realize the minimum attenuation coefficient, or to choose a neutron temperature that would increase the attenuation by inclusions, defects, etc., or to choose a neutron temperature that provides a good balance between sample transmission and defect attenuation. Other neutron temperatures might also be chosen for other reasons. This may be done by having a source of neutrons embedded in a moderator material, such as solid methane, and cooling the moderator material to the desired temperature by a cryogenic refrigerator. In another embodiment, neutrons from a nuclear reactor are passed through a moderator cooled by a cryogenic refrigerator. Since the neutron temperature is matched to the material being radiographically inspected, improved contrast and resolution can be obtained through thicker materials than it has heretofore been possible to analyze by cold neutron radiography. More optimum filtering of a neutron beam is also achieved by using a cryogenic refrigerator to cool the neutron beam filter. (auth)

Accelerator driven reactors, - the significance of the energy distribution of spallation neutrons on the neutron statistics

Energy Technology Data Exchange (ETDEWEB)

Fhager, V

2000-01-01

In order to make correct predictions of the second moment of statistical nuclear variables, such as the number of fissions and the number of thermalized neutrons, the dependence of the energy distribution of the source particles on their number should be considered. It has been pointed out recently that neglecting this number dependence in accelerator driven systems might result in bad estimates of the second moment, and this paper contains qualitative and quantitative estimates of the size of these efforts. We walk towards the requested results in two steps. First, models of the number dependent energy distributions of the neutrons that are ejected in the spallation reactions are constructed, both by simple assumptions and by extracting energy distributions of spallation neutrons from a high-energy particle transport code. Then, the second moment of nuclear variables in a sub-critical reactor, into which spallation neutrons are injected, is calculated. The results from second moment calculations using number dependent energy distributions for the source neutrons are compared to those where only the average energy distribution is used. Two physical models are employed to simulate the neutron transport in the reactor. One is analytical, treating only slowing down of neutrons by elastic scattering in the core material. For this model, equations are written down and solved for the second moment of thermalized neutrons that include the distribution of energy of the spallation neutrons. The other model utilizes Monte Carlo methods for tracking the source neutrons as they travel inside the reactor material. Fast and thermal fission reactions are considered, as well as neutron capture and elastic scattering, and the second moment of the number of fissions, the number of neutrons that leaked out of the system, etc. are calculated. Both models use a cylindrical core with a homogenous mixture of core material. Our results indicate that the number dependence of the energy

Accelerator driven reactors, - the significance of the energy distribution of spallation neutrons on the neutron statistics

International Nuclear Information System (INIS)

Fhager, V.

2000-01-01

In order to make correct predictions of the second moment of statistical nuclear variables, such as the number of fissions and the number of thermalized neutrons, the dependence of the energy distribution of the source particles on their number should be considered. It has been pointed out recently that neglecting this number dependence in accelerator driven systems might result in bad estimates of the second moment, and this paper contains qualitative and quantitative estimates of the size of these efforts. We walk towards the requested results in two steps. First, models of the number dependent energy distributions of the neutrons that are ejected in the spallation reactions are constructed, both by simple assumptions and by extracting energy distributions of spallation neutrons from a high-energy particle transport code. Then, the second moment of nuclear variables in a sub-critical reactor, into which spallation neutrons are injected, is calculated. The results from second moment calculations using number dependent energy distributions for the source neutrons are compared to those where only the average energy distribution is used. Two physical models are employed to simulate the neutron transport in the reactor. One is analytical, treating only slowing down of neutrons by elastic scattering in the core material. For this model, equations are written down and solved for the second moment of thermalized neutrons that include the distribution of energy of the spallation neutrons. The other model utilizes Monte Carlo methods for tracking the source neutrons as they travel inside the reactor material. Fast and thermal fission reactions are considered, as well as neutron capture and elastic scattering, and the second moment of the number of fissions, the number of neutrons that leaked out of the system, etc. are calculated. Both models use a cylindrical core with a homogenous mixture of core material. Our results indicate that the number dependence of the energy

Evaluation of Cm-247 neutron cross sections in the resonance region

International Nuclear Information System (INIS)

Martinelli, T.; Menapace, E.; Motta, M.; Vaccari, M.

1980-01-01

The neutron cross sections of Cm-247 are evaluated in the resonance (resolved and unresolved) region up to 10 keV. Average resonance parameters (i.e. spacing D, fission and radiative widths, neutron strength functions) are determined for unresolved region calculations. Moreover for a better comparison with the experimental data, fission cross section is calculated up to 10 MeV. In addition, the average number of neutrons emitted per fission as a function of energy is estimated

Neutron spectra characteristics for the intense neutron source, INS

International Nuclear Information System (INIS)

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

1977-01-01

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

TLD-300 detectors for separate measurement of total and gamma absorbed dose distributions of single, multiple, and moving-field neutron treatments

International Nuclear Information System (INIS)

Rassow, J.

1984-01-01

Fast neutron therapy requirements, because of the poor depth dose characteristic of present therapeutical sources, are at least as complex in treatment plans as photon therapy. The physical part of the treatment planning is very important; however, it is much more complicated than for photons or electrons owing to the need for: Separation of total and gamma absorbed dose distributions (Dsub(T) and Dsub(G)); and more stringent tissue-equivalence conditions of phantoms than in photon therapy. Therefore, methods of clinical dosimetry for the separate determination of total and gamma absorbed dose distributions in irregularly shaped (inhomogeneous) phantoms are needed. A method using TLD-300 (CaF 2 :Tm) detectors is described, which is able to give an approximate solution of the above-mentioned dosimetric requirements. The two independent doses, Dsub(T) and Dsub(G), can be calculated by an on-line computer analysis of the digitalized glow curve of TLD-300 detectors, irradiated with d(14)+Be neutrons of the cyclotron isocentric neutron therapy facility CIRCE in Essen. Results are presented for depth and lateral absorbed dose distributions (Dsub(T) and Dsub(G)) for fixed neutron beams of different field sizes compared with measurements by standard procedures (TE-TE ionization chamber, GM counter) in an A-150 phantom. The TLD-300 results for multiple and moving-field treatments (with and without wedge filters) in a patient simulating irregularly shaped (inhomogeneous) phantoms, are shown together with computer calculations of these dose distributions. The probable causes for some systematic deviations are discussed, which lead to open problems for further investigations owing to features of the detector material and the evaluation method, but mainly to differences in the composition of phantom materials used for the calculations (standard dose distributions) and TLD-300 measurements. (author)

Cosmogenic excess of 40K and the flux of fast neutrons in meteorites

International Nuclear Information System (INIS)

Stegmann, W.; Begemann, F.

1975-01-01

Results are reported of a mass spectrometric investigation of the content and isotopic composition of potassium from the mesosiderite Emery (silicate phase) and the chondrite Elenovka (bulk). Normalized to the Nier value of 39 K/ 41 K = 13.47, the 40 K in Emery (K-content 220 +- 25 ppm) was found to be enriched by (4.03 +- 0.30)%, the potassium from Elenovka (760 +- 50 ppm) to be indistinguishable from terrestrial potassium. Evidence is presented that the excess 40 K in the silicates from Emery (Ca-content 6.06 weight %) has been produced essentially by secondary cosmic ray neutrons via the 40 Ca(n,p)-reaction. The total excess of (2.57 +- 0.39) x 10 14 40 K-atoms/gCa together with the excitation function of the 40 Ca(n,p)-reaction and the neutron flux spectrum of Arnold, Honda and Lal yields a dose of fast neutrons (2 MeV 16 neutrons/cm 2 and an average flux during the cosmic ray exposure age T = (134 +- 12) Myrs of PHI = (17.4 +- 3.1) neutrons/cm 2 sec. (orig./BJ) [de

Experimental characterization of the neutron spectra generated by a high-energy clinical LINAC

Energy Technology Data Exchange (ETDEWEB)

Amgarou, K., E-mail: khalil.amgarou@uab.e [Institut de Radioprotection et de Surete Nucleaire (IRSN), Laboratoire de Metrologie et de Dosimetrie des Neutrons, F-13115 Saint Paul-Lez-Durance (France); Lacoste, V.; Martin, A. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Laboratoire de Metrologie et de Dosimetrie des Neutrons, F-13115 Saint Paul-Lez-Durance (France)

2011-02-11

The production of unwanted neutrons by electron linear accelerators (LINACs) has attracted a special attention since the early 50s. The renewed interest in this topic during the last years is due mainly to the increased use of such machines in radiotherapy. Specially, in most of developing countries where many old teletherapy irradiators, based on {sup 60}Co and {sup 137}Cs radioactive sources, are being replaced with new LINAC units. The main objective of this work is to report the results of an experimental characterization of the neutron spectra generated by a high-energy clinical LINAC. Measurements were carried out, considering four irradiation configurations, by means of our recently developed passive Bonner sphere spectrometer (BSS) using pure gold activation foils as central detectors. This system offers the possibility to measure neutrons over a wide energy range (from thermal up to a few MeV) at pulsed, intense and complex mixed n-{gamma} fields. A two-step unfolding method that combines the NUBAY and MAXED codes was applied to derive the final neutron spectra as well as their associated integral quantities (in terms of total neutron fluence and ambient dose equivalent rates) and fluence-averaged energies.

Neutron-energy-dependent cell survival and oncogenic transformation.

Science.gov (United States)

Miller, R C; Marino, S A; Martin, S G; Komatsu, K; Geard, C R; Brenner, D J; Hall, E J

1999-12-01

Both cell lethality and neoplastic transformation were assessed for C3H10T1/2 cells exposed to neutrons with energies from 0.040 to 13.7 MeV. Monoenergetic neutrons with energies from 0.23 to 13.7 MeV and two neutron energy spectra with average energies of 0.040 and 0.070 MeV were produced with a Van de Graaff accelerator at the Radiological Research Accelerator Facility (RARAF) in the Center for Radiological Research of Columbia University. For determination of relative biological effectiveness (RBE), cells were exposed to 250 kVp X rays. With exposures to 250 kVp X rays, both cell survival and radiation-induced oncogenic transformation were curvilinear. Irradiation of cells with neutrons at all energies resulted in linear responses as a function of dose for both biological endpoints. Results indicate a complex relationship between RBEm and neutron energy. For both survival and transformation, RBEm was greatest for cells exposed to 0.35 MeV neutrons. RBEm was significantly less at energies above or below 0.35 MeV. These results are consistent with microdosimetric expectation. These results are also compatible with current assessments of neutron radiation weighting factors for radiation protection purposes. Based on calculations of dose-averaged LET, 0.35 MeV neutrons have the greatest LET and therefore would be expected to be more biologically effective than neutrons of greater or lesser energies.

A system for fast neutron radiography

International Nuclear Information System (INIS)

Klann, R.T.

1996-01-01

A system has been designed and a neutron generator installed to perform fast neutron radiography. With this sytem, objects as small as a coin or as large as a waste drum can be radiographed. The neutron source is an MF Physics A-711 neutron generator which produces 3x10 10 neutrons/second with an average energy of 14.5 MeV. The radiography system uses x-ray scintillation screens and film in commercially available cassettes. The cassettes have been modified to include a thin sheet of plastic to convert neutrons to protons through elastic scattering from hydrogen and other low Z materials in the plastic. For film densities from 1.8 to 3.0, exposures range from 1.9x10 7 to 3.8x10 8 n/cm 2 depending on the type of screen and film

Polarizing neutron by light-irradiated graphene

International Nuclear Information System (INIS)

Peng, Feng

2015-01-01

We study the spin orientation of the neutron scattered by light-irradiated graphene and calculate the average value of spin z-component of the neutron in terms of a generating functional technique. Our calculation results indicate that there is a remarkable neutron polarization effect when a neutron penetrates graphene irradiated by a circularly polarized light. We analyse the dynamical source of generating this effect from the aspect of photon-mediated interaction between the neutron spin and valley pseudospin. By comparing with the polarization induced by a magnetic field, we find that this polarization may be equivalent to the one led by a magnetic field of several hundred Teslas if the photon frequency is in the X-ray frequency range. This provides an approach of polarizing neutrons. (copyright 2015 by WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Hexagonal boron nitride neutron detectors with high detection efficiencies

Science.gov (United States)

Maity, A.; Grenadier, S. J.; Li, J.; Lin, J. Y.; Jiang, H. X.

2018-01-01

Neutron detectors fabricated from 10B enriched hexagonal boron nitride (h-10BN or h-BN) epilayers have demonstrated the highest thermal neutron detection efficiency among solid-state neutron detectors to date at about 53%. In this work, photoconductive-like vertical detectors with a detection area of 1 × 1 mm2 were fabricated from 50 μm thick free-standing h-BN epilayers using Ni/Au and Ti/Al bilayers as ohmic contacts. Leakage currents, mobility-lifetime (μτ) products under UV photoexcitation, and neutron detection efficiencies have been measured for a total of 16 different device configurations. The results have unambiguously identified that detectors incorporating the Ni/Au bilayer on both surfaces as ohmic contacts and using the negatively biased top surface for neutron irradiation are the most desired device configurations. It was noted that high growth temperatures of h-10BN epilayers on sapphire substrates tend to yield a higher concentration of oxygen impurities near the bottom surface, leading to a better device performance by the chosen top surface for irradiation than by the bottom. Preferential scattering of oxygen donors tends to reduce the mobility of holes more than that of electrons, making the biasing scheme with the ability of rapidly extracting holes at the irradiated surface while leaving the electrons to travel a large average distance inside the detector at a preferred choice. When measured against a calibrated 6LiF filled micro-structured semiconductor neutron detector, it was shown that the optimized configuration has pushed the detection efficiency of h-BN neutron detectors to 58%. These detailed studies also provided a better understanding of growth-mediated impurities in h-BN epilayers and their effects on the charge collection and neutron detection efficiencies.

SORCE Level 3 Total Solar Irradiance Daily Average V016

Data.gov (United States)

National Aeronautics and Space Administration — The Total Solar Irradiance (TSI) data set SOR3TSID contains the total solar irradiance (a.k.a solar constant) data collected by the Total Irradiance Monitor (TIM)...

Total Cross-Sections of U, UO{sub 2} and ThO{sub 2} for Thermal and Subthermal Neutrons

Energy Technology Data Exchange (ETDEWEB)

Beshai, S F [IAEA-fellow from Atomic Energy Establishment (Egypt)

1966-03-15

The total neutron cross-sections of U, UO{sub 2} and ThO{sub 2} have been measured from 0.0045 eV to 0.028 eV, using the time-of-flight technique. The samples were measured at 20 deg C. ThO{sub 2} was also measured at 750 deg C. The neutron source was the reactor Rl, Stockholm. The experimental results presented as graphs in the report show in detail the influence of Bragg scattering. The results further show that the increase of the temperature for the ThO{sub 2} sample gives an increase in the cross-section. The work also contains some calculations of the position in energy of Bragg edges for the three materials. These calculations show a very good agreement with the experiments. For uranium metal some calculations have been carried out also for the height ({sigma}{sup g}{sub hkl}) of the edges The agreement with the experiments is reasonable.

Non-Fick ian law for the neutron density current; Atomos para el desarrollo de Mexico

Energy Technology Data Exchange (ETDEWEB)

Espinosa P, G.; Vazquez R, R. [UAM-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, Mexico D.F. 09340 (Mexico); Morales S, J. [UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: gepe@xanum.uam.mx

2008-07-01

In this paper, a fractional wave equation for the average neutron motion in a nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fick ian and non-Fick ian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion. The relaxation time associated with a rapid variation in the neutron flux contains an adjustable parameter, which can be manipulated to obtain the best representation of the neutron transport phenomena. (Author)

Uncertainty Assessments in Fast Neutron Activation Analysis

International Nuclear Information System (INIS)

W. D. James; R. Zeisler

2000-01-01

Fast neutron activation analysis (FNAA) carried out with the use of small accelerator-based neutron generators is routinely used for major/minor element determinations in industry, mineral and petroleum exploration, and to some extent in research. While the method shares many of the operational procedures and therefore errors inherent to conventional thermal neutron activation analysis, its unique implementation gives rise to additional specific concerns that can result in errors or increased uncertainties of measured quantities. The authors were involved in a recent effort to evaluate irreversible incorporation of oxygen into a standard reference material (SRM) by direct measurement of oxygen by FNAA. That project required determination of oxygen in bottles of the SRM stored in varying environmental conditions and a comparison of the results. We recognized the need to accurately describe the total uncertainty of the measurements to accurately characterize any differences in the resulting average concentrations. It is our intent here to discuss the breadth of potential parameters that have the potential to contribute to the random and nonrandom errors of the method and provide estimates of the magnitude of uncertainty introduced. In addition, we will discuss the steps taken in this recent FNAA project to control quality, assess the uncertainty of the measurements, and evaluate results based on the statistical reproducibility

Definition of neutron lifespan and neutron lifetime in MCNP4B

International Nuclear Information System (INIS)

Busch, R.D.; Spriggs, G.D.; Hendricks, J.S.

1997-01-01

MCNP4B was released in early 1997. In this new version, several major changes were made to the underlying theory used to estimate the non-adjoint-weighted removal, fission, capture, and escape prompt-neutron lifetimes. These four lifetimes are now being calculated in accordance to the neutron-balance theory described by Spriggs et al. in which the non-adjoint-weighted lifetime for a particular type of reaction (i.e., fission, capture, escape, removal, etc.) is defined as the total neutron population in the system divided by that reaction rate

Importance of the elemental composition in brachytherapy with neutrons

International Nuclear Information System (INIS)

Paredes G, L.; Balcazar G, M.; Azorin N, J.; Francois L, J.L.

2004-01-01

An analysis is presented of as the small differences that exist in the elementary composition of the wicked tumors, healthy fabrics and some material substitutes of fabric employees in dosimetry, they generate variations in the value of the kerma coefficient and consequently in the absorbed dose of neutrons in the interval 11 eV to 29 MeV. These differences make that the coefficient of kerma of neutrons average for the considered wicked tumors, be between 6% and 7% smaller that the coefficient of kerma of neutrons average for soft fabric, in the interval of interest in therapy with quick neutrons. These results have a special importance during the process of planning of brachytherapy treatments with sources of 252 Cf, to optimize and to individualize the treatments to the patients. (Author)

Neutron Multiplicity Analysis

Energy Technology Data Exchange (ETDEWEB)

Frame, Katherine Chiyoko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-06-28

Neutron multiplicity measurements are widely used for nondestructive assay (NDA) of special nuclear material (SNM). When combined with isotopic composition information, neutron multiplicity analysis can be used to estimate the spontaneous fission rate and leakage multiplication of SNM. When combined with isotopic information, the total mass of fissile material can also be determined. This presentation provides an overview of this technique.

Developments in Pulsed Neutron Activation for Determination of Water Flow in Pipes

CERN Document Server

Mattsson, H

2003-01-01

In PNA (pulsed neutron activation) it is important that the measured data can be related to the total mass flow. In this thesis two fundamental problems of the measurement technique and data treatment have been investigated: transport/mixing and background radiation. The principle of PNA is to introduce a radioactive substance into a pipe by bombarding fluid in the pipe with neutron pulses. The fluid in the pipe is activated and subsequently transported and mixed with the flow. Gamma radiation emitted from the activity is measured with one or two detectors downstream from the activation point. The time-resolved signal from the detectors is used to calculate the average velocity of the water flow. Due to the short distance between the neutron generator and the pipe the activity in the pipe becomes highly inhomogeneous. The transport and mixing of the activity were simulated using colour which was injected into the flow. It was found that the inhomogeneous activity distribution must be taken into account if the...

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Differential neutron spectrometry in the very low neutron energy range. Neutron cross sections for Zr, Al, polyethylene and liquid fluoropolymers

International Nuclear Information System (INIS)

Pokotilovskij, Yu.N.; Novopol'tsev, M.I.; Geltenbort, P.; Brenner, T.

2003-01-01

Some results of the test of the time-of-flight neutron spectrometers in the energy range (0.05-2.5)μeV are described. The measurements of total and differential cross sections were performed for several substances relevant to the experiments in the physics of ultracold neutrons: Zr, Al, polyethylene and liquid fluoropolymers

Neutron activation analysis at the Californium User Facility for Neutron Science

International Nuclear Information System (INIS)

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

1997-12-01

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

Failure of Stadard Optical Models to Reproduce Neutron Total Cross Section Difference in the W Isotopes

Energy Technology Data Exchange (ETDEWEB)

Anderson, J D; Bauer, R W; Dietrich, F S; Grimes, S M; Finlay, R W; Abfalterer, W P; Bateman, F B; Haight, R C; Morgan, G L; Bauge, E; Delaroche, J P; Romain, P

2001-11-01

Recently cross section differences among the isotopes{sup 182,184,186}W have been measured as part of a study of total cross sections in the 5-560 MeV energy range. These measurements show oscillations up to 150 mb between 5 and 100 MeV. Spherical and deformed phenomenological optical potentials with typical radial and isospin dependences show very small oscillations, in disagreement with the data. In a simple Ramsauer model, this discrepancy can be traced to a cancellation between radial and isospin effects. Understanding this problem requires a more detailed model that incorporates a realistic description of the neutron and proton density distributions. This has been done with results of Hartree-Fock-Bogolyubov calculations using the Gogny force, together with a microscopic folding model employing a modification of the JLM potential as an effective interaction. This treatment yields a satisfactory interpretation of the observed total cross section differences.

Polarized Epithermal Neutron Studies of Magnetic Domains

International Nuclear Information System (INIS)

Alfimenkov, V.P.; Chernikov, A.N.; Lason, L.; Mareev, Yu. D.; Novitsky, V.V.; Pikelner, L.B.; Skoy, V.R.; Tsulaya, M.I.; Gould, C.R.; Haase, D.G.; Roberson, N.R.

1997-01-01

The average size and shape of magnetic domains in a material can be determined from the precession of polarized neutrons traversing the material. Epithermal neutrons (0.5eV< En<100eV), which process more slowly than thermals, effectively probe the internal structure of samples that are thick or have large domains or large internal fields. Such epithermal neutron measurements require a neutron polarizer and analyzer based on cryogenically polarized spin filters. We discuss the measurements at JINR, Dubna, of magnetic domains in a 2.0 cm. diam. crystal of holmium using 1.7 to 59eV neutrons polarized by a dynamically polarized proton target and analyzed with a statically polarized dysprosium target

A measurement of. Delta. sigma. sub L (np), the difference between neutron-proton total cross sections in pure longitudinal spin states

Energy Technology Data Exchange (ETDEWEB)

Beddo, M.E.

1990-10-01

A measurement off {Delta}{sigma}{sub L}(np), the difference between neutron-proton total cross sections in pure longitudinal spin states, is described. The results will help determine the isospin-zero (I = 0) scattering amplitudes, which are not well known above laboratory energies of 500 MeV, whereas the isospin-one (I = 1) amplitudes are fairly well-determined to 1 GeV. Data points were taken at the Los Alamos Meson Physics Facility (LAMPF) at Los Alamos, New Mexico, for five neutron beam energies: 484, 568, 634,720 and 788 MeV; they are the first in this energy range. Polarized neutrons were produced by charge-exchange of polarized protons on a liquid deuterium target (LD{sub 2}). Large-volume neutron counters detected the neutrons that passed through a polarized proton target. The counters subtended a range of solid angles large enough to allow extrapolation of the scattered neutrons to 0{degree}. Two modifications to the LAMPF accelerator system which were made for this work are described. They included a beam buncher,'' which modified the normal rf-time structure of the proton beam and allowed for the selection of peak-energy neutrons by time-of-flight means, and a computerized beam steering program, which reduced systematic effects due to beam motion at the LD{sub 2} target. The experimental values of {Delta}{sigma}{sub L}(np) are found to be consistent with other np data, including preliminary data from SIN and Saclay, but not with some results from Argonne which used a polarized proton beam and a polarized deuteron target. The I = 0 component was extracted from {Delta}{sigma}{sub L}(np) using existing pp data (I = 1), with the unexpected result that {Delta}{sigma}{sub L}(I = 0) was found to be essentially identical in shape to {Delta}{sigma}{sub L}(I = 1). The significance of this is not yet understood.

238U subthreshold neutron induced fission cross section

International Nuclear Information System (INIS)

Difilippo, F.C.; Perez, R.B.; De Saussure, G.; Olsen, D.K.; Ingle, R.W.

1976-01-01

High resolution measurements of the 238 U neutron induced fission cross section are reported for neutron energies between 600 eV and 2 MeV. The average subthreshold fission cross section between 10 and 100 keV was found to be 44 +- 6 μb

Measurement of the average number of prompt neutrons emitted per fission of 235U relative to 252Cf for the energy region 500 eV to 10 MeV

International Nuclear Information System (INIS)

Gwin, R.; Spencer, R.R.; Ingle, R.W.; Todd, J.H.; Weaver, H.

1980-01-01

The average number of prompt neutrons emitted per fission ν/sub p/-bar(E), was measured for 235 U relative to ν/sub p/-bar for the spontaneous fission of 252 Cf over the neutron energy range from 500 eV to 10 MeV. The samples of 235 U and 252 Cf were contained in fission chambers located in the center of a large liquid scintillator. Fission neutrons were detected by the large liquid scintillator. The present values of ν/sub p/-bar(E) for 235 U are about 0.8% larger than those measured by Boldeman. In earlier work with the present system, it was noted that Boldeman's value of ν/sub p/-bar(E) for thermal energy neutrons was about 0.8% lower than obtained at ORELA. It is suggested that the thickness of the fission foil used in Boldeman's experiment may cause some of the discrepancy between his and the present values of ν/sub p/-bar(E). For the energy region up to 700 keV, the present values of ν/sub p/-bar(E) for 235 U agree, within the uncertainty, with those given in ENDF/B-V. Above 1 MeV the present results for ν/sub p/-bar(E) range about the ENDF/B-V values with differences up to 1.3%. 6 figures, 1 table

Neutrons production of very low energy by filtration through graphite and measurement of total cross sections; Production de neutrons de tres basse energie par filtration a travers du graphite et mesure de sections efficaces totales

Energy Technology Data Exchange (ETDEWEB)

Ertaud, A; Beauge, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1953-07-01

In this report, we done on the reactor of Chatillon analogous experiences to those of Fermi in Argonne. The temperature of the neutrons filtered ba a more or less thickness of graphite is measured by the method of absorption. This method permits the production and the study of a very low energy neutron beam. (M.B.) [French] Dans ce rapport, nous avons repris sur la pile de Chatillon des experiences analogues a celles de FERMI a Argonne. La temperature des neutrons filtree par une epaisseur plus ou moins grande de graphite est mesuree par la methode d'absorption. Cette methode permettant la production et l'etude d'un faisceau de neutron de tres basse energie. (M.B.)

Utilization of the intense pulsed neutron source (IPNS) at Argonne National Laboratory for neutron activation analysis

International Nuclear Information System (INIS)

Heinrich, R.R.; Greenwood, L.R.; Popek, R.J.; Schulke, A.W. Jr.

1983-01-01

The Intense Pulsed Neutron Source (IPNS) neutron scattering facility (NSF) has been investigated for its applicability to neutron activation analysis. A polyethylene insert has been added to the vertical hole VT3 which enhances the thermal neutron flux by a factor of two. The neutron spectral distribution at this position has been measured by the multiple-foil technique which utilized 28 activation reactions and the STAYSL computer code. The validity of this spectral measurement was tested by two irradiations of National Bureau of Standards SRM-1571 (orchard leaves), SRM-1575 (pine needles), and SRM-1645 (river sediment). The average thermal neutron flux for these irradiations normalized to 10 μamp proton beam is 4.0 x 10 11 n/cm 2 -s. Concentrations of nine trace elements in each of these SRMs have been determined by gamma-ray spectrometry. Agreement of measured values to certified values is demonstrated to be within experiment error

Neutron dose equivalent next to the target shield of a neutron therapy facility using an LET counter

International Nuclear Information System (INIS)

Stinchcomb, T.G.; Kuchnir, F.T.

1981-01-01

The use of a spherical tissue-equivalent proportional counter for measurements of the lineal energy (y) and derivations of the linear energy transfer (LET) for fast neutrons has the advantage of giving distributions of dose and dose equivalent as functions of either LET or y. A measurement next to the target shielding of the neutron therapy facility at the University of Chicago Hospitals and Clinics (UCHC) is described, and the data processing is outlined. The distributions are presented and compared to those from measurements in the neutron beam. The average quality factors are presented

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Measurements of neutron radiation in aircraft

International Nuclear Information System (INIS)

Vukovic, B.; Poje, M.; Varga, M.; Radolic, V.; Miklavcic, I.; Faj, D.; Stanic, D.; Planinic, J.

2010-01-01

Radiation environment is a complex mixture of charged particles of the solar and galactic origin, as well as of secondary particles created in an interaction of galactic cosmic particles with the nuclei of the Earth's atmosphere. A radiation field at aircraft altitude consists of different types of particles, mainly photons, electrons, positrons and neutrons, with a large energy range. In order to measure a neutron component of the cosmic radiation, we investigated a few combinations of a track etch detector (CR-39, LR-115) with a plastic converter or boron foil. Detector calibration was performed on neutrons coming from the nuclear reactor, as well as in the CERN-EU high-energy Reference Field (CERF) facility. From November 2007 to September 2008, the neutron dose equivalent was measured by the track detectors during five aircraft flights, in the north geographical latitude from 21 o to 58 o ; the respective average dose rate, determined by using the D-4 detector (CR-39/B), was H n =5.9 μSv/h. The photon dose rate, measured by the electronic dosimeter RAD-60 SE, had the average value of H f =1.4 μSv/h.

Measurements of neutron radiation in aircraft

Energy Technology Data Exchange (ETDEWEB)

Vukovic, B.; Poje, M.; Varga, M.; Radolic, V.; Miklavcic, I. [Department of Physics, University of Osijek, Osijek, P.O. Box 125 (Croatia); Faj, D. [Clinical Hospital Osijek (Croatia); Stanic, D. [Department of Physics, University of Osijek, Osijek, P.O. Box 125 (Croatia); Planinic, J., E-mail: planinic@ffos.h [Department of Physics, University of Osijek, Osijek, P.O. Box 125 (Croatia)

2010-12-15

Radiation environment is a complex mixture of charged particles of the solar and galactic origin, as well as of secondary particles created in an interaction of galactic cosmic particles with the nuclei of the Earth's atmosphere. A radiation field at aircraft altitude consists of different types of particles, mainly photons, electrons, positrons and neutrons, with a large energy range. In order to measure a neutron component of the cosmic radiation, we investigated a few combinations of a track etch detector (CR-39, LR-115) with a plastic converter or boron foil. Detector calibration was performed on neutrons coming from the nuclear reactor, as well as in the CERN-EU high-energy Reference Field (CERF) facility. From November 2007 to September 2008, the neutron dose equivalent was measured by the track detectors during five aircraft flights, in the north geographical latitude from 21{sup o} to 58{sup o}; the respective average dose rate, determined by using the D-4 detector (CR-39/B), was H{sub n}=5.9 {mu}Sv/h. The photon dose rate, measured by the electronic dosimeter RAD-60 SE, had the average value of H{sub f}=1.4 {mu}Sv/h.

Neutron transmission through pyrolytic graphite crystals

Energy Technology Data Exchange (ETDEWEB)

Adib, M. [Reactor Physics Department NRC, Reactor Physics Division, Nuclear Research Center, Egyptian Atomic Energy Authority, Cairo 13759 (Egypt); Habib, N. [Reactor Physics Department NRC, Reactor Physics Division, Nuclear Research Center, Egyptian Atomic Energy Authority, Cairo 13759 (Egypt)]. E-mail: nadiahabib15@yahoo.com; Fathaalla, M. [Reactor Physics Department NRC, Reactor Physics Division, Nuclear Research Center, Egyptian Atomic Energy Authority, Cairo 13759 (Egypt)

2006-05-15

Calculation of the total cross-section, neutron transmission and removal coefficient of pyrolytic graphite (PG) for thermal neutron energies were carried out using an additive formula. The formula takes into account the variation of thermal diffuse and Bragg scattering cross-sections in terms of PG temperature and mosaic spread for neutron energies in the range 1 meV to 1 eV. A computer code PG has been developed which allow calculations for the graphite in its hexagonal close-packed structure, when its c-direction is parallel with incident neutron beam (parallel orientation). The calculated total neutron cross-sections for PG in parallel orientation at different mosaic spreads were compared with the measured values. An overall agreement is indicated between the formula fits and experimental data at room and liquid nitrogen temperatures. A feasibility study for use of PG crystals as second-order neutron filter is detailed in terms of mosaic spread, optimum thickness and temperature. The calculated removal coefficients of PG crystals show that such crystals are high efficiency second-order filter within neutron energy intervals (4-7 meV) and (10-15 meV)

Neutron shielding properties of a borated high-density glass

Directory of Open Access Journals (Sweden)

Saeed Aly Abdallah

2017-01-01

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

Design of neutron diagnostic for MTX

International Nuclear Information System (INIS)

Ogawa, Toshihide; Oasa, Kazumi; Hoshino, Katsumichi; Odajima, Kazuo; Maeda, Hikosuke

1990-07-01

A neutron diagnostic system was designed for the Microwave Tokamak Experiment being carried out at the lawrence Livermore National Laboratory. High speed measurements are important to this experiment. Plastic scintillator is used for this fast response detection of neutron. Proportional counters and fission counters are used for the total neutron emission rate measurements. (author)

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.

Large area solid target neutron source

International Nuclear Information System (INIS)

Crawford, J.C.; Bauer, W.

1974-01-01

A potentially useful neutron source may result from the combination of a solid deuterium-tritium loaded target with the large area, high energy ion beams from ion sources being developed for neutral beam injection. The resulting neutron source would have a large radiating area and thus produce the sizable experimental volume necessary for future studies of bulk and synergistic surface radiation effects as well as experiments on engineering samples and small components. With a 200 keV D + T + beam and 40 kW/cm 2 power dissipation on a 200 cm 2 target spot, a total neutron yield of about 4 x 10 15 n/sec may be achieved. Although the useable neutron flux from this source is limited to 1 to 2 x 10 13 n/cm 2 /sec, this flux can be produced 3 cm in front of the target and over about 300 cm 3 of experimental volume. Problems of total power dissipation, sputtering, isotopic flushing and thermal dissociation are reviewed. Neutron flux profiles and potential experimental configurations are presented and compared to other neutron source concepts. (U.S.)

Contributions to the theory of fission neutron emission

International Nuclear Information System (INIS)

Seeliger, D.; Maerten, H.; Ruben, A.

1990-03-01

This report gives a compilation of recent work performed at Technical University, Dresden by D. Seeliger, H. Maerten and A. Ruben on the topic of fission neutron emission. In the first paper calculated fission neutron spectra are presented using the temperature distribution model FINESSE for fissioning actinide nuclei. In the second paper, starting from a general energy balance, Terrell's approach is generalized to describe average fragment energies as a function of incident energy; trends of fragment energy data in the Th-Pu region are well reproduced. In the third contribution, prompt fission neutron spectra and fragment characteristics for spontaneous fission of even Pu-isotopes are presented and discussed in comparison with experimental data using a phenomenological scission point model including temperature dependent shell effects. In the fourth paper, neutron multiplicities and energy spectra as well as average fragment energies for incident energies from threshold to 20 MeV (including multiple-chance fission) for U-238 are compared with traditional data representations. (author). Refs, figs and tabs

The study of iodine in Chinese total diets

International Nuclear Information System (INIS)

Hou, Xiaolin; Chai, Chifang; Qian, Qinfang; Liu, Guodong; Zhang, Yongbao; Wang, Ke

1997-01-01

In this work, China was divided into four area groups according to their geographical positions and dietary habits. All foods were divided into 12 types and the iodine contents in various diets were determined using epithermal neutron activation analysis (NAA). The intakes for China were evaluated. The results indicate that the intakes of iodine in northern areas are slightly higher and in south areas lower than the lowest recommended intake, and the average intake in China is 166 μg/person per day, which is within the recommended range. In addition, one province was chosen from each area groups. The dietary intakes of iodine were investigated in different ages and sex using total mixed diet method. Our results indicate that the average iodine intake of four provinces was lower than the recommended value, which suggests that it is necessary to supplement iodine in foods in China

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.)

An application of reactor noise techniques to neutron transport problems in a random medium

International Nuclear Information System (INIS)

Sahni, D.C.

1989-01-01

Neutron transport problems in a random medium are considered by defining a joint Markov process describing the fluctuations of one neutron population and the random changes in the medium. Backward Chapman-Kolmogorov equations are derived which yield an adjoint transport equation for the average neutron density. It is shown that this average density also satisfied the direct transport equation as given by the phenomenological model. (author)

Neutron emission in heavy ion induced reactions at 10 MeV/A

International Nuclear Information System (INIS)

Benrachi, F.

1984-01-01

The neutron emission mechanism in reactions induced by heavy ions (in which a high energy component had already been observed) has been investigated. To get informations on the desexcitation mode which is responsible of that component and on the sharing of the excitation energy between the fragments, a very disymmetric system 14 N+ 165 Ho at E inc=10,5 MeV/a.m.u. was studied. The neutrons are preferentially emitted forward in the direction of the quasi-projectile and are in average highly energetic. The calculations of the excitation energies and of the multiplicities showed that the detected events are mainly coming from peripheral processes with a weak angular momentum exchange. The analysis of the energy spectra and of the invariance cross section maps pointed out a strong asymmetry and then proved that at least two neutron sources were involved in that reaction mechanism. Studying these two components in terms of a sequential mechanism i.e. an emission from the totally equilibrated quasi-target and quasi-projectile, the whole neutron emission could not be explained. To interpret the experimental results, non statistical models with a preequilibrium emission at the collision beginning were used [fr

Neutron scattering from 12C in the few-MeV region

International Nuclear Information System (INIS)

Smith, A.; Holt, R.; Whalen, J.

1978-09-01

Neutron total cross sections of natural carbon are deduced from the observed transmission of approximately monoenergetic neutrons through carbon samples of varying thickness. The measurements extend from approximately equal to 0.1 to 4.5 MeV with resolutions of approximately equal to 2 to 100 keV. Neutron differential-elastic-scattering cross sections of natural carbon are measured from 1.5 to 4.0 MeV at incident-neutron energy intervals of approximately less than 100 keV, over an angular range of approximately equal to 20 to 160 0 and with energy resolutions of 20 to 50 keV. The experimental results are interpreted in terms of a multilevel R-function analysis. Results are compared with the large body of measured and evaluated neutron total and scattering cross sections and scattered neutron polarizations reported in the literature. It is suggested that the observed neutron total and scattering cross sections of carbon are physically consistent and suitable for use as a reference standard in experimental studies of neutron processes. The R-function description should provide a convenient description of neutron total and scattering cross sections of carbon as a function of both angle and energy. 88 references

Fast neutron flux and intracranial dose distribution at a neutron irradiation facility

International Nuclear Information System (INIS)

Matsumoto, Tetsuo; Aizawa, Otohiko; Nozaki, Tetsuya

1981-01-01

A head phantom filled with water was used to measure the fast neutron flux using 115 In(n, n')sup(115m)In and 103 Rh(n, n')sup(103m)Rh reactions. γ-ray from sup(115m)In and x-ray from sup(103m)Rh were detected by a Ge(Li) and a Na(Tl)I counter, respectively. TLD was used to investigate the γ-dose rate distribution inside the phantom. Flux of fast neutron inside the phantom was about 1 x 10 6 n/cm 2 sec, which was 3 order smaller than that of thermal neutron. The fast neutron flux decreased to 1/10 at 15 cm depth, and γ-dose rate was about 200 R/h at 100 kW inside the phantom. Total dose at the surface was 350 rad/h, to which, fast neutrons contributed more than γ-rays. The rate of fast neutron dose was about 10% of thermal neutron's in Kerma dose unit (rad), however, the rate was highly dependent on RBE value. (Nakanishi, T.)

Single Crystal Filters for Neutron Spectrometry

International Nuclear Information System (INIS)

Habib, N.

2008-01-01

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

Impurity radiation from a beam-plasma neutron source

International Nuclear Information System (INIS)

Molvik, A.W.

1995-01-01

Impurity radiation, in a worst case evaluation for a beam-plasma neutron source (BPNS), does not limit performance. Impurities originate from four sources: (a) sputtering from walls by charge exchange or alpha particle bombardment, (b) sputtering from limiters, (c) plasma desorption of gas from walls and (d) injection with neutral beams. Sources (c) and (d) are negligible; adsorbed gas on the walls of the confinement chamber and the neutral beam sources is removed by the steady state discharge. Source (b) is negligible for impinging ion energies below the sputtering threshold (T i ≤ 0.025 keV on tungsten) and for power densities to the limiter within the capabilities of water cooling (30-40 MW/m 2 ); both conditions can be satisfied in the BPNS. Source (a) radiates 0.025 MW/m 2 to the neutron irradiation samples, compared with 5 to 10 MW/m 2 of neutrons; and radiates a total of 0.08 MW from the plasma column, compared with 60 MW of injected power. The particle bombardment that yields source (a) deposits an average of 2.7 MW/m 2 on the samples, within the capabilities of helium gas cooling (10 MW/m 2 ). An additional worst case for source (d) is evaluated for present day 2 to 5 s pulsed neutral beams with 0.1% impurity density and is benchmarked against 2XIIB. The total radiation would increase a factor of 1.5 to ≤ 0.12 MW, supporting the conclusion that impurities will not have a significant impact on a BPN. (author). 61 refs, 7 figs, 2 tabs

Importance of isovector effects in reproducing neutron total cross section differences in the W isotopes

International Nuclear Information System (INIS)

Dietrich, F.S.; Anderson, J.D.; Bauer, R.W.; Grimes, S.M.; Finlay, R.W.; Abfalterer, W.P.; Bateman, F.B.; Haight, R.C.; Morgan, G.L.; Bauge, E.; Delaroche, J.-P.; Romain, P.

2003-01-01

Cross section differences among the isotopes 182,184,186 W have been measured as a part of a study of total cross sections in the 5-560 MeV energy range. These difference measurements show oscillations up to 150 mb between 5 and 100 MeV. Calculations with spherical and deformed phenomenological optical potentials employing standard radial and isospin dependences show much smaller oscillations than the experimental data. In a simple Ramsauer model, this discrepancy can be traced to a cancellation between radial and isospin effects. Understanding this problem requires a more detailed model that incorporates a realistic description of the neutron and proton density distributions. This has been done with the results of Hartree-Fock-Bogoliubov calculations using the Gogny force, together with a microscopic folding model employing a modification of the Jeukenne, Lejeune, and Mahaux potential as an effective interaction. This treatment yields a satisfactory interpretation of the observed total cross section differences up to 200 MeV. The calculations have been extended above that energy with a folding model based on an empirical effective interaction

Pulsed neutron source and instruments at neutron facility

Energy Technology Data Exchange (ETDEWEB)

Teshigawara, Makoto; Aizawa, Kazuya; Suzuki, Jun-ichi; Morii, Yukio; Watanabe, Noboru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-11-01

We report the results of design studies on the optimal target shape, target - moderator coupling, optimal layout of moderators, and neutron instruments for a next generation pulsed spallation source in JAERI. The source utilizes a projected high-intensity proton accelerator (linac: 1.5 GeV, {approx}8 MW in total beam power, compressor ring: {approx}5 MW). We discuss the target neutronics, moderators and their layout. The sources is designed to have at least 30 beam lines equipped with more than 40 instruments, which are selected tentatively to the present knowledge. (author)

Polarized epithermal neutron studies of magnetic domains

International Nuclear Information System (INIS)

Alfimenkov, V.P.; Chernikov, A.N.; Lason, L.; Mareev, Y.D.; Novitsky, V.V.; Pikelner, L.B.; Skoy, V.R.; Tsulaya, M.I.; Gould, C.R.; Haase, D.G.; the Triangle Universities Nuclear Laboratory, Durham, North Carolina; Roberson, N.R.; the Triangle Universities Nuclear Laboratory, Durham, North Carolina

1997-01-01

The average size and shape of magnetic domains in a material can be determined from the precession of polarized neutrons traversing the material. Epithermal neutrons (0.5eV n <100eV), which precess more slowly than thermals, effectively probe the internal structure of samples that are thick or have large domains or large internal fields. Such epithermal neutron measurements require a neutron polarizer and analyzer based on cryogenically polarized spin filters. We discuss the measurement at JINR, Dubna, of magnetic domains in a 2.0 cm. diam. crystal of holmium using 1.7 to 59 eV neutrons polarized by a dynamically polarized proton target and analyzed with a statically polarized dysprosium target. copyright 1997 American Institute of Physics

Two-phase flow measurement by pulsed neutron activation techniques

International Nuclear Information System (INIS)

Kehler, P.

1978-01-01

The Pulsed Neutron Activation (PNA) technique for measuring the mass flow velocity and the average density of two-phase mixtures is described. PNA equipment can be easily installed at different loops, and PNA techniques are non-intrusive and independent of flow regimes. These features of the PNA technique make it suitable for in-situ measurement of two-phase flows, and for calibration of more conventional two-phase flow measurement devices. Analytic relations governing the various PNA methods are derived. The equipment and procedures used in the first air-water flow measurement by PNA techniques are discussed, and recommendations are made for improvement of future tests. In the present test, the mass flow velocity was determined with an accuracy of 2 percent, and average densities were measured down to 0.08 g/cm 3 with an accuracy of 0.04 g/cm 3 . Both the accuracy of the mass flow velocity measurement and the lower limit of the density measurement are functions of the injected activity and of the total number of counts. By using a stronger neutron source and a larger number of detectors, the measurable density can be decreased by a factor of 12 to .007 g/cm 3 for 12.5 cm pipes, and to even lower ranges for larger pipes

Validation of evaluated neutron standard cross sections

International Nuclear Information System (INIS)

Badikov, S.; Golashvili, T.

2008-01-01

Some steps of the validation and verification of the new version of the evaluated neutron standard cross sections were carried out. In particular: -) the evaluated covariance data was checked for physical consistency, -) energy-dependent evaluated cross-sections were tested in most important neutron benchmark field - 252 Cf spontaneous fission neutron field, -) a procedure of folding differential standard neutron data in group representation for preparation of specialized libraries of the neutron standards was verified. The results of the validation and verification of the neutron standards can be summarized as follows: a) the covariance data of the evaluated neutron standards is physically consistent since all the covariance matrices of the evaluated cross sections are positive definite, b) the 252 Cf spectrum averaged standard cross-sections are in agreement with the evaluated integral data (except for 197 Au(n,γ) reaction), c) a procedure of folding differential standard neutron data in group representation was tested, as a result a specialized library of neutron standards in the ABBN 28-group structure was prepared for use in reactor applications. (authors)

Testing of ENDF/B cross section data in the Californium-252 neutron benchmark field

International Nuclear Information System (INIS)

Mannhart, W.

1979-01-01

The fission neutron field of 252 Cf presently represents one of the most well-known neutron benchmark fields. For 13 neutron reactions which are of importance in reactor metrology, measurements of spectrum-averaged cross sections, [sigma], performed in this neutron field were compared with calculated average cross sections. This comparison allows one to draw conclusions as to the quality of different sigma(E) data taken from ENDF/B-IV, from ENDF/B-V, and from recent experiments and used in the calculation of average cross sections. The comparison includes an uncertainty analysis regarding the different uncertainty contributions of [sigma], of sigma(E), and of the spectral distribution of 252 Cf fission neutrons. Additionally, in a few examples, sensitivity studies were carried out. The sensitivity of the spectrum-averaged cross sections to individual characteristics of the sigma(E) data, such as normalization factors or shifts in the energy scale, was investigated. Similarly, the sensitivity of [sigma] to the spectral distribution of 252 Cf was determined. 4 figures, 2 tables

High energy neutron generator

International Nuclear Information System (INIS)

Barjon, R.; Breynat, G.

1987-01-01

This patent describes a generator of fast neutrons only slightly contaminated by neutrons of energy less than 15 MeV, comprising a source of charged particles of energy equal to at least 15 MeV, a target made of lithium deuteride, and means for cooling the target. The target comprises at least two elements placed in series in the path of the charged particles and separated from each other, the thickness of each of the elements being selected as a function of the average energy of the charged particles emitted from the source and the energy of the fast neutrons to be generated such that neutrons of energy equal to at least 15 MeV are emitted in the forward direction in response to the bombardment of the target from behind by the charged particles. The target cooling means comprises means for circulating between and around the elements a gas which does not chemically react with lithium deuteride

PNG-300 a nanosecond pulsed neutron generator

International Nuclear Information System (INIS)

Sztaricskai, T.; Vasvary, L.; Petoe, G.C.; Devkin, B.V.

1985-01-01

The design and operation of a nanosecond-pulse neutron generator is reported. It was constructed for the measurement of prompt neutron and gamma radiation in experimental studies of fast neutron reactions by time of flight techniques. The acceleration voltage is 300 kV and the total resolution of the generator-neutron spectrometer system is 2 ns. The ion-optical system, the vacuum system and the control of the neutron generator is described in detail. The equipment was used for prompt neutron and gamma radiation induced in construction materials. (R.P.)

Experimental determination of one- and two-neutron separation energies for neutron-rich copper isotopes

Science.gov (United States)

Yu, Mian; Wei, Hui-Ling; Song, Yi-Dan; Ma, Chun-Wang

2017-09-01

A method is proposed to determine the one-neutron S n or two-neutron S 2n separation energy of neutron-rich isotopes. Relationships between S n (S 2n) and isotopic cross sections have been deduced from an empirical formula, i.e., the cross section of an isotope exponentially depends on the average binding energy per nucleon B/A. The proposed relationships have been verified using the neutron-rich copper isotopes measured in the 64A MeV 86Kr + 9Be reaction. S n, S 2n, and B/A for the very neutron-rich 77,78,79Cu isotopes are determined from the proposed correlations. It is also proposed that the correlations between S n, S 2n and isotopic cross sections can be used to find the location of neutron drip line isotopes. Supported by Program for Science and Technology Innovation Talents at Universities of Henan Province (13HASTIT046), Natural and Science Foundation in Henan Province (162300410179), Program for the Excellent Youth at Henan Normal University (154100510007) and Y-D Song thanks the support from the Creative Experimental Project of National Undergraduate Students (CEPNU 201510476017)

Atomic structure of radiation damages in FCC-metals after neutron irradiation

International Nuclear Information System (INIS)

Popova, E.V.; Ivchenko, V.A.; Kozlov, A.V.

2005-01-01

Full text: Radiation clusters, formed at a neutron irradiation, are a product of evolution of cascade areas. The quantitative information about clusters can be used for verification of calculations of cascade damage ability, in particular, cascade efficiency. Data about concentration clusters and an average of the vacancies containing in them, allow to receive total of the vacancies reserved in them and to use them for comparison to results of calculations of cascade damage ability. A correctness of such comparison by that above, than below temperature of a neutron irradiation. The purpose of work was experimental studying radiation clusters formed in FCC-metals at a low temperature neutron irradiation methods of dilatometry, field ion (FIM) and transmission electronic microscopy (TEM). Radiation clusters were studied: in industrial austenite steel C0.05Crl6Nil5Mo2Mnl, irradiated in reactor Rw-2a at temperature 310 K up to fluence intermediate and fast neutrons (with E > 0,1 MeV) 6.7·l0 21 m -2 ; in a modelling material - Pt (cleanliness of 99.99 %) with the same - FCC-structure in an initial condition and after an irradiation in reactor RWW-2M at temperature 310 K up to fluence intermediate and fast neutrons (with E > 0.1 MeV) 3.5·10 22 m -2 . As a result of an irradiation of steel and pure Pt, in these materials by methods FIM and TEM many radiation clusters, the accelerated neutrons initiated by interaction with substance was revealed. It is established that these damage areas represent the depleted zones containing separate vacancies, and also small vacancy complexes, with the 'belt' interstitial atoms. The quantitative estimation of the sizes of such radiating defects is lead and their density in volume is experimentally established. So the neutron irradiation of steel at temperature 310 K up to fluence 6.7·10 21 m -2 causes formation radiation clusters which average diameter according to TEM makes 3 nanometers. Observable by methods FIM clusters have the

Energy corrections in pulsed neutron measurements for cylindrical geometry

International Nuclear Information System (INIS)

Drozdowicz, K.; Woznicka, U.

1982-01-01

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

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

The neutron lifetime experiment PENeLOPE

Energy Technology Data Exchange (ETDEWEB)

Schreyer, Wolfgang [Technische Universitaet Muenchen (Germany); Collaboration: PENeLOPE-Collaboration

2015-07-01

The neutron lifetime τ{sub n}=880.3±1.1 s is an important parameter in the Standard Model of particle physics and in Big Bang cosmology. Several systematic corrections of previously published results reduced the PDG world average by several σ in the last years and call for a new experiment with complementary systematics. The experiment PENeLOPE, currently under construction at the Physik-Department of Technische Universitaet Muenchen, aims to determine the neutron lifetime with a precision of 0.1 s. It will trap ultra-cold neutrons in a magneto-gravitational trap using a large superconducting magnet and will measure their lifetime by both neutron counting and online proton detection. This presentation gives an overview over the latest developments of the experiment.

Delayed neutron kinetic functions for /sup 232/Th and /sup 238/U mixtures

Energy Technology Data Exchange (ETDEWEB)

Ganich, P P; Goshovskij, M V; Lendel, A I; Lomonosov, V I; Sikora, D I; Sychev, S I

1984-11-01

In order to investigate the applicability of the method based on using kinetic functions, describing the emission of delayed neutrons by samples for determination of the content of fissionable nuclides in binary mixtures, the /sup 232/Th+/sup 238/U mixtures have been analyzed with the M-30 microtron. Fresh samples containing ThO/sub 2/, U/sub 3/O/sub 8/ and their mixtures are irradiated by bremstrahlung at the 15.5 MeV energy of accelerated electrons and 9 ..mu..A average current. The mass of samples is about 6 g. To determine the kinetic functions, temporal distributions of delayed neutron pulses are used, their maximum number for different samples being (1.7-3.0) x 10/sup 4/. In processing the data obtained two methods of normalization of the delayed neutron number in the kinetic functions are used: to the total yield of delayed neutrons and to the yield of /sup 133/I ..gamma..-quanta. The conclusion is drawn that the method investigated permits to determine relative /sup 238/U concentrations in the mixtures considered with 0.06-0.2 errors. Error reduction is achieved during the normalization of the number of delayed neutrons to the yield of /sup 130/I ..gamma..-quanta.

High count problems in elemental analysis using pulsed neutron inelastic scattering

Energy Technology Data Exchange (ETDEWEB)

Vartsky, D; Wielopolski, L; Ellis, K J; Cohn, S H [Brookhaven National Lab., Upton, NY (USA). Medical Dept.

1983-03-01

Elemental analysis by neutron inelastic scattering using a miniature intense pulsed neutron source ('Zetatron') was evaluated. The particular problems associated with detector pulse-pile-up during the neutron burst and the limited ability of the analyzer to process on average more than one detector pulse per neutron burst were examined. The severity of these problems is described and a solution using a multiple ADC system is proposed.

Detailed characterisation of the incident neutron beam on the TOSCA spectrometer

Science.gov (United States)

Pinna, Roberto S.; Rudić, Svemir; Capstick, Matthew J.; McPhail, David J.; Pooley, Daniel E.; Howells, Gareth D.; Gorini, Giuseppe; Fernandez-Alonso, Felix

2017-10-01

We report a detailed characterisation of the incident neutron beam on the TOSCA spectrometer. A bespoke time-of-flight neutron monitor has been designed, constructed and used to perform extensive spatially resolved measurements of the absolute neutron flux and its underlying time structure at the instrument sample position. The obtained data give a quantitative understanding of the current instrument beyond neutronic simulations and provide a baseline in order to assess the performance of the upgraded instrument. At an average proton current-on-target of 153 μA (ISIS Target Station 1; at the time of measurements) we have found that the wavelength-integrated neutron flux (from 0.28 Å to 4.65 Å) at the position of the TOSCA instrument sample (spatially averaged across the 3 × 3cm2 surface centred around (0,0) position) is approximately 1 . 2 × 106 neutrons cm-2s-1, while the whole beam has a homogeneous distribution across the 3 . 0 × 3 . 5cm2 sample surface. The spectra reproduced the well-known shape of the neutrons moderated by the room temperature water moderator and exhibit a neutron flux of 7 . 3 × 105 neutrons cm-2s-1Å-1 at 1 Å.

An attempt of application of short lived 44K activity induced in the 44Ca(n,p)44K reaction using 14 MeV neutrons for total body calcium assessment in human subject

International Nuclear Information System (INIS)

Haratym, Z.; Kempisty, T.; Mikolajewski, S.; Rurarz, E.

1999-01-01

The status of in vivo neutron activation analysis techniques for the measurement of total body calcium in human subject is reviewed. Relevant data on the nuclear characteristics of calcium isotopes during interaction with neutrons ranging from slow up to 14 MeV neutrons are presented. Physical aspects of the measurement of in vivo total body calcium (TBCa) using 44 K activity induced in the 44 Ca(n,p) 44 K(T 1/2 =22.3 min) reaction by 14 MeV neutrons are discussed. The measurement of delayed γ-ray emitted during decay of activities induced in enriched 44 Ca, nat Ca, phantom filled with water solution of natural calcium and skeletal arm are considered. Results of measurements on the phantom and skeletal arm indicate a possibility to measure the TBCa using the 44 K activity. (author)

Lead-Bismuth-Eutectic Spallation Neutron Source for Nuclear Transmuter

International Nuclear Information System (INIS)

Gohar, Y.; Herceg, J.; Krajtl, L.; Micklich, B.; Pointer, D.; Saiveau, J.; Sofu, T.; Finck, P.

2002-01-01

A lead-bismuth eutectic (LBE) spallation target design concept has been developed for the subcritical multiplier (SCM) design of the accelerator-driven test facility (ADTF). The design is based on a coaxial geometrical configuration, which has been carefully analyzed and designed to achieve an optimum performance. The target design description, the results from the parametric studies, and the design analyses including neutronics, heat transfer, and hydraulics analyses are given in this paper. A detailed MCNPX geometrical model for the target has been developed to generate heating rates and nuclear responses in the structural material for the design process. The beam has a uniform distribution of 600 MeV protons and 5-MW total power. A small LBE buffer is optimized to reduce the irradiation damage in the SCM fuel elements from the scatter protons and the high-energy neutrons, to maximize the neutron yield to the SCM operation, and to provide inlet and outlet manifolds for the LBE coolant. A special attention has been given to the target window design to enhance its lifetime. The window volumetric heating is 766 W/cm 3 relative to 750 W/cm 3 in LBE for a 40-μA/cm 2 current density. The results show that the nuclear heating from the proton beam diminishes at about 32 cm along the beam axis in the LBE target material. The neutron contribution to the atomic displacement is in the range of 94 to ∼100% for the structure material outside the proton beam path. In the beam window, the neutron contribution is ∼74% and the proton beam is responsible for more than 95% of the total gas production. The proton contribution to the gas production vanishes outside the beam path. The LBE average velocity is ∼2 m/s. The heat transfer and the hydraulics analyses have been iterated to reduce the maximum temperature and the thermal stress level in the target window to enhance its operating life. (authors)

LANL sunnyside experiment: Study of neutron production in accelerator-driven targets

International Nuclear Information System (INIS)

Morgan, G.; Butler, G.; Cappiello, M.; Carius, S.; Daemen, L.; DeVolder, B.; Frehaut, J.; Goulding, C.; Grace, R.; Green, R.; Lisowski, P.; Littleton, P.; King, J.; King, N.; Prael, R.; Stratton, T.; Turner, S.; Ullmann, J.; Venneri, F.; Yates, M.

1995-01-01

Measurements have been made of the neutron production in prototypic targets for accelerator driven systems. Studies were conducted on four target assemblies containing lead, lithium, tungsten, and a thorium-salt mixture. Integral data on total neutron production were obtained as well as more differential data on neutron leakage and neutron flux profiles in the blanket/moderator region. Data analysis on total neutron production is complete and shows excellent agreement with calculations using the LAHET/MCNP code system

LANL sunnyside experiment: Study of neutron production in accelerator-driven targets

Energy Technology Data Exchange (ETDEWEB)

Morgan, G.; Butler, G.; Cappiello, M. [Los Alamos National Laboratory, NM (United States)] [and others

1995-10-01

Measurements have been made of the neutron production in prototypic targets for accelerator driven systems. Studies were conducted on four target assemblies containing lead, lithium, tungsten, and a thorium-salt mixture. Integral data on total neutron production were obtained as well as more differential data on neutron leakage and neutron flux profiles in the blanket/moderator region. Data analysis on total neutron production is complete and shows excellent agreement with calculations using the LAHET/MCNP code system.

Prompt neutrons from {sup 236}U fission fragments

Energy Technology Data Exchange (ETDEWEB)

Boldeman, J W; Musgrove, A.R. de L.; Walsch, R L

1971-03-01

Measurements were made of prompt neutron emission in the thermal neutron fission of {sup 235}U. The mean neutron emission per fragment was obtained for particular values of the fragment mass and total kinetic energy. A direct neutron counting method was employed and a comparison made with data from previous experiments of this type. (author)

Neutron total cross-section of hydrogenous and deuterated 1- and 2-propanol and n-butanol measured using the VESUVIO spectrometer

Science.gov (United States)

Rodríguez Palomino, L. A.; Dawidowski, J.; Márquez Damián, J. I.; Cuello, G. J.; Romanelli, G.; Krzystyniak, M.

2017-10-01

This work presents the total cross sections of a set of normal and deuterated alcohols (hydrogenous 1- and 2-propanol and n-butanol, 1-propanol(OD) and fully deuterated 2-propanol and n-butanol), measured at spectrometer VESUVIO (ISIS spallation neutron source, United Kingdom). Granada's Synthetic Model was applied to describe those systems and a satisfactory agreement with the measured total cross section was achieved in the range of energies from 10-3 to 100 eV. The input parameters of the model were determined from the essential features of the vibrational spectra of the atoms that compose the systems, which were studied using Molecular Dynamics.

Multiscale simulation of neutron induced damage in tritium breeding blankets with different spectral shifters

Energy Technology Data Exchange (ETDEWEB)

Choi, Yong Hee; Joo, Han Gyu, E-mail: joohan@snu.ac.kr

2013-10-15

Highlights: • A multiscale defect simulation system tailored for neutron damage estimation is introduced. • The new recoil spectrum code can use the most recent ENDF-B/VII nuclear data. • The high energy cascades are broken into subcascades using the INCAS model. • OKMC simulation provides data for shear stress estimation using dislocation dynamics formula. • Demonstration is made with a fusion blanket design having different spectral shifters. -- Abstract: A multiscale material defect simulation established to evaluate neutron induced damages on metals is applied to an estimation of material degradation in helium cooled molten lithium blankets in which four different spectral shifter materials are examined as a means of maximizing the tritium breeding ratio through proper shaping of the neutron spectrum. The multiscale system consists of a Monte Carlo neutron transport code, a recoil spectrum generation code, a molecular dynamics code, a high energy cascade breakup model, an object kinetic Monte Carlo code, and a simple formula as the shear stress estimator. The average recoil energy of the primary knock-on atoms, the total concentration of the defects, average defect sizes, and the increase in shear stress after a certain irradiation time are calculated for each spectral shifter. Among the four proposed materials of B4C, Be, Graphite and TiC, B4C reveals the best shielding performance in terms of neutron radiation hardening. The result for the increase in shear stress after 100 days of irradiation indicates that the increased shear stress is 1.5 GPa for B4C which is about 40% less than that of the worst one, the graphite spectral shifter. The other damage indicators show consistent trends.

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.)

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)

Effects of fuel particle size distributions on neutron transport in stochastic media

International Nuclear Information System (INIS)

Liang, Chao; Pavlou, Andrew T.; Ji, Wei

2014-01-01

Highlights: • Effects of fuel particle size distributions on neutron transport are evaluated. • Neutron channeling is identified as the fundamental reason for the effects. • The effects are noticeable at low packing and low optical thickness systems. • Unit cells of realistic reactor designs are studied for different size particles. • Fuel particle size distribution effects are not negligible in realistic designs. - Abstract: This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (k eff ) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron

Inhomogeneity of neutron and gamma-ray attenuation in biological shields

Energy Technology Data Exchange (ETDEWEB)

El-bakkoush, F A; El-Ghobary, A M; Megahid, R M [Reactor and Neutron physics Department, Nuclear Research Center, A.E.A., Cairo (Egypt)

1997-12-31

Measurements have been carried-out to investigate the attenuation properties of some materials which are used as biological shields around nuclear radiation sources. Investigation was performed by measuring the transmitted fast neutron and gamma-spectra through cylindrical samples of magnetite- limonite, steel and cellulose shields. The neutron and gamma spectra were measured by a neutron-gamma spectrometer with stilbene scintillator. Discrimination between neutron and gamma pulses was achieved by a discrimination method. The obtained results are displayed in the form of neutron and gamma spectra and attenuation relations which are used to derive the total macroscopic cross-sections for neutrons and total linear attenuation coefficients for gamma-rays. The values of neutron and gamma relaxation lengths are also derived for the investigated materials. 10 figs., 1 tabs.

[Summary talk for the fourth international symposium on neutron induced reactions

International Nuclear Information System (INIS)

Raman, S.

1985-01-01

Comments are made on presentations on these topics: how the mean field theories account for average nuclear properties determined by the dynamics of the system and how the fluctuations about this average can be simulated in terms of a Hamiltonian consisting of a Gaussian orthogonal ensemble; a full Hamiltonian consisting of at least nine terms; multi-step direct reaction theory; the verification of the hypothesis of an E1 giant resonance built on every excited state of the nucleus; neutron resonance averaging; level density of compound resonances; unequal parity distribution to explain the gamma competition in the continuum of 56 Fe; a survey of the most recent measurements of (n,p) and n,α) reactions with thermal and resonance neutrons; and data indicating a hard component of fission neutrons from 252 Cf

Tritium and neutron measurements from deuterated Pd-Si

International Nuclear Information System (INIS)

Claytor, T.N.; Tuggle, D.G.; Menlove, H.O.; Seeger, P.A.; Doty, W.R.; Rohwer, R.K.

1990-01-01

Evidence has been found for tritium and neutron production in palladium and silicon stacks when pulsed with a high electric current. These palladium-silicon stacks consist of alternating layers of pressed palladium and silicon powder. A pulsed high electric current is thought to promote non equilibrium conditions important for tritium and neutron production. More than 2000 hours of neutron counting time has been accumulated in a underground, low background, environment with high efficiency counters (21%). Neutron emission has occurred as infrequent burst or as low level emission lasting for up to 20 hours. In eight of 30 cells, excess tritium greater than 3 sigma has been observed. In each of these measurements, with the powder system, the ratio of tritium detected to total integrated total neutrons inferred has been anomalously high. Recent cells have shown reproducible tritium generation at a level of about 0.5 nCi/hr. Several hydrogen and air control cells have been run with no anomalous excess tritium or neutron emission above background. A significant amount of the total palladium inventory (18%) has been checked for tritium contamination by three independent means. 12 refs., 6 figs., 2 tabs

Residual insufficiency of hematopoiesis after acute or chronic exposure to gamma radiation or neutrons

International Nuclear Information System (INIS)

Wangenheim, K.H. v.; Peterson, H.P.; Feinendegen, L.E.

1983-01-01

Recovery of the stem cell quality is possible after acute exposure to 500 rad γ radiation up to a period of 6 months. Beyond this data, a significant residual damage remains. The same applies to quantitative stem cell recovery. Chronic γ exposure leads to less radiation damage than acute exposure. After a total accumulation of 500 rad, the proliferation factors after chronic exposure were, on an average 20% higher than after acute radiation exposure. 6 MeV neutron exposure reduced the stem cell quality and stem cell count much more efficiently than γ exposure. The relative biological effect of neutrons is at least 2.5 times as high as the γ effect, both for the stem cell count and the stem cell quality. (orig.) [de

Monitor for reactor neutron detector

International Nuclear Information System (INIS)

Shirakami, Hisayuki; Shibata, Masatoshi

1992-01-01

The device of the present invention judges as to whether a neutron detector is normal or not while considering the change of indication value depending on the power change of a reactor core. That is, the device of the present invention comprises a standard value setting device for setting the standard value for calibrating the neutron detector and an abnormality judging device for comparing the standard value with a measured value of the neutron detector and judging the abnormality when the difference is greater than a predetermined value. The measured value upon initialization of each of the neutron detectors is determined as a quasi-standard value. An average value of the difference between the measured value and the quasi-standard value of a plurality of effective neutron detectors at a same level for the height of the reactor core is multiplied to a power rate based on the reactor core power at a position where the neutron detector is disposed upon calibration. The value obtained by adding the multiplied value and the quasi-standard value is determined as a standard value. The abnormality judging device compares the standard value with the measured value of the neutron detector and, if the difference is greater than a predetermined value, the neutron detector is determined as abnormal. As a result, judgement can be conducted more accurately than conventional cases. (I.S.)

Superintensive pulse slow neutron source SIN based on kaon factory

International Nuclear Information System (INIS)

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

1991-01-01

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

Neutron Measurements At Hanford's Plutonium Finishing Plant

International Nuclear Information System (INIS)

Conrady, Matthew M.; Berg, Randal K.; Scherpelz, Robert I.; Rathbone, Bruce A.

2009-01-01

The Pacific Northwest National Laboratory (PNNL) conducted neutron measurements at Hanford's Plutonium Finishing Plant (PFP). The measurements were performed to evaluate the performance of the Hanford Standard Dosimeter (HSD) and the 8816 TLD component of the Hanford Combination Neutron Dosimeter (HCND) in the neutron fields responsible for worker neutron exposures. For this study, TEPC detectors and multisphere spectrometers were used to measure neutron dose equivalent rate, and multispheres were used to measure average neutron energy. Water-filled phantoms holding Hanford dosimeters were positioned at each measurement location. The phantoms were positioned in the same location where a multisphere measurement was taken and TEPCs were also positioned there. Plant survey meters were also used to measure neutron dose rates at all locations. Three measurement locations were chose near the HC-9B glovebox in room 228A of Building 234-5. The multisphere spectrometers measured average neutron energies in the range of 337 to 555 keV at these locations. Personal dose equivalent, Hp(10)n, as measured by the multisphere and TEPC, ranged from 2.7 to 9.7 mrem/h in the three locations. Effective dose assuming a rotational geometry (EROT) was substantially lower than Hp(10), ranging from 1.3 to 3.6 mrem/h. These values were lower than the reported values from dosimeters exposed on a rotating phantom. Effective dose assuming an AP geometry (EAP) was also substantially lower than Hp(10), ranging from 2.3 to 6.5 mrem/h. These values were lower than the reported values from the dosimeters on slab phantoms. Since the effective dose values were lower than reported values from dosimeters, the dosimeters were shown to be conservative estimates of the protection quantities.

Attenuation of thermal neutrons by an imperfect single crystal

Energy Technology Data Exchange (ETDEWEB)

Naguib, K.; Adib, M. [National Research Centre, Cairo (Egypt). Reactor and Neutron Physics Dept.

1996-06-14

A semi-empirical formula is given which allows one to calculate the total thermal cross section of an imperfect single crystal as a function of crystal constants, temperature and neutron energy E, in the energy range between 3 meV and 10 eV. The formula also includes the contribution of the parasitic Bragg scattering to the total cross section that takes into account the crystal mosaic spread value and its orientation with respect to the neutron beam direction. A computer program (ISCANF) was developed to calculate the total attenuation of neutrons using the proposed formula. The ISCANF program was applied to investigate the neutron attenuation through a copper single crystal. The calculated values of the neutron transmission through the imperfect copper single crystal were fitted to the measured ones in the energy range 3-40 meV at different crystal orientations. The result of fitting shows that use of the computer program ISCANF allows one to predict the behaviour of the total cross section of an imperfect copper single crystal for the whole energy range. (author).

Attenuation of thermal neutrons by an imperfect single crystal

Science.gov (United States)

Naguib, K.; Adib, M.

1996-06-01

A semi-empirical formula is given which allows one to calculate the total thermal cross section of an imperfect single crystal as a function of crystal constants, temperature and neutron energy E, in the energy range between 3 meV and 10 eV. The formula also includes the contribution of the parasitic Bragg scattering to the total cross section that takes into account the crystal mosaic spread value and its orientation with respect to the neutron beam direction. A computer program (ISCANF) was developed to calculate the total attenuation of neutrons using the proposed formula. The ISCANF program was applied to investigate the neutron attenuation through a copper single crystal. The calculated values of the neutron transmission through the imperfect copper single crystal were fitted to the measured ones in the energy range 3 - 40 meV at different crystal orientations. The result of fitting shows that use of the computer program ISCANF allows one to predict the behaviour of the total cross section of an imperfect copper single crystal for the whole energy range.

Optimal timing of neutron irradiation for boron neutron capture therapy after intravenous infusion of sodium borocaptate in patients with glioblastoma

International Nuclear Information System (INIS)

Kageji, Teruyoshi; Nagahiro, Shinji; Kitamura, Katsushi; Nakagawa, Yoshinobu; Hatanaka, Hiroshi; Haritz, Dietrich; Grochulla, Frank; Haselsberger, Klaus; Gabel, Detlef

2001-01-01

Purpose: A cooperative study in Europe and Japan was conducted to determine the pharmacokinetics and boron uptake of sodium borocaptate (BSH: Na 2 B 12 H 11 SH), which has been introduced clinically as a boron carrier for boron neutron capture therapy in patients with glioblastoma. Methods and Materials: Data from 56 patients with glioblastoma who received BSH intravenous infusion were retrospectively reviewed. The pharmacokinetics were evaluated in 50 patients, and boron uptake was investigated in 47 patients. Patients received BSH doses between 12 and 100 mg/kg of body weight. For the evaluation, the infused boron dose was scaled linearly to 100 mg/kg BSH. Results: In BSH pharmacokinetics, the average value for total body clearance, distribution volume of steady state, and mean residence time was 3.6±1.5 L/h, 223.3±160.7 L, and 68.0±52.5 h, respectively. The average values of the boron concentration in tumor adjusted to 100 mg/kg BSH, the boron concentration in blood adjusted to 100 mg/kg BSH, and the tumor/blood boron concentration ratio were 37.1±35.8 ppm, 35.2±41.8 ppm, and 1.53±1.43, respectively. A good correlation was found between the logarithmic value of T adj and the interval from BSH infusion to tumor tissue sampling. About 12-19 h after infusion, the actual values for T adj and tumor/blood boron concentration ratio were 46.2±36.0 ppm and 1.70±1.06, respectively. The dose ratio between tumor and healthy tissue peaked in the same interval. Conclusion: For boron neutron capture therapy using BSH administered by intravenous infusion, this work confirms that neutron irradiation is optimal around 12-19 h after the infusion is started

Attenuation of Thermal Neutrons by Crystalline Silicon

International Nuclear Information System (INIS)

Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

2002-01-01

A simple formula is given which allows to calculate the contribution of the total neutron cross - section including the Bragg scattering from different (hkt) planes to the neutron * transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy .A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500μ eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

Summary report of the consultants' meeting on neutron sources spectra for EXFOR

International Nuclear Information System (INIS)

Simakov, S.P.; Kaeppeler, F.

2011-10-01

The participants highlighted the importance of complementing the averaged cross section data already stored in EXFOR by the incident neutron energy spectra. They shared their experience on measurement and simulation of neutron fields produced at reactors and accelerators over a wide energy range. The source characteristics, format and rules needed for storage in EXFOR were discussed. The participants submitted the numerical information on spectra that will essentially increase the number of 'complete' data sets in EXFOR. The report additionally provides an overview of (i) neutron production cross sections and thick target yields missing from the EXFOR database; (ii) codes for neutron spectra calculations; (iii) informational resources for reactor, radioactive and spallation neutron sources; (iv) codes for spectrum unfolding and (v) EXFOR compilation rules for the Maxwellian averaged cross sections measured for the reactor and astrophysical applications. (author)

Measurement of cross sections of threshold detectors with spectrum average technique

International Nuclear Information System (INIS)

Agus, Y.; Celenk, I.; Oezmen, A.

2004-01-01

Cross sections of the reactions 103 Rh(n, n') 103m Rh, 115 In(n, n') 115m In, 232 Th(n, f), 47 Ti(n, p) 47 Sc, 64 Zn(n, p) 64 Cu, 58 Ni(n, p) 58 Co, 54 Fe(n, p) 54 Mn, 46 Ti(n, p) 46 Sc, 27 Al(n, p) 27 Mg, 56 Fe(n, p) 56 Mn, 24 Mg(n, p) 24 Na, 59 Co(n, α) 56 Mn, 27 Al(n, α) 24 Na and 48 Ti(n, p) 48 Sc were measured with average neutron energies above effective threshold by using the activation method through usage of spectrum average technique in an irradiation system where there are three equivalent Am/Be sources, each of which has 592 GBq activity. The cross sections were determined with reference to the fast neutron fission cross section of 238 U. The measured values and published values are generally in agreement. (orig.)

A method to compare calculated and experimental 14 MeV neutron attenuation coefficient and to determine the total removal cross-section

International Nuclear Information System (INIS)

Elay, A.G.

1978-01-01

A method to compare calculated and experimental neutron attenuation coefficients (chi) when samples are o, different geometries but the same material is proposed. The best Σ (total removal cross section) is determined by using the fact that the logarithm of the attenuation coefficient varies linearly with respect to Σ i.e. lg chi = + asub(s) Σ, where asub(s) is a parameter that characterises all the geometrical experimental conditions of the neutron source, the sample and the relative source-to-sample geometry. In order to increase the precision, samples of different geometries but the same material were used. Values of chi are determined experimentally and asub(s) calculated for these geometries. The graph of lg chi as a function of asub(s) together with a simple fit to a straight line is sufficient to determine Σ (the slope of the line). (T.G.)

Deterministic calculation of the effective delayed neutron fraction without using the adjoint neutron flux - 299

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.

2010-01-01

In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)

Measurement of the neutron capture cross-section of 232Th using the neutron activation technique

International Nuclear Information System (INIS)

Naik, H.; Singh, Sarbjit; Goswami, A.; Manchanda, V.K.; Prajapati, P.M.; Surayanarayana, S.V.; Nayak, B.K.; Sharma, S.C.; Jagadeesan, K.C.; Thakare, S.V.; Raj, D.; Ganesan, S.; Mulik, V.K.; Sivashankar, B.S.; Mukherjee, S.

2011-01-01

The 232 Th(n, γ) reaction cross-section at average neutron energies of 3.7±0.3 MeV and 9.85±0.38 MeV from the 7 Li(p, n) reaction has been determined for the first time using activation and off-line γ -ray spectrometric technique. The 232 Th(n, 2n) reaction cross-section at the average neutron energy of 9.85±0.38 MeV has been also determined using the same technique. The experimentally determined 232 Th(n, γ) and 232 Th(n, 2n) reaction cross-sections were compared with the evaluated data of ENDF/B-VII, JENDL-4.0 and JEFF-3.1 and were found to be in good agreement. The present data along with literature data in a wide range of neutron energies were interpreted in terms of competition between different reaction channels including fission. The 232 Th(n, γ) and 232 Th(n, 2n) reaction cross-sections were also calculated theoretically using the TALYS 1.2 computer code and were found to be slightly higher than the experimental data. (orig.)

Studies with radioactive beams - properties of neutron halo

International Nuclear Information System (INIS)

Tanihata, I.

1992-01-01

Interaction cross sections σ I and 9 Li transverse momentum distributions of 11 Li reactions were measured using p, d, Be and C targets at 800 A and 400 A MeV. The density distribution of 11 Li nucleus has been determined, for the first time, combining the interaction cross sections with various targets and energies. It was confirmed that only the distribution with long tail describe the observed data. The momentum correlation of two neutrons in the neutron halo is extracted from the P T distribution of 9 Li and that of neutron. It is found that the two neutrons are moving in the same direction in average and thus strongly suggests a formation of di-neutron in 11 Li

Measurement of the scattering cross section of slow neutrons on liquid parahydrogen from neutron transmission

Science.gov (United States)

Grammer, K. B.; Alarcon, R.; Barrón-Palos, L.; Blyth, D.; Bowman, J. D.; Calarco, J.; Crawford, C.; Craycraft, K.; Evans, D.; Fomin, N.; Fry, J.; Gericke, M.; Gillis, R. C.; Greene, G. L.; Hamblen, J.; Hayes, C.; Kucuker, S.; Mahurin, R.; Maldonado-Velázquez, M.; Martin, E.; McCrea, M.; Mueller, P. E.; Musgrave, M.; Nann, H.; Penttilä, S. I.; Snow, W. M.; Tang, Z.; Wilburn, W. S.

2015-05-01

Liquid hydrogen is a dense Bose fluid whose equilibrium properties are both calculable from first principles using various theoretical approaches and of interest for the understanding of a wide range of questions in many-body physics. Unfortunately, the pair correlation function g (r ) inferred from neutron scattering measurements of the differential cross section d/σ d Ω from different measurements reported in the literature are inconsistent. We have measured the energy dependence of the total cross section and the scattering cross section for slow neutrons with energies between 0.43 and 16.1 meV on liquid hydrogen at 15.6 K (which is dominated by the parahydrogen component) using neutron transmission measurements on the hydrogen target of the NPDGamma collaboration at the Spallation Neutron Source at Oak Ridge National Laboratory. The relationship between the neutron transmission measurement we perform and the total cross section is unambiguous, and the energy range accesses length scales where the pair correlation function is rapidly varying. At 1 meV our measurement is a factor of 3 below the data from previous work. We present evidence that these previous measurements of the hydrogen cross section, which assumed that the equilibrium value for the ratio of orthohydrogen and parahydrogen has been reached in the target liquid, were in fact contaminated with an extra nonequilibrium component of orthohydrogen. Liquid parahydrogen is also a widely used neutron moderator medium, and an accurate knowledge of its slow neutron cross section is essential for the design and optimization of intense slow neutron sources. We describe our measurements and compare them with previous work.

Cobalt, fast neutrons and physical models: Nuclear data and measurements series

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.; Lawson, R.D.

1987-07-01

Energy-averaged neutron total cross sections of cobalt were measured from ≅0.5 to 12.0 MeV. Differential elastic- and inelastic-scattering cross sections were measured from ≅1.5 to 10.0 MeV over the scattering-angle range ≅18 0 to 160 0 , with sufficient detail to define the energy-averaged behavior. Inelastic neutron groups were observed corresponding to ''levels'' at: 1115 +- 29, 1212 +- 24, 1307 +- 24, 1503 +- 33, 1778 +- 40, 2112 +- 40, 2224 +- 35, 2423 +- 39, 2593 +- 41 and 2810 keV. The experimental results were interpreted in terms of the spherical optical-statistical and coupled-channels models. An unusually successful description of observables was achieved over a wide energy range ( 20.0 MeV) with a spherical model having energy-dependent strengths and geometries. The energy dependencies are large below ≅7.0 MeV (i.e., ≅19.0 MeV above the Fermi energy), but become smaller and similar to those reported for ''global'' potentials at higher energies. The imaginary strength is large and decreases with energy. These imaginary-potential characteristics are attributed to neutron shell closure and collective-vibrational processes. The weak-coupling model also offers an explanation of the unusual negative energy slope and relatively small radius of the imaginary potential. The spherical optical model derived from the neutron-scattering results was extrapolated to bound energies using the dispersion relationship and the method of moments. The resulting real-potential strength and radius peak at ≅-10.0 MeV, while concurrently the real diffuseness is at a minimum. The extrapolated potential is ≅8% larger than that implied by reported particle-state energies, and ≅13% smaller than indicated by hole-state energies. 68 refs., 15 figs., 1 tab

Hematologic status of mice submitted to sublethal total body irradiation with mixed neutron-gamma radiation

International Nuclear Information System (INIS)

Herodin, F.; Court, L.

1989-01-01

The hematologic status of mice exposed to sublethal whole body irradiation with mixed neutron-gamma radiation (mainly neutrons) is studied. A slight decrease of the blood cell count is still observed below 1 Gy. The recovery of bone marrow granulocyte-macrophage progenitors seems to require more time than after pure gamma irradiation [fr

Total Ambient Dose Equivalent Buildup Factor Determination for Nbs04 Concrete.

Science.gov (United States)

Duckic, Paulina; Hayes, Robert B

2018-06-01

Buildup factors are dimensionless multiplicative factors required by the point kernel method to account for scattered radiation through a shielding material. The accuracy of the point kernel method is strongly affected by the correspondence of analyzed parameters to experimental configurations, which is attempted to be simplified here. The point kernel method has not been found to have widespread practical use for neutron shielding calculations due to the complex neutron transport behavior through shielding materials (i.e. the variety of interaction mechanisms that neutrons may undergo while traversing the shield) as well as non-linear neutron total cross section energy dependence. In this work, total ambient dose buildup factors for NBS04 concrete are calculated in terms of neutron and secondary gamma ray transmission factors. The neutron and secondary gamma ray transmission factors are calculated using MCNP6™ code with updated cross sections. Both transmission factors and buildup factors are given in a tabulated form. Practical use of neutron transmission and buildup factors warrants rigorously calculated results with all associated uncertainties. In this work, sensitivity analysis of neutron transmission factors and total buildup factors with varying water content has been conducted. The analysis showed significant impact of varying water content in concrete on both neutron transmission factors and total buildup factors. Finally, support vector regression, a machine learning technique, has been engaged to make a model based on the calculated data for calculation of the buildup factors. The developed model can predict most of the data with 20% relative error.

A set-up for measuring neutron cross sections and radiation multiplicity from neutron-nucleus interaction

International Nuclear Information System (INIS)

Georgiev, G.P.; Ermakov, V.A.; Grigor'ev, Yu.V.

1988-01-01

A multiplicity detector of the ''Romashka'' type has been used on the 500 m flight part of the IBR-30 pulsed reactor. The detector consists of 16 independent sections with NaJ(Tl) crystals with a total volume of 36 liters. The geometric efficiency of single-ray detection is ∼ 80%. The gamma-ray to neutron detection efficiency ratio is ≥600 for neutrons with energies below 200 keV. This detector allows one to perform neutron capture and fission cross section measurements and to study gamma-ray multiplicity and resonance selfabsorption effects in the 20 eV-200keV neutron energy range

Proposal of a wide-band mirror polarizer of slow neutrons at a pulsed neutron source

International Nuclear Information System (INIS)

Nikitenko, Yu.V.; Ostanevich, Yu.M.

1992-01-01

The new type wide-band mirror-based neutron polarizer to be operated at a pulsed neutron source is suggested. The idea is to use a movable polarizing mirror system, which, be the incoming beam monochromatized by the time-of-flight, would allow one to tune glancing angles in time so, that the total reflection condition is always fulfilled only for one of the two neutron spin eigenstates. Estimates show, that with the pulsed reactor IBR-2 such polarizer allows one to build a small-angle neutron scattering instrument capable to effectively use the wave-length band from 2 to 15 A. 9 refs.; 1 fig

Low-dose neutron dose response of zebrafish embryos obtained from the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility

International Nuclear Information System (INIS)

Ng, C.Y.P.; Kong, E.Y.; Konishi, T.; Kobayashi, A.; Suya, N.; Cheng, S.H.; Yu, K.N.

2015-01-01

The dose response of embryos of the zebrafish, Danio rerio, irradiated at 5 h post fertilization (hpf) by 2-MeV neutrons with ≤100 mGy was determined. The neutron irradiations were made at the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility in the National Institute of Radiological Sciences (NIRS), Chiba, Japan. A total of 10 neutron doses ranging from 0.6 to 100 mGy were employed (with a gamma-ray contribution of 14% to the total dose), and the biological effects were studied through quantification of apoptosis at 25 hpf. The responses for neutron doses of 10, 20, 25, and 50 mGy approximately fitted on a straight line, while those for neutron doses of 0.6, 1 and 2.5 mGy exhibited neutron hormetic effects. As such, hormetic responses were generically developed by different kinds of ionizing radiations with different linear energy transfer (LET) values. The responses for neutron doses of 70 and 100 mGy were significantly below the lower 95% confidence band of the best-fit line, which strongly suggested the presence of gamma-ray hormesis. - Highlights: • Neutron dose response was determined for embryos of the zebrafish, Danio rerio. • Neutron doses of 0.6, 1 and 2.5 mGy led to neutron hormetic effects. • Neutron doses of 70 and 100 mGy accompanied by gamma rays led to gamma-ray hormesis

Electron-capture delayed fission properties of neutron-deficient einsteinium nuclei

International Nuclear Information System (INIS)

Shaughnessy, Dawn A.

2000-01-01

Electron-capture delayed fission (ECDF) properties of neutron-deficient einsteinium isotopes were investigated using a combination of chemical separations and on-line radiation detection methods. 242 Es was produced via the 233 U( 14 N,5n) 242 Es reaction at a beam energy of 87 MeV (on target) in the lab system, and was found to decay with a half-life of 11 ± 3 seconds. The ECDF of 242 Es showed a highly asymmetric mass distribution with an average pre-neutron emission total kinetic energy (TKE) of 183 ± 18 MeV. The probability of delayed fission (P DF ) was measured to be 0.006 ± 0.002. In conjunction with this experiment, the excitation functions of the 233 U( 14 N,xn) 247-x Es and 233 U( 15 N,xn) 248-x Es reactions were measured for 243 Es, 244 Es and 245 Es at projectile energies between 80 MeV and 100 MeV

Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

Science.gov (United States)

Kim, Myong-Seop; Park, Sang-Jun

2009-08-01

Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

Neutron intensity of fast reactor spent fuel

Energy Technology Data Exchange (ETDEWEB)

Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1998-03-01

Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

Characterization of a high repetition-rate laser-driven short-pulsed neutron source

Science.gov (United States)

Hah, J.; Nees, J. A.; Hammig, M. D.; Krushelnick, K.; Thomas, A. G. R.

2018-05-01

We demonstrate a repetitive, high flux, short-pulsed laser-driven neutron source using a heavy-water jet target. We measure neutron generation at 1/2 kHz repetition rate using several-mJ pulse energies, yielding a time-averaged neutron flux of 2 × 105 neutrons s‑1 (into 4π steradians). Deuteron spectra are also measured in order to understand source characteristics. Analyses of time-of-flight neutron spectra indicate that two separate populations of neutrons, ‘prompt’ and ‘delayed’, are generated at different locations. Gamma-ray emission from neutron capture 1H(n,γ) is also measured to confirm the neutron flux.

Determination of the in-core power and the average core temperature of low power research reactors using gamma dose rate measurements

International Nuclear Information System (INIS)

Osei Poku, L.

2012-01-01

Most reactors incorporate out-of-core neutron detectors to monitor the reactor power. An accurate relationship between the powers indicated by these detectors and actual core thermal power is required. This relationship is established by calibrating the thermal power. The most common method used in calibrating the thermal power of low power reactors is neutron activation technique. To enhance the principle of multiplicity and diversity of measuring the thermal neutron flux and/or power and temperature difference and/or average core temperature of low power research reactors, an alternative and complimentary method has been developed, in addition to the current method. Thermal neutron flux/Power and temperature difference/average core temperature were correlated with measured gamma dose rate. The thermal neutron flux and power predicted using gamma dose rate measurement were in good agreement with the calibrated/indicated thermal neutron fluxes and powers. The predicted data was also good agreement with thermal neutron fluxes and powers obtained using the activation technique. At an indicated power of 30 kW, the gamma dose rate measured predicted thermal neutron flux of (1* 10 12 ± 0.00255 * 10 12 ) n/cm 2 s and (0.987* 10 12 ± 0.00243 * 10 12 ) which corresponded to powers of (30.06 ± 0.075) kW and (29.6 ± 0.073) for both normal level of the pool water and 40 cm below normal levels respectively. At an indicated power of 15 kW, the gamma dose rate measured predicted thermal neutron flux of (5.07* 10 11 ± 0.025* 10 11 ) n/cm 2 s and (5.12 * 10 11 ±0.024* 10 11 ) n/cm 2 s which corresponded to power of (15.21 ± 0.075) kW and (15.36 ± 0.073) kW for both normal levels of the pool water and 40 cm below normal levels respectively. The power predicted by this work also compared well with power obtained from a three-dimensional neutronic analysis for GHARR-1 core. The predicted power also compares well with calculated power using a correlation equation obtained from

Determination of europium content in Li_2SiO_3(Eu) by neutron activation analysis using Am-Be neutron source

International Nuclear Information System (INIS)

Naik, Yeshwant; Tapase, Anant Shamrao; Mhatre, Amol; Datrik, Chandrashekhar; Tawade, Nilesh; Kumar, Umesh; Naik, Haladhara

2016-01-01

Circulardiscs of Li_2SiO_3 doped with europium were prepared and a new activation procedure for the neutron dose estimation in a breeder blanket of fusion reactor is described. The amount of europium in the disc was determined by neutron activation analysis (NAA) using an isotopic neutron source. The average neutron absorption cross section for the reaction was calculated using neutron distribution of the Am-Be source and available neutron absorption cross section data for the "1"5"1Eu(n,γ)"1"5"2"mEu reaction, which was used for estimation of europium in the pallet. The cross section of the elements varies with neutron energy, and the flux of the neutrons in each energy range seen by the nuclei under investigation also varies. Neutron distribution spectrum of the Am-Be source was worked out prior to NAA and the effective fractional flux for the nuclear reaction considered for the flux estimation was also determined. - Highlights: • Lithium meta-silicate is breeder materials for a fusion reactor. • Europium is used for neutron dose estimation in a breeder blanket. • It is important to determine amount of europium in lithium meta-silicate. • Amount of europium in lithium meta-silicate was determined by neutron activation and off-line gamma spectrometry.

Design, Construction, and Modeling of a 252Cf Neutron Irradiator

Directory of Open Access Journals (Sweden)

Blake C. Anderson

2016-01-01

Full Text Available Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

Maxwellian-averaged cross sections calculated from JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Tsuneo; Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ohsaka, Toshiro; Igashira, Masayuki [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

2000-02-01

Maxwellian-averaged cross sections of neutron capture, fission, (n,p) and (n,{alpha}) reactions are calculated from the Japanese Evaluated Nuclear Data Library, JENDL-3.2, for applications in the astrophysics. The calculation was made in the temperature (kT) range from 1 keV to 1 MeV. Results are listed in tables. The Maxwellian-averaged capture cross sections were compared with recommendations of other authors and recent experimental data. Large discrepancies were found among them especially in the light mass nuclides. Since JENDL-3.2 reproduces relatively well the recent experimental data, we conclude that JENDL-3.2 is superior to the others in such a mass region. (author)

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

Neutron scattering lengths of 3He

International Nuclear Information System (INIS)

Alfimenkov, V.P.; Akopian, G.G.; Wierzbicki, J.; Govorov, A.M.; Pikelner, L.B.; Sharapov, E.I.

1976-01-01

The total neutron scattering cross-section of 3 He has been measured in the neutron energy range from 20 meV to 2 eV. Together with the known value of coherent scattering amplitude it leads to the two sts of n 3 He scattering lengths

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

CERN Document Server

Vega, H R

2002-01-01

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

Determination of the average number of neutrons per fission event for californium-252

International Nuclear Information System (INIS)

Aleksandrov, B.M.; Belov, L.M.; Drapchinskij, L.V.

1982-01-01

By means of a separate determination of neutron yields and fission event rates, the value of #betta#-bar( 252 Cf) has been measured for a series of new high-purity sources. The improved quality of the source active layers has reduced the error in determining the fission rate to 0.35%. The value obtained for #betta#-bar( 252 Cf) is 3.747+-0.036. A description is given of the design and the parameters of a spherical manganese bath in which the work on refining the value of #betta#-bar( 252 Cf) will be continued. (author)

Average and local structure of α-CuI by configurational averaging

International Nuclear Information System (INIS)

Mohn, Chris E; Stoelen, Svein

2007-01-01

Configurational Boltzmann averaging together with density functional theory are used to study in detail the average and local structure of the superionic α-CuI. We find that the coppers are spread out with peaks in the atom-density at the tetrahedral sites of the fcc sublattice of iodines. We calculate Cu-Cu, Cu-I and I-I pair radial distribution functions, the distribution of coordination numbers and the distribution of Cu-I-Cu, I-Cu-I and Cu-Cu-Cu bond-angles. The partial pair distribution functions are in good agreement with experimental neutron diffraction-reverse Monte Carlo, extended x-ray absorption fine structure and ab initio molecular dynamics results. In particular, our results confirm the presence of a prominent peak at around 2.7 A in the Cu-Cu pair distribution function as well as a broader, less intense peak at roughly 4.3 A. We find highly flexible bonds and a range of coordination numbers for both iodines and coppers. This structural flexibility is of key importance in order to understand the exceptional conductivity of coppers in α-CuI; the iodines can easily respond to changes in the local environment as the coppers diffuse, and a myriad of different diffusion-pathways is expected due to the large variation in the local motifs

Dispersion of the Neutron Emission in U{sup 235} Fission

Science.gov (United States)

Feynman, R. P.; de Hoffmann, F.; Serber, R.

1955-01-01

Equations are developed which allow the calculation of the average number of neutrons per U{sup235} fission from experimental measurements. Experimental methods are described, the results of which give a value of (7.8 + 0.6){sup ½} neutrons per U{sup 235} thermal fission.

Calculation of displacement, gas, and transmutation production in stainless steel irradiated with spallation neutrons

International Nuclear Information System (INIS)

Wechsler, M.A.; Ramavarapu, R.; Daugherty, E.L.; Palmer, R.C.; Bullen, D.B.; Sommer, W.F.

1993-01-01

Calculations using the high-energy transport code LAHET have been made for the production of displacements, helium gas, and transmuted atoms for stainless steel (Fe-18 wt % Cr-10 wt % Ni) irradiated with spallation neutrons at energies of 100 to 1600 MeV. The damage energy cross section increased from about 250 to 350 b keV for increasing neutron energies from 100 to 1600 MeV with a spallation spectrum average of 281 barns-keV. For a displacement threshold energy of 33 eV, the corresponding spectrum-average displacement cross section is 3400 barns. The PKA spectrum was found to be fairly independent of the incident neutron energy, with an average damage energy of 0.25--0.30 MeV. The helium production cross section increased monotonically with increasing neutron energy, with a spectrum average of 0.32 barns. The maximum transmutation yield was observed near manganese (Z = 25), corresponding to a production cross section of about 0.2 barns. Relevance to fusion materials is discussed

Estimation of 242Cm neutron cross sections for total file creation

International Nuclear Information System (INIS)

Bakhanovich, L.A.; Klepatskij, A.B.; Maslov, V.M.; Porodzinskij, Yu.V.; Sukhovitskij, E.Sh.

1989-01-01

Description of evaluation of 242 Cm neutron cross sections in 10 -5 eV-20 MeV energy range is given. Due to the lack of experimental data the evaluation is largely based on the application of theoretical models and systematics. The data obtained are compared to evaluations by other authors. 10 refs.; 2 figs.; 2 tabs

ATLAS MDT neutron sensitivity measurement and modeling

International Nuclear Information System (INIS)

Ahlen, S.; Hu, G.; Osborne, D.; Schulz, A.; Shank, J.; Xu, Q.; Zhou, B.

2003-01-01

The sensitivity of the ATLAS precision muon detector element, the Monitored Drift Tube (MDT), to fast neutrons has been measured using a 5.5 MeV Van de Graaff accelerator. The major mechanism of neutron-induced signals in the drift tubes is the elastic collisions between the neutrons and the gas nuclei. The recoil nuclei lose kinetic energy in the gas and produce the signals. By measuring the ATLAS drift tube neutron-induced signal rate and the total neutron flux, the MDT neutron signal sensitivities were determined for different drift gas mixtures and for different neutron beam energies. We also developed a sophisticated simulation model to calculate the neutron-induced signal rate and signal spectrum for ATLAS MDT operation configurations. The calculations agree with the measurements very well. This model can be used to calculate the neutron sensitivities for different gaseous detectors and for neutron energies above those available to this experiment

Local magnetic structure determination using polarized neutron holography

International Nuclear Information System (INIS)

Szakál, Alex; Markó, Márton; Cser, László

2015-01-01

A unique and important property of the neutron is that it possesses magnetic moment. This property is widely used for determination of magnetic structure of crystalline samples observing the magnetic components of the diffraction peaks. Investigations of diffraction patterns give information only about the averaged structure of a crystal but for discovering of local spin arrangement around a specific (e.g., impurity) nucleus remains still a challenging problem. Neutron holography is a useful tool to investigate the local structure around a specific nucleus embedded in a crystal lattice. The method has been successfully applied experimentally in several cases using non-magnetic short range interaction of the neutron and the nucleus. A mathematical model of the hologram using interaction between magnetic moment of the atom and the neutron spin for polarized neutron holography is provided. Validity of a polarized neutron holographic experiment is demonstrated by applying the proposed method on model systems

Future prospects of imaging at spallation neutron sources

International Nuclear Information System (INIS)

Strobl, M.

2009-01-01

The advent of state-of-the-art spallation neutron sources is a major step forward in efficient neutron production for most neutron scattering techniques. Although they provide lower time-averaged neutron flux than high flux reactor sources, advantage for different instrumental techniques can be derived from the pulsed time structure of the available flux, which can be translated into energy, respectively, wavelength resolution. Conventional neutron imaging on the other hand relies on an intense continuous beam flux and hence falls short in profiting from the new development. Nevertheless, some recently developed novel imaging techniques require and some can benefit from energy resolution. The impact of the emerging spallation sources on different imaging techniques has been investigated, ways to benefit will be identified (where possible) and prospects of future imaging instruments and possible options and layouts at a spallation neutron source will be discussed and outlined.

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

International Nuclear Information System (INIS)

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

1997-03-01

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

The recovery of bone marrow derived GM-CFU in baboons unilaterally exposed to a total body LD50/30d mixed neutron-gamma irradiation

International Nuclear Information System (INIS)

Herodin, F.; Orfeuvre, H.; Janodet, D.; Mestries, J.C.; Fatome, M.

1990-01-01

The unilateral exposure of baboons to a total body LD 50/30d mixed neutron/gamma irradiation was characterized to be non uniform in dose distribution. The pattern of recovery of granulocyte-macrophage progenitors in bone marrow samples collected from entrance and exit sides respectively is consistent with this observed heterogeneity [fr

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

Science.gov (United States)

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

2018-01-01

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

Dosimetric comparative analysis between 10 MV Megavoltage unidirectional beam and boron neutron capture therapy for brain tumors treatment

International Nuclear Information System (INIS)

Brandao, Samia F.; Campos, Tarcisio P.R.

2011-01-01

This paper present a comparative dosimetric analysis between boron neutron capture therapy and 10 MV megavoltage employed in brain tumor treatments, limited to a unidirectional beam. A computational phantom of a human head was developed to be used in computational simulations of the two protocols, conducted in MCNP5 code. This phantom represents several head's structures, mainly, the central nervous system and a tumor that represents a Glioblastoma Multiform - one of the most malignant and aggressive brain tumors. Absorbed and biological weighted dose rates and neutron fluency in the computational phantom were evaluated from the MCNP5 code. The biologically weighted dose rate to 10 MV megavoltage beam presented no specificity in deposited dose in tumor. The average total biologically weighted dose rate in tumor was 9.93E-04 RBE.Gy.h"-"1/Mp.s"-"1 while in healthy tissue it was 8.67E-04 RBE.Gy.h"-"1/Mp.s"-1. On the BNCT simulations the boron concentration was particularly relevant since the largest dose deposition happened in borate tissues. The average total biologically weighted dose rate in tumor was 3.66E-02 RBE.Gy.h"-"1/Mp.s"-"1 while in healthy tissue it was 1.39E-03 RBE.Gy.h"-"1/Mp.s"-"1. In comparison to the 10 MV megavoltage beam, BNCT showed clearly a largest dose deposition in the tumor, on average, 37 times larger than in the megavoltage beam, while in healthy tissue that average was only 1,6 time larger in BNCT. (author)

Fullerene films and fullerene-dodecylamine adduct monolayers at air-water interfaces studied by neutron and x-ray reflection

DEFF Research Database (Denmark)

Wang, J.Y.; Vaknin, D.; Uphaus, R.A.

1994-01-01

Neutron and X-ray reflection measurements and surface pressure isotherms of spread films of the fullerene-dodecylamine adduct C60-[NH2(CH2)11CH3]x all indicate that this material may form monomolecular layers on water surfaces. The reflection data sets (neutron on both H2O and D2O) can be accounted...... for by a single model structure defined in terms of the dimensions of an average cell and its chemical composition. This model ascribes a total thickness of about 29 angstrom to the molecular interface layer with the following internal structure. The fullerenes (with several alkyl chains attached) form a central...... stratum and the remainder alkyl tails are located close to both the air and the water interfaces. The alkyl moieties close to the aqueous substrate are hydrated. The reflection experiments and the isotherms suggest that on average 8 +/- 3 dodecylamine molecules are present per fullerene, consistent within...

In-core neutron flux measurements at PARR using self powered neutron detector

International Nuclear Information System (INIS)

Hussain, A.; Ansari, S.A.

1989-10-01

This report describes experimental reactor physics measure ments at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response time determination and neutron flux measurements. The detectors were calibrated at low power using gold foils and full power by the thermal channel. Based on this calibration it was observed that the detector response remains almost linear throughout the power range. The self powered detectors were used for on-line determination of absolute neutron flux in the core as well as the spatial distribution of neutron flux or reactor power. The experimental, axial and horizontal flux mapping results at certain locations in the core are presented. The total response time of rhodium detector was experimentally determined to be about 5 minutes, which agree well with the theoretical results. Because of longer response time of SPND of the detectors it is not possible to use them in the reactor protection system. (author). 10 figs

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.

Parameters measurement for the thermal neutron beam in the thermal column hole of Xi’an pulse reactor

Institute of Scientific and Technical Information of China (English)

无

2010-01-01

The distribution of the neutron spectra in the thermal column hole of Xi’an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.

Benchmarking a first-principles thermal neutron scattering law for water ice with a diffusion experiment

Directory of Open Access Journals (Sweden)

Holmes Jesse

2017-01-01

Full Text Available The neutron scattering properties of water ice are of interest to the nuclear criticality safety community for the transport and storage of nuclear materials in cold environments. The common hexagonal phase ice Ih has locally ordered, but globally disordered, H2O molecular orientations. A 96-molecule supercell is modeled using the VASP ab initio density functional theory code and PHONON lattice dynamics code to calculate the phonon vibrational spectra of H and O in ice Ih. These spectra are supplied to the LEAPR module of the NJOY2012 nuclear data processing code to generate thermal neutron scattering laws for H and O in ice Ih in the incoherent approximation. The predicted vibrational spectra are optimized to be representative of the globally averaged ice Ih structure by comparing theoretically calculated and experimentally measured total cross sections and inelastic neutron scattering spectra. The resulting scattering kernel is then supplied to the MC21 Monte Carlo transport code to calculate time eigenvalues for the fundamental mode decay in ice cylinders at various temperatures. Results are compared to experimental flux decay measurements for a pulsed-neutron die-away diffusion benchmark.

Evaluation and calculation of neutron transactinide cross-sections

International Nuclear Information System (INIS)

Konshin, V.A.

1980-01-01

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

The fluence research of filter material for fast neutron fluence measurement

International Nuclear Information System (INIS)

Tang Xiding

2010-01-01

When the fast neutron fluence is measured by radioactivation techniques in the nuclear reactor the fast neutron is also filtered a little by the thermal neutron filter material, and if the filter material thickness increase the filtered fast neutron increases therewith. For fast neutron fluenc measurement, there are only cadmium, boron and gadolinium three elements filtering fluence can be calculated ordinarily. In order to calculate the filtered fast neutron fluence of the all elements in the filter material, the many total cross sections of nuclides had checked out from nuclear cross section data library, converted them into the same energy group structure, then element's total cross section, compound's total cross section and multilayer filters' total cross section had calculated from these total cross sections with same energy group structure, a new cross section data library can be obtained lastly through merging these cross sections into the old cross section data library used for neutron fluence measurement. The calculation analysis indicates that the results of the unit 2 surveillance capsule U of DAYA Bay NPP and the unit 1 surveillance capsule A of the Second Nuclear Power Plant of Qinshan by considering the all elements subtracting iron are smaller about 1.5% and 2.6% respectively than the ones only to consider cadmium, boron. The old measured results accord with the new values under the measurement uncertainty, are reliable. The new results are more accuracy. (authors)

Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

International Nuclear Information System (INIS)

Rossa, Riccardo; Borella, Alessandro; Labeau, Pierre-Etienne; Pauly, Nicolas; Meer, Klaas van der

2015-01-01

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of 239 Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of 239 Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a 239 Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to 239 Pu, in comparison with a 235 U fission chamber, with a 3 He proportional counter, and with a 10 B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the 239 Pu and 235 U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the 3 He and 10 B proportional counters to increase the sensitivity to 239 Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies

Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

Energy Technology Data Exchange (ETDEWEB)

Rossa, Riccardo, E-mail: rrossa@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Borella, Alessandro, E-mail: aborella@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Labeau, Pierre-Etienne, E-mail: pelabeau@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Pauly, Nicolas, E-mail: nipauly@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Meer, Klaas van der, E-mail: kvdmeer@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium)

2015-08-11