The TENDL neutron data library and the TEND1038 38-group neutron constant system

International Nuclear Information System (INIS)

Abramovich, S.N.; Gorelov, V.P.; Gorshikhin, A.A.; Grebennikov, A.N.; Il'in, V.N.; Krut'ko, N.A.; Farafontov, G.G.

2002-01-01

The library contains neutron data for 103 nuclei - i.e. for 38 actinide nuclei (from 232 Th to 249 Cm), 26 fission fragment nuclei and 39 nuclei in structural and technological materials. The 38-group constants were obtained from TENDL. The high-energy group boundary is 20 MeV. The energy range below 1.2 eV contains 11 groups. Temperature and resonance effects were taken into account. The delayed neutron parameters for 6 groups and the yields of 40 fission fragments were obtained (light and heavy, stable and non-stable). The fast neutron features of spherical critical assemblies were calculated using constants from TEND1038. (author)

Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

Energy Technology Data Exchange (ETDEWEB)

Hosseini, Seyed Abolfazl, E-mail: sahosseini@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Tehran 8639-11365 (Iran, Islamic Republic of); Afrakoti, Iman Esmaili Paeen [Faculty of Engineering & Technology, University of Mazandaran, Pasdaran Street, P.O. Box: 416, Babolsar 47415 (Iran, Islamic Republic of)

2017-04-11

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The {sup 241}Am-{sup 9}Be and {sup 252}Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions. - Highlights: • The neutron pulse height distribution was simulated using MCNPX-ESUT. • The energy spectrum of the neutron source was unfolded using GMDH. • The energy spectrum of the neutron source was

Neutron excess generation by fusion neutron source for self-consistency of nuclear energy system

International Nuclear Information System (INIS)

Saito, Masaki; Artisyuk, V.; Chmelev, A.

1999-01-01

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities. (author)

Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

Multi-group neutron transport theory

International Nuclear Information System (INIS)

Zelazny, R.; Kuszell, A.

1962-01-01

Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr

8-group relative delayed neutron yields for monoenergetic neutron induced fission of 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

The energy dependence of the relative yield of delayed neutrons in an 8-group model representation was obtained for monoenergetic neutron induced fission of 239 Pu. A comparison of this data with the available experimental data by other authors was made in terms of the mean half-life of the delayed neutron precursors. (author)

Three-group albedo method applied to the diffusion phenomenon with up-scattering of neutrons

International Nuclear Information System (INIS)

Terra, Andre M. Barge Pontes Torres; Silva, Jorge A. Valle da; Cabral, Ronaldo G.

2007-01-01

The main objective of this research is to develop a three-group neutron Albedo algorithm considering the up-scattering of neutrons in order to analyse the diffusion phenomenon in nonmultiplying media. The neutron Albedo method is an analytical method that does not try to solve describing explicit equations for the neutron fluxes. Thus the neutron Albedo methodology is very different from the conventional methodology, as the neutron diffusion theory model. Graphite is analyzed as a model case. One major application is in the determination of the nonleakage probabilities with more understandable results in physical terms than conventional radiation transport method calculations. (author)

Effects of core models and neutron energy group structures on xenon oscillation in large graphite-moderated reactors

International Nuclear Information System (INIS)

Yamasita, Kiyonobu; Harada, Hiroo; Murata, Isao; Shindo, Ryuichi; Tsuruoka, Takuya.

1993-01-01

Xenon oscillations of large graphite-moderated reactors have been analyzed by a multi-group diffusion code with two- and three-dimensional core models to study the effects of the geometric core models and the neutron energy group structures on the evaluation of the Xe oscillation behavior. The study clarified the following. It is important for accurate Xe oscillation simulations to use the neutron energy group structure that describes well the large change in the absorption cross section of Xe in the thermal energy range of 0.1∼0.65 eV, because the energy structure in this energy range has significant influences on the amplitude and the period of oscillations in power distributions. Two-dimensional R-Z models can be used instead of three-dimensional R-θ-Z models for evaluation of the threshold power of Xe oscillation, but two-dimensional R-θ models cannot be used for evaluation of the threshold power. Although the threshold power evaluated with the R-θ-Z models coincides with that of the R-Z models, it does not coincide with that of the R-θ models. (author)

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

International Nuclear Information System (INIS)

Pazirandeh, A.; Shekarian, E.

2006-01-01

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

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

International Nuclear Information System (INIS)

Zalan, T.A.

1988-01-01

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

High energy neutron radiography

International Nuclear Information System (INIS)

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

1996-01-01

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

Prototype Neutron Energy Spectrometer

International Nuclear Information System (INIS)

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

2010-01-01

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

Prototype Neutron Energy Spectrometer

Energy Technology Data Exchange (ETDEWEB)

Stephen Mitchell, Sanjoy Mukhopadhyay, Richard Maurer, Ronald Wolff

2010-06-16

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

Neutron matter, symmetry energy and neutron stars

Energy Technology Data Exchange (ETDEWEB)

Stefano, Gandolfi [Los Alamos National Laboratory (LANL); Steiner, Andrew W [ORNL

2016-01-01

Recent progress in quantum Monte Carlo with modern nucleon-nucleon interactions have enabled the successful description of properties of light nuclei and neutron-rich matter. Of particular interest is the nuclear symmetry energy, the energy cost of creating an isospin asymmetry, and its connection to the structure of neutron stars. Combining these advances with recent observations of neutron star masses and radii gives insight into the equation of state of neutron-rich matter near and above the saturation density. In particular, neutron star radius measurements constrain the derivative of the symmetry energy.

Energy group structure determination using particle swarm optimization

International Nuclear Information System (INIS)

Yi, Ce; Sjoden, Glenn

2013-01-01

Highlights: ► Particle swarm optimization is applied to determine broad group structure. ► A graph representation of the broad group structure problem is introduced. ► The approach is tested on a fuel-pin model. - Abstract: Multi-group theory is widely applied for the energy domain discretization when solving the Linear Boltzmann Equation. To reduce the computational cost, fine group cross libraries are often down-sampled into broad group cross section libraries. Cross section data collapsing generally involves two steps: Firstly, the broad group structure has to be determined; secondly, a weighting scheme is used to evaluate the broad cross section library based on the fine group cross section data and the broad group structure. A common scheme is to average the fine group cross section weighted by the fine group flux. Cross section collapsing techniques have been intensively researched. However, most studies use a pre-determined group structure, open based on experience, to divide the neutron energy spectrum into thermal, epi-thermal, fast, etc. energy range. In this paper, a swarm intelligence algorithm, particle swarm optimization (PSO), is applied to optimize the broad group structure. A graph representation of the broad group structure determination problem is introduced. And the swarm intelligence algorithm is used to solve the graph model. The effectiveness of the approach is demonstrated using a fuel-pin model

The applied research program of the High Flux Neutron Generator at the National Nuclear Center, Havana

International Nuclear Information System (INIS)

Perez, G.; Martin, G.; Ceballos, C.; Padron, I.; Shtejer, K.; Perez, N.; Guibert, R.; Ledo, L.M.; Cruz Inclan, Carlos

2001-01-01

The Havana High Flux Neutron Generator facility is an intense neutron source based on a 20 mA duoplasmatron ion source and a 250 kV high voltage power supply. It has been installed in the Neutron Generator Laboratory at the Center of Applied Technologies and Nuclear Research in 1997. This paper deal outlined the future applied program to be carried out in this facility in the next years. The Applied Research Program consists on install two nuclear analytic techniques: the PELAN technique which uses the neutron generator in the pulse mode and the Low Energy PIXE technique which uses the same facility as a low energy proton accelerator for PIXE analysis. (author)

Energy dependence of fast neutron dosimetry using electrochemical etching

International Nuclear Information System (INIS)

Su, S.J.; Morgan, K.Z.

1978-01-01

Registration of fast-neutron induced recoil tracks by the electrochemical etching technique as applied to sensitive Lexan polycarbonate foils provides a simple and inexpensive means of fast neutron personnel dosimetry. The sensitivity (tracks/neutron) of recoil particle registration is given as a function of neutron energy. Neutrons of 7 Li (p,n) 7 Be, 3 T (d,n) 4 He and 9 B, respectively. Results are compared with other studies using other neutron sources and conventional etching method

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

International Nuclear Information System (INIS)

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

2014-01-01

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

An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

International Nuclear Information System (INIS)

Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

2013-01-01

Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

Research activities on dosimetry for high energy neutrons

Energy Technology Data Exchange (ETDEWEB)

Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

The external dosimetry research group of JAERI has been calculating dose conversion coefficients for high-energy radiations using particle transport simulation codes. The group has also been developing radiation dose measurement techniques for high-energy neutrons in collaboration with some university groups. (author)

Intercomparison of medium-energy neutron attenuation in iron and concrete

International Nuclear Information System (INIS)

Hirayama, H.

1999-01-01

Neutron attenuation of medium energy below 1 GeV has not been well understood until now. It is desired to obtain common agreements concerning the behaviours of neutrons inside various materials. This is necessary in order to agree on definitions of the attenuation length, which is very important for shielding calculations involving high energy accelerators. As one attempt, it was proposed by Japanese attendants of SATIF-2 to compare the attenuation of medium-energy neutrons inside iron and concrete shields between various computer codes and data, and was cited as a suitable action for SATIF. The first results from three groups were presented at SATIF-3. It has become clear that neutrons above 20 MeV are important for understanding the attenuation inside materials and that the geometry, planar or spherical, does not affect the results very much. Considering the CPU times required for Monte Carlo calculations and this result, revised problems to be calculated were prepared by the Japanese Working Group and sent to the participants of this action. The geometry is only plane, and calculations are required only for neutrons above 20 MeV. The secondary neutrons from high energy protons, which were calculated by H. Nakashima, are also included in the problem. The results from four groups were sent to the organizer at the end of August. This paper presents a comparison between groups concerning the attenuation length together with the neutron spectrum and the future themes which come from this intercomparison. (author)

High energy neutron dosimetry for the fusion program

International Nuclear Information System (INIS)

Barr, D.W.; Norris, A.E.

1977-01-01

Neutron dosimetry by the foil activation method offers a flexible technique for characterizing neutron spectra ranging from thermal energies to 30 MeV with the potential for extension to higher neutron energies as investigated by the Los Alamos Radiochemistry Group at the Los Alamos Meson Physics Facility and in the Apollo-Soyuz Test Project. The use of this method for the neutron flux description in thermal, resonance, and fission spectrum assemblies has been demonstrated. An extension of the method to environments involving thermonuclear processes was developed at Los Alamos in the early 1950's to characterize mixed fission-thermonuclear systems

Discrete energy formulation of neutron transport theory applied to solving the discrete ordinates equations

International Nuclear Information System (INIS)

Ching, J.; Oblow, E.M.; Goldstein, H.

1976-01-01

An algebraic equivalence between the point-energy and multigroup forms of the Boltzmann transport equation is demonstrated that allows the development of a discrete energy, discrete ordinates method for the solution of radiation transport problems. In the discrete energy method, the group averaging required in the cross-section processing for multigroup calculations is replaced by a faster numerical quadrature scheme capable of generating transfer cross sections describing all the physical processes of interest on a fine point-energy grid. Test calculations in which the discrete energy method is compared with the multigroup method show that, for the same energy grid, the discrete energy method is much faster, although somewhat less accurate, than the multigroup method. However, the accuracy of the discrete energy method increases rapidly as the spacing between energy grid points is decreased, approaching that of multigroup calculations. For problems requiring great detail in the energy spectrum, the discrete energy method is therefore expected to be far more economical than the multigroup technique for equivalent accuracy solutions. This advantage of the point method is demonstrated by application to the study of neutron transport in a thick iron slab

Euratom Neutron Radiography Working Group

DEFF Research Database (Denmark)

Domanus, Joseph Czeslaw

1986-01-01

reactor fuel as well as establish standards for radiographic image quality of neutron radiographs. The NRWG meets once a year in each of the neutron radiography centers to review the progress made and draw plans for the future. Besides, ad-hoc sub-groups or. different topics within the field of neutron......In 1979 a Neutron Radiography Working Group (NRWG) was constituted within Buratom with the participation of all centers within the European Community at which neutron facilities were available. The main purpose of NRWG was to standardize methods and procedures used in neutron radiography of nuclear...... radiography are constituted. This paper reviews the activities and achievements of the NRWG and its sub-groups....

Reactor-moderated intermediate-energy neutron beams for neutron-capture therapy

International Nuclear Information System (INIS)

Less, T.J.

1987-01-01

One approach to producing an intermediate energy beam is moderating fission neutrons escaping from a reactor core. The objective of this research is to evaluate materials that might produce an intermediate beam for NCT via moderation of fission neutrons. A second objective is to use the more promising moderator material in a preliminary design of an NCT facility at a research reactor. The evaluations showed that several materials or combinations of materials could produce a moderator source for an intermediate beam for NCT. The best neutron spectrum for use in NCT is produced by Al 2 O 3 , but mixtures of Al metal and D 2 O are also attractive. Using the best moderator materials, results were applied to the design of an NCT moderator at the Georgia Institute of Technology Research Reactor's bio-medical facility. The amount of photon shielding and thermal neutron absorber were optimized with respect to the desired photon dose rate and intermediate neutron flux at the patient position

Neutron Field Characterization of Irradiation Locations Applied to the Slovenian TRIGA Reactor

International Nuclear Information System (INIS)

Barbot, Loic; Domergue, Christophe; Breaud, Stephane; Destouches, Christophe; Villard, Jean-Francois; Snoj, Luka; Stancar, Ziga; Radulovic, Vladimir; Trkov, Andrej

2013-06-01

This work deals with several neutron flux measurement instruments and particle transport calculations combined in a method to assess the neutron field in experimental locations in nuclear reactor core or reflector. First test of this method in the TRIGA Mark II of Slovenia led to the assessment of three energy groups neutron fluxes in central irradiation locations within reactor core. (authors)

Determining space-energy distribution of thermal neutrons in heterogeneous cylindrically symmetric reactor cell, Master Thesis

International Nuclear Information System (INIS)

Matausek, M. V.

1966-06-01

A combination of multigroup method and P 3 approximation of spherical harmonics method was chosen for calculating space-energy distribution of thermal neutron flux in elementary reactor cell. Application of these methods reduced solution of complicated transport equation to the problem of solving an inhomogeneous system of six ordinary firs-order differential equations. A procedure is proposed which avoids numerical solution and enables analytical solution when applying certain approximations. Based on this approach, computer codes were written for ZUSE-Z-23 computer: SIGMA code for calculating group constants for a given material; MULTI code which uses results of SIGMA code as input and calculates spatial ana energy distribution of thermal neutron flux in a reactor cell. Calculations of thermal neutron spectra for a number of reactor cells were compared to results available from literature. Agreement was satisfactory in all the cases, which proved the correctness of the applied method. Some possibilities for improving the precision and acceleration of the calculation process were found during calculation. (author)

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.

Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

1993-02-01

A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

Measurement of high energy neutrons via Lu(n,xn) reactions

International Nuclear Information System (INIS)

Henry, E.A.; Becker, J.A.; Archer, D.E.; Younes, W.; Stoyer, M.A.; Slaughter, D.

1997-07-01

High energy neutrons can be assayed by the use of the nuclear diagnostic material lutetium. We are measuring the (n,xn) cross sections for natural lutetium in order to develop it as a detector material. We are applying lutetium to diagnose the high energy neutrons produced in test target/blanket systems appropriate for the Accelerator Production of Tritium Project. 3 refs., 5 figs., 1 tab

“Influence Method” applied to measure a moderated neutron flux

International Nuclear Information System (INIS)

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

2016-01-01

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

Direct capture of low-energy neutrons by {sup 16}O

Energy Technology Data Exchange (ETDEWEB)

Kitazawa, Hideo [Tokyo Inst. of Tech., Nagatsuta, Yokohama (Japan). Interdisciplinary Graduate School of Science; Igashira, Masayuki

1998-03-01

A dispersive optical potential for the interaction between low-energy neutrons and {sup 16}O-nuclei is derived from a dispersion relation based on the Feshbach generalized optical model. This potential is applied to direct-capture model calculations in explaining the observed off-resonance capture transitions to the ground (5/2{sup +}) and 871 keV(1/2{sup +}) levels in {sup 17}O at neutron energies of 20-70 keV. The model calculations take account of the spatial nonlocality of the neutron-nucleus interaction potential. (author)

Generating energy dependent neutron flux maps for effective ...

African Journals Online (AJOL)

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

Logic Estimation of the Optimum Source Neutron Energy for BNCT of Brain Tumors

International Nuclear Information System (INIS)

Dorrah, M.A.; Gaber, F.A.; Abd Elwahab, M.A.; Kotb, M.A.; Mohammed, M.M.

2012-01-01

BNCT is very complicated technique; primarily due to the complexity of element composition of the brain. Moreover; numerous components contributes to the over all radiation dose both to normal brain and to tumor. Simple algebraic summation cannot be applied to these dose components, since each component should at first be weighed by its relative biological effectiveness (RBE) value. Unfortunately, there is no worldwide agreement on these RBE values. For that reason, the parameters required for accurate planning of BNCT of brain tumors located at different depths in brain remained obscure. The most important of these parameters is; the source neutron energy. Thermal neutrons were formerly employed for BNCT, but they failed to prove therapeutic efficacy. Later on; epithermal neutrons were suggested proposing that they would be enough thermalized while transporting in the brain tissues. However; debate aroused regarding the source neutrons energy appropriate for treating brain tumors located at different depths in brain. Again, the insufficient knowledge regarding the RBE values of the different dose components was a major obstacle. A new concept was adopted for estimating the optimum source neutrons energy appropriate for different circumstances of BNCT. Four postulations on the optimum source neutrons energy were worked out, almost entirely independent of the RBE values of the different dose components. Four corresponding condition on the optimum source neutrons energy were deduced. An energy escalation study was carried out investigating 65 different source neutron energies, between 0.01 eV and 13.2 MeV. MCNP4B Monte C arlo neutron transport code was utilized to study the behavior of neutrons in the brain. The deduced four conditions were applied to the results of the 65 steps of the neutron energy escalation study. A source neutron energy range of few electron volts (eV) to about 30 keV was estimated to be the most appropriate for BNCT of brain tumors located at

Report from the neutron diffraction work group

International Nuclear Information System (INIS)

1978-08-01

This progress report of the neutron diffraction group at the Hahn Meitner Institute in Berlin comprises the following contributions: Three-dimensional critical properties of CsNiF 3 around the Neel point; Spin waves in CsNiF 3 with an applied magnetic field; Solitons in CsNiF 3 : Their experimental evidence and their thermodynamics; Neutron diffraction study of DAG at very low temperatures and in external magnetic field; Neutron diffraction investigation of tricritical behaviour in DyPO 4 ; Crystalline modifications and structural phase transitions of NaOH; Gitterdynamik von Cerhydrid; Investigation of the ferroelectric-ferroelastic phase transition in KH 2 PO 4 and RbH 2 PO 4 by means of γ-ray diffractometry; A γ-ray diffractometer for systematic measurements of absolute structure factors; Electron density in pyrite by combined γ-ray and neutron diffraction measurements: Thermal parameters from short wavelength neutron data; Accurate determination of temperature parameters from neutron diffraction data: Direct observation of the thermal diffuse scattering from silicon using perfect crystals; A Compton spectrometer for momentum density studies using 412 keV γ-radiation; Investigation of the electronic structure of Niobiumhydrides by means of gamma-ray Compton scattering; Interpretation of Compton profile data in position space; High resolution neutron scattering measurements on single crystals using a horizontally bent monochromator and a multidetecter; Statistical analysis of neutron diffraction studies of proteins. (orig.) [de

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

A multi-group neutron noise simulator for fast reactors

International Nuclear Information System (INIS)

Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

2013-01-01

Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

One-velocity neutron diffusion calculations based on a two-group reactor model

Energy Technology Data Exchange (ETDEWEB)

Bingulac, S; Radanovic, L; Lazarevic, B; Matausek, M; Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

1965-07-01

Many processes in reactor physics are described by the energy dependent neutron diffusion equations which for many practical purposes can often be reduced to one-dimensional two-group equations. Though such two-group models are satisfactory from the standpoint of accuracy, they require rather extensive computations which are usually iterative and involve the use of digital computers. In many applications, however, and particularly in dynamic analyses, where the studies are performed on analogue computers, it is preferable to avoid iterative calculations. The usual practice in such situations is to resort to one group models, which allow the solution to be expressed analytically. However, the loss in accuracy is rather great particularly when several media of different properties are involved. This paper describes a procedure by which the solution of the two-group neutron diffusion. equations can be expressed analytically in the form which, from the computational standpoint, is as simple as the one-group model, but retains the accuracy of the two-group treatment. In describing the procedure, the case of a multi-region nuclear reactor of cylindrical geometry is treated, but the method applied and the results obtained are of more general application. Another approach in approximate solution of diffusion equations, suggested by Galanin is applicable only in special ideal cases.

Mean energy polarized neutron source

International Nuclear Information System (INIS)

Aleshin, V.A.; Zaika, N.I.; Kolotyj, V.V.; Prokopenko, V.S.; Semenov, V.S.

1988-01-01

Physical bases and realization scheme of a pulsed source of polarized neutrons with the energy of up to 75 MeV are described. The source comprises polarized deuteron source, transport line, low-energy ion and axial injector to the accelerator, U-240 isochronous cyclotron, targets for polarized neutron production, accelerated deuteron transport line and flight bases. The pulsed source of fast neutrons with the energy of up to 75 MeV can provide for highly polarized neutron beams with the intensity by 2-3 orders higher than in the most perfect source of this range which allows one to perform various experiments with high efficiency and energy resolution. 9 refs.; 1 fig

Neutron energy measurement for practical applications

Science.gov (United States)

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

2018-03-01

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

Measurement and analysis of leakage neutron energy spectra around the Kinki University Reactor, UTR-KINKI

CERN Document Server

Ogawa, Y; Sagawa, H; Tsujimoto, T

2002-01-01

The highly sensitive cylindrical multi-moderator type neutron spectrometer was constructed for measurement of low level environmental neutrons. This neutron spectrometer was applied for the determination of leakage neutron energy spectra around the Kinki University Reactor. The analysis of the leakage neutron energy spectra was performed by MCNP Monte Carlo code. From the obtained results, the agreement between the MCNP predictions and the experimentally determined values is fairly good, which indicates the MCNP model is correctly simulating the UTR-KINKI.

BARC 75 - A 75 group neutron-photon coupled cross-section library with P5- anisotropic scattering matrices

International Nuclear Information System (INIS)

Garg, S.B.

1990-01-01

A 75 group neutron-photon coupled cross-section library has been developed for 42 reactor nuclides utilizing the basic cross-section files - ENDF/B-IV for neutrons and DLC-7F for photons. 50 neutron energy groups and gamma energy groups are included in this library which should be well suited to carry out safety, shielding and core physics studies of nuclear reactors based on fission or fusion processes. This library is also adequate for oil logging and mineral exploration investigations. (author). 11 refs., 3 tabs

Future possibilities with intermediate-energy neutron beams

International Nuclear Information System (INIS)

Brady, F.P.

1987-01-01

Future possibilities for using neutrons of intermediate energies (50 - 200 MeV) as a probe of the nucleus are discussed. Some of the recent thinking concerning a systematic approach for studying elastic and inelastic scattering of electrons and hadrons and the important role of medium- and intermediate-energy neutrons in such a programme is reviewed. The advantages of neutrons in this energy range over neutrons with lower energies and over intermediate-energy pions for determining nuclear-transition and ground state densities, and for distinguishing proton from neutron density (isovector sensitivity), are noted. The important role of (n,p) charge exchange reactions in nuclear excitation studies is also reviewed. Experimental methods for utilizing neutrons as probes in elastic, inelastic, and charge exchange studies at these energies are discussed

Intercomparison of high energy neutron personnel dosimeters

International Nuclear Information System (INIS)

McDonald, J.C.; Akabani, G.; Loesch, R.M.

1993-03-01

An intercomparison of high-energy neutron personnel dosimeters was performed to evaluate the uniformity of the response characteristics of typical neutron dosimeters presently in use at US Department of Energy (DOE) accelerator facilities. It was necessary to perform an intercomparison because there are no national or international standards for high-energy neutron dosimetry. The testing that is presently under way for the Department of Energy Laboratory Accreditation Program (DOELAP) is limited to the use of neutron sources that range in energy from about 1 keV to 2 MeV. Therefore, the high-energy neutron dosimeters presently in use at DOE accelerator facilities are not being tested effectively. This intercomparison employed neutrons produced by the 9 Be(p,n) 9 B interaction at the University of Washington cyclotron, using 50-MeV protons. The resulting neutron energy spectrum extended to a maximum of approximately 50-MeV, with a mean energy of about 20-MeV. Intercomparison results for currently used dosimeters, including Nuclear Type A (NTA) film, thermoluminescent dosimeter (TLD)-albedo, and track-etch dosimeters (TEDs), indicated a wide variation in response to identical doses of high-energy neutrons. Results of this study will be discussed along with a description of plans for future work

Neutron applications in materials for energy

CERN Document Server

Kearley, Gordon J

2015-01-01

Neutron Applications in Materials for Energy collects results and conclusions of recent neutron-based investigations of materials that are important in the development of sustainable energy. Chapters are authored by leading scientists with hands-on experience in the field, providing overviews, recent highlights, and case-studies to illustrate the applicability of one or more neutron-based techniques of analysis. The theme follows energy production, storage, and use, but each chapter, or section, can also be read independently, with basic theory and instrumentation for neutron scattering being

Time-grated energy-selected cold neutron radiography

International Nuclear Information System (INIS)

McDonald, T.E. Jr.; Brun, T.O.; Claytor, T.N.; Farnum, E.H.; Greene, G.L.; Morris, C.

1998-01-01

A technique is under development at the Los Alamos Neutron Science Center (LANSCE), Manuel Lujan Jr. Neutron Scattering Center (Lujan Center) for producing neutron radiography using only a narrow energy range of cold neutrons. The technique, referred to as Time-Gated Energy-Selected (TGES) neutron radiography, employs the pulsed neutron source at the Lujan Center with time of flight to obtain a neutron pulse having an energy distribution that is a function of the arrival time at the imager. The radiograph is formed on a short persistence scintillator and a gated, intensified, cooled CCD camera is employed to record the images, which are produced at the specific neutron energy range determined by the camera gate. The technique has been used to achieve a degree of material discrimination in radiographic images. For some materials, such as beryllium and carbon, at energies above the Bragg cutoff the neutron scattering cross section is relatively high while at energies below the Bragg cutoff the scattering cross section drops significantly. This difference in scattering characteristics can be recorded in the TGES radiography and, because the Bragg cutoff occurs at different energy levels for various materials, the approach can be used to differentiate among these materials. This paper outlines the TGES radiography technique and shows an example of radiography using the approach

Proton energy dependence of slow neutron intensity

International Nuclear Information System (INIS)

Teshigawara, Makoto; Harada, Masahide; Watanabe, Noboru; Kai, Tetsuya; Sakata, Hideaki; Ikeda, Yujiro

2001-01-01

The choice of the proton energy is an important issue for the design of an intense-pulsed-spallation source. The optimal proton beam energy is rather unique from a viewpoint of the leakage neutron intensity but no yet clear from the slow-neutron intensity view point. It also depends on an accelerator type. Since it is also important to know the proton energy dependence of slow-neutrons from the moderators in a realistic target-moderator-reflector assembly (TMRA). We studied on the TMRA proposed for Japan Spallation Neutron Source. The slow-neutron intensities from the moderators per unit proton beam power (MW) exhibit the maximum at about 1-2 GeV. At higher proton energies the intensity per MW goes down; at 3 and 50 GeV about 0.91 and 0.47 times as low as that at 1 GeV. The proton energy dependence of slow-neutron intensities was found to be almost the same as that of total neutron yield (leakage neutrons) from the same bare target. It was also found that proton energy dependence was almost the same for the coupled and decoupled moderators, regardless the different moderator type, geometry and coupling scheme. (author)

MPA-11: Materials Synthesis and Integrated Devices; Overview of an Applied Energy Group

Energy Technology Data Exchange (ETDEWEB)

Dattelbaum, Andrew Martin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-16

Our mission is to provide innovative and creative chemical synthesis and materials science solutions to solve materials problems across the LANL missions. Our group conducts basic and applied research in areas related to energy security as well as problems relevant to the Weapons Program.

Martian Neutron Energy Spectrometer (MANES)

Science.gov (United States)

Maurer, R. H.; Roth, D. R.; Kinnison, J. D.; Goldsten, J. O.; Fainchtein, R.; Badhwar, G.

2000-01-01

High energy charged particles of extragalactic, galactic, and solar origin collide with spacecraft structures and planetary atmospheres. These primaries create a number of secondary particles inside the structures or on the surfaces of planets to produce a significant radiation environment. This radiation is a threat to long term inhabitants and travelers for interplanetary missions and produces an increased risk of carcinogenesis, central nervous system (CNS) and DNA damage. Charged particles are readily detected; but, neutrons, being electrically neutral, are much more difficult to monitor. These secondary neutrons are reported to contribute 30-60% of the dose equivalent in the Shuttle and MIR station. The Martian atmosphere has an areal density of 37 g/sq cm primarily of carbon dioxide molecules. This shallow atmosphere presents fewer mean free paths to the bombarding cosmic rays and solar particles. The secondary neutrons present at the surface of Mars will have undergone fewer generations of collisions and have higher energies than at sea level on Earth. Albedo neutrons produced by collisions with the Martian surface material will also contribute to the radiation environment. The increased threat of radiation damage to humans on Mars occurs when neutrons of higher mean energy traverse the thin, dry Martian atmosphere and encounter water in the astronaut's body. Water, being hydrogeneous, efficiently moderates the high energy neutrons thereby slowing them as they penetrate deeply into the body. Consequently, greater radiation doses can be deposited in or near critical organs such as the liver or spleen than is the case on Earth. A second significant threat is the possibility of a high energy heavy ion or neutron causing a DNA double strand break in a single strike.

Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

Lee, Dong Hyuk; Shim, Hyung Jin

2015-01-01

The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

Neutrons and sustainable energy research

International Nuclear Information System (INIS)

Peterson, V.

2009-01-01

Full text: Neutron scattering is essential for the study of sustainable energy materials, including the areas of hydrogen research (such as its separation, storage, and use in fuel-cells) and energy transport (such as fuel-cell and battery materials). Researchers at the Bragg Institute address critical questions in sustainable energy research, with researchers providing a source of expertise for external collaborators, specialist analysis equipment, and acting as a point of contact for the study of sustainable energy materials using neutron scattering. Some recent examples of sustainable energy materials research using neutron scattering will be presented. These examples include the storage of energy, in the form of hydrogen through a study of its location in and interaction with new porous hydrogen storage materials [1-3] and in battery materials through in-situ studies of structure during charge-discharge cycling, and use of energy in fuel cells by studying proton diffusion through fuel cell membranes.

Improving the neutron-to-photon discrimination capability of detectors used for neutron dosimetry in high energy photon beam radiotherapy

International Nuclear Information System (INIS)

Irazola, L.; Terrón, J.A.; Bedogni, R; Pola, A.; Lorenzoli, M.; Sánchez-Nieto, B.; Gómez, F.; Sánchez-Doblado, F.

2016-01-01

The increasing interest of the medical community to radioinduced second malignancies due to photoneutrons in patients undergoing high-energy radiotherapy, has stimulated in recent years the study of peripheral doses, including the development of some dedicated active detectors. Although these devices are designed to respond to neutrons only, their parasitic photon response is usually not identically zero and anisotropic. The impact of these facts on measurement accuracy can be important, especially in points close to the photon field-edge. A simple method to estimate the photon contribution to detector readings is to cover it with a thermal neutron absorber with reduced secondary photon emission, such as a borated rubber. This technique was applied to the TNRD (Thermal Neutron Rate Detector), recently validated for thermal neutron measurements in high-energy photon radiotherapy. The positive results, together with the accessibility of the method, encourage its application to other detectors and different clinical scenarios. - Highlights: • Neutron-to-photon discrimination of a thermal neutron detector used in radiotherapy. • Photon and anisotropic response study with distance and beam incidence of thermal neutron detector. • Borated rubber for estimating photon contribution in any thermal neutron detector.

Thermal neutron group constants in monoatomic-gas approximation

Energy Technology Data Exchange (ETDEWEB)

Matausek, M V; Bosevski, T [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-12-15

To solve the problem of space-energy neutron distribution in an elementary reactor cell, a combination of the multigroup procedure and the P{sub 3} approximation of the spherical harmonics method was chosen. The calculation was divided into two independent parts: the first part was to provide multigroup constants which serve as input data for the second part - the determination of the slow neutron spectra. In the present report only the first part of the problem will be discussed. The velocity dependence of cross-sections and scattering function in thermal range was interpreted by the monoatomic-gas model. A digital computer program was developed for the evaluation of the group values for these quantities (author00.

Standardization activities of the Euratom Neutron Radiography Working Group

International Nuclear Information System (INIS)

Domanus, J.

1982-06-01

In 1979 a working group on neutron radiography was formed at Euratom. The purpose of this group is the standardization of neutron radiographic methods in the field of nuclear fuel. Activities of this Neutron Radiography Working Group are revised. Classification of defects revealed by neutron radiography is illustrated in a special atlas. Beam purity and sensitivity indicators are tested together with a special calibration fuel pin. All the Euratom neutron radiography centers will perform comparative neutron radiography with those items. The measuring results obtained, using various measuring aparatus will form the basis to formulate conclusions about the best measuring methods and instruments to be used in that field. Besides the atlas of neutron radiographic findings in light water reactor fuel, the Euratom Neutron Radiogrphy Working Group has published a neutron radiography handbook in which the neutron radiography installations in the European Community are also described. (author)

Semi-empirical neutron tool calibration (one and two-group approximation)

International Nuclear Information System (INIS)

Czubek, J.A.

1988-01-01

The physical principles of the new method of calibration of neutron tools for the rock porosity determination are given. A short description of the physics of neutron transport in the matter is presented together with some remarks on the elementary interactions of neutrons with nuclei (cross sections, group cross sections etc.). The definitions of the main integral parameters characterizing the neutron transport in the rock media are given. The three main approaches to the calibration problem: empirical, theoretical and semi-empirical are presented with some more detailed description of the latter one. The new semi-empirical approach is described. The method is based on the definition of the apparent slowing down or migration length for neutrons sensed by the neutron tool situated in the real borehole-rock conditions. To calculate this apparent slowing down or migration lengths the ratio of the proper space moments of the neutron distribution along the borehole axis is used. Theoretical results are given for one- and two-group diffusion approximations in the rock-borehole geometrical conditions when the tool is in the sidewall position. The physical and chemical parameters are given for the calibration blocks of the Logging Company in Zielona Gora. Using these data the neutron parameters of the calibration blocks have been calculated. An example, how to determine the calibration curve for the dual detector tool applying this new method and using the neutron parameters mentioned above together with the measurements performed in the calibration blocks, is given. The most important advantage of the new semi-empirical method of calibration is the possibility of setting on the unique calibration curve all experimental calibration data obtained for a given neutron tool for different porosities, lithologies and borehole diameters. 52 refs., 21 figs., 21 tabs. (author)

Comparison of Bonner sphere responses calculated by different Monte Carlo codes at energies between 1 MeV and 1 GeV – Potential impact on neutron dosimetry at energies higher than 20 MeV

CERN Document Server

Rühm, W; Pioch, C; Agosteo, S; Endo, A; Ferrarini, M; Rakhno, I; Rollet, S; Satoh, D; Vincke, H

2014-01-01

Bonner Spheres Spectrometry in its high-energy extended version is an established method to quantify neutrons at a wide energy range from several meV up to more than 1 GeV. In order to allow for quantitative measurements, the responses of the various spheres used in a Bonner Sphere Spectrometer (BSS) are usually simulated by Monte Carlo (MC) codes over the neutron energy range of interest. Because above 20 MeV experimental cross section data are scarce, intra-nuclear cascade (INC) and evaporation models are applied in these MC codes. It was suspected that this lack of data above 20 MeV may translate to differences in simulated BSS response functions depending on the MC code and nuclear models used, which in turn may add to the uncertainty involved in Bonner Sphere Spectrometry, in particular for neutron energies above 20 MeV. In order to investigate this issue in a systematic way, EURADOS (European Radiation Dosimetry Group) initiated an exercise where six groups having experience in neutron transport calcula...

Two-neutron “halo” from the low-energy limit of neutron–neutron interaction: Applications to drip-line nuclei 22C and 24O

Directory of Open Access Journals (Sweden)

Toshio Suzuki

2016-02-01

Full Text Available The formation of two-neutron “halo”, a low-density far-extended surface of weakly-bound two neutrons, is described using the neutron–neutron (nn interaction fixed at the low-energy nn scattering limit. This method is tested for loosely-bound two neutrons in 24O, where a good agreement with experimental data is found. It is applied to halo neutrons in 22C in two ways: with the 20C core being closed or correlated (due to excitations from the closed core. This nn interaction is shown to be strong enough to produce a two-neutron halo in both cases, locating 22C on the drip line, while 21C remains unbound. A unique relation between the two neutron separation energy, S2n, and the radius of neutron halo is presented. New predictions for S2n and the radius of neutron halo are given for 22C. The appearance of Efimov states is also discussed.

Neutron energy measurement for practical applications

Indian Academy of Sciences (India)

M V Roshan

2018-02-07

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

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

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.

Spectral correction factors for conventional neutron dosemeters used in high-energy neutron environments

International Nuclear Information System (INIS)

Lee, K.W.; Sheu, R.J.

2015-01-01

High-energy neutrons (>10 MeV) contribute substantially to the dose fraction but result in only a small or negligible response in most conventional moderated-type neutron detectors. Neutron dosemeters used for radiation protection purpose are commonly calibrated with 252 Cf neutron sources and are used in various workplace. A workplace-specific correction factor is suggested. In this study, the effect of the neutron spectrum on the accuracy of dose measurements was investigated. A set of neutron spectra representing various neutron environments was selected to study the dose responses of a series of Bonner spheres, including standard and extended-range spheres. By comparing 252 Cf-calibrated dose responses with reference values based on fluence-to-dose conversion coefficients, this paper presents recommendations for neutron field characterisation and appropriate correction factors for responses of conventional neutron dosemeters used in environments with high-energy neutrons. The correction depends on the estimated percentage of high-energy neutrons in the spectrum or the ratio between the measured responses of two Bonner spheres (the 4P6-8 extended-range sphere versus the 6'' standard sphere). (authors)

Calculated intensity of high-energy neutron beams

International Nuclear Information System (INIS)

Mustapha, B.; Nolen, J.A.; Back, B.B.

2004-01-01

The flux, energy and angular distributions of high-energy neutrons produced by in-flight spallation and fission of a 400 MeV/A 238 U beam and by the break-up of a 400 MeV/A deuteron beam are calculated. In both cases very intense secondary neutron beams are produced, peaking at zero degrees, with a relatively narrow energy spread. Such secondary neutron beams can be produced with the primary beams from the proposed rare isotope accelerator driver linac. The break-up of a 400 kW deuteron beam on a liquid-lithium target can produce a neutron flux of >10 10 neutrons/cm 2 /s at a distance of 10 m from the target

Report of the Working Group on low-temperature neutron irradiation

International Nuclear Information System (INIS)

1982-07-01

This report summarizes deliberations at a Working Group meeting sponsored by the Department of Energy, Division of Materials Sciences for the purpose of: (1) assessing the need for maintaining a low temperature neutron irradiation program in the United States; and (2) recommending a course of action based on this assessment

Very High Energy Neutron Scattering from Hydrogen

International Nuclear Information System (INIS)

Cowley, R A; Stock, C; Bennington, S M; Taylor, J; Gidopoulos, N I

2010-01-01

The neutron scattering from hydrogen in polythene has been measured with the direct time-of flight spectrometer, MARI, at the ISIS facility of the Rutherford Appleton Laboratory with incident neutron energies between 0.5 eV and 600 eV. The results of experiments using the spectrometer, VESUVIO, have given intensities from hydrogen containing materials that were about 60% of the intensity expected from hydrogen. Since VESUVIO is the only instrument in the world that routinely operates with incident neutron energies in the eV range we have chosen to measure the scattering from hydrogen at high incident neutron energies with a different type of instrument. The MARI, direct time-of-flight, instrument was chosen for the experiment and we have studied the scattering for several different incident neutron energies. We have learnt how to subtract the gamma ray background, how to calibrate the incident energy and how to convert the spectra to an energy plot . The intensity of the hydrogen scattering was independent of the scattering angle for scattering angles from about 5 degrees up to 70 degrees for at least 3 different incident neutron energies between 20 eV and 100 eV. When the data was put on an absolute scale, by measuring the scattering from 5 metal foils with known thicknesses under the same conditions we found that the absolute intensity of the scattering from the hydrogen was in agreement with that expected to an accuracy of ± 5.0% over a wide range of wave-vector transfers between 1 and 250 A -1 . These measurements show that it is possible to measure the neutron scattering with incident neutron energies up to at least 100 eV with a direct geometry time-of-flight spectrometer and that the results are in agreement with conventional scattering theory.

Neutron energy spectra calculations in the low power research reactor

International Nuclear Information System (INIS)

Omar, H.; Khattab, K.; Ghazi, N.

2011-01-01

The neutron energy spectra have been calculated in the fuel region, inner and outer irradiation sites of the zero power research reactor using the MCNP-4C code and the combination of the WIMS-D/4 transport code for generation of group constants and the three-dimensional CITATION diffusion code for core analysis calculations. The neutron energy spectrum has been divided into three regions and compared with the proposed empirical correlations. The calculated thermal and fast neutron fluxes in the low power research reactor MNSR inner and outer irradiation sites have been compared with the measured results. Better agreements have been noticed between the calculated and measured results using the MCNP code than those obtained by the CITATION code. (author)

Multisphere system neutron spectrometry applied to dosimetry for the personnel

International Nuclear Information System (INIS)

Allinei, P.G.

1992-01-01

Neutron dosimetry is a necessity that must be dealt with in order to ensure efficient monitoring of all personnel regarding radiology safety. Dosimetric variables are difficult to measure for they are dependent on complex functions evolving with the energy of neutrons, which forces us to determine their energetic distribution. We have chosen to use the multisphere system associated to an unfolding code in order to perform neutron spectrometry, our purpose being to determine these dosimetric variables. The initial stage consists in modifying a research code, the code SOHO, in order to adapt it to our needs. The resulting new version was subsequently tested and proven successful by means of computerized simulations. Afterwards, we used reference dosimetric and spectral beams to confirm the position results previously obtained. At the time of this test, the code SOHO yielded results coherent with the theoretical values, and even allowed the quantity of radiation diffused by the laboratory structures to be estimated. The final part of this study consists in applying the previously perfected technique to authentic situations. The results thus obtained are compared to those obtained by conventional methods in order to reveal the interest of neutron spectrometry used for dosimetry of the personnel

Comparison of pressure vessel neutron fluences for the Balakovo-3 reactor with measurements and investigation of the influence of neutron cross sections and number of groups on the results

Energy Technology Data Exchange (ETDEWEB)

Barz, H U; Boehmer, B; Konheiser, J; Stephan, I

1998-10-01

The general methodical questions of experimental and theoretical determination of neutron fluences have been described in connection with the measurements and 3-D Monte Carlo calculation for the Rovno-3 reactor. The same calculation and measurement methods were applied for the Balakovo-3 reactor. In the first part, the results of the comparison for Balakovo will be given and discussed. However, for this reactor the main attention was focussed on investigations of the accuracy of the calculation. In this connection an important question is the influence of neutron data on the results. With this respect not only the source of the data but also the number of energy groups is important. (orig.)

Quasi-energy of ultracold neutrons

International Nuclear Information System (INIS)

Frank, A.I.; Nosov, V.G.

1992-01-01

A solution is found to the problem of the propagation of a neutron beam transmitted through a periodically acting high-speed chopper. It is a generalization of the Moshinsky's problem of the evolution of a plane wave in the right half-space after an ideal absorber at the origin of coordinates has been instantaneously removed. The energy spectrum of transmitted neutrons is found to be discrete and corresponding to their quasi-energy. Interference of the states corresponding to different satellite lines leads to a complex spatial pattern with typical beats. A number of experiments with ultracold neutrons are suggested and discussed. 12 refs.; 1 fig

Moderation of neutron energy

International Nuclear Information System (INIS)

Marlatt, G.R.

1986-01-01

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

Free neutron-proton analyzing power at medium energies

International Nuclear Information System (INIS)

Newsom, C.R.

1980-01-01

In recent years, increasing efforts have been made to measure the nucleon-nucleon polarization parameters. To date, no free neutron-proton spin correlated parameters have been published in the energy range 500 to 800 MeV. Existing analyzing power data is of low precision and in most cases was obtained by quasi-free proton scattering. As a first step in determining the neutron-proton scattering matrix, the free neutron-proton analyzing power has been measured at the Los Alamos Physics Facility as a function of energy and angle. The experiment was performed by scattering a neutron beam from a polarized proton target. The neutron beam was generated by scattering 800 MeV protons from a Beryllium target and using the neutrons produced at 0 degrees. The incident energy ranged from 300 MeV to 800 MeV. The energy spread of the neutron beam made it possible to measure the analyzing power at different energies simultaneously. Angular distributions were taken from 60 to 170 degrees in the center of mass system (c.m.)

Recent applications of neutron activation analysis at Lucas Heights

International Nuclear Information System (INIS)

Fardy, J.J.

1978-01-01

The use of neutron activation analysis to determine key elemental distribution patterns in samples from both the energy industry and health science field is summarised. Instrumental neutron activation analysis has successfully measured simultaneously more than twenty elements in a sample of brown coal from Victoria, black coal from New South Wales and samples from the product stream of ACIRL's batch autoclave, solvent-refined, coal hydrogenation process. Four gallstones removed from the same gallbladder have been examined instrumentally by neutron activation analysis. A total of sixteen trace elements were detected with concentrations in the range 0.8 ng g -1 for gold to 7,800 μg g -1 for calcium

A neutron irradiator applied to cancer treatment

International Nuclear Information System (INIS)

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

2000-01-01

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

Low energy neutron scattering for energy dependent cross sections. General considerations

Energy Technology Data Exchange (ETDEWEB)

Rothenstein, W; Dagan, R [Technion-Israel Inst. of Tech., Haifa (Israel). Dept. of Mechanical Engineering

1996-12-01

We consider in this paper some aspects related to neutron scattering at low energies by nuclei which are subject to thermal agitation. The scattering is determined by a temperature dependent joint scattering kernel, or the corresponding joint probability density, which is a function of two variables, the neutron energy after scattering, and the cosine of the angle of scattering, for a specified energy and direction of motion of the neutron, before the interaction takes place. This joint probability density is easy to calculate, when the nucleus which causes the scattering of the neutron is at rest. It can be expressed by a delta function, since there is a one to one correspondence between the neutron energy change, and the cosine of the scattering angle. If the thermal motion of the target nucleus is taken into account, the calculation is rather more complicated. The delta function relation between the cosine of the angle of scattering and the neutron energy change is now averaged over the spectrum of velocities of the target nucleus, and becomes a joint kernel depending on both these variables. This function has a simple form, if the target nucleus behaves as an ideal gas, which has a scattering cross section independent of energy. An energy dependent scattering cross section complicates the treatment further. An analytic expression is no longer obtained for the ideal gas temperature dependent joint scattering kernel as a function of the neutron energy after the interaction and the cosine of the scattering angle. Instead the kernel is expressed by an inverse Fourier Transform of a complex integrand, which is averaged over the velocity spectrum of the target nucleus. (Abstract Truncated)

Application of the variational method for calculation of neutron spectra and group constants - Master thesis

International Nuclear Information System (INIS)

Milosevic, M.

1979-01-01

One-dimensional variational method for cylindrical configuration was applied for calculating group constants, together with effects of elastic slowing down, anisotropic elastic scattering, inelastic scattering, heterogeneous resonance absorption with the aim to include the presence of a number of different isotopes and effects of neutron leakage from the reactor core. Neutron flux shape P 3 and adjoint function are proposed in order to enable calculation of smaller size reactors and inclusion of heterogeneity effects by cell calculations. Microscopic multigroup constants were prepared based on the UKNDL data library. Analytical-numerical approach was applied for solving the equations of the P 3 approximation to obtain neutron flux moments and adjoint functions

MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors

International Nuclear Information System (INIS)

Camiciola, P.; Cundari, D.; Montagnini, B.

1992-01-01

1 - Description of program or function: The program solves the 1-D time-dependent one and two group coarse-mesh neutron diffusion equations, coupled with the equations for the delayed-neutron precursor, in plane geometry. 2 - Method of solution: The program is based on a simple coarse-mesh cubic approximation formula for the spatial behaviour of the flux inside each interval. An implicit scheme (the time-integrated method) is used for the advancement of the solution. The resulting (block three-diagonal) matrix is inverted at each time step by Thomas' method. 3 - Restrictions on the complexity of the problem: Number of coarse- mesh intervals LE 80; number of material regions LE 10; number of delayed-neutron precursor groups LE 10. Typical mesh sizes range from 5 cm to 20 cm; typical step length (non-prompt critical transients) ranges from 0.005 to 0.1 seconds

Calculating disadvantage factor for fuel taking into account the neutron energy distribution

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1964-01-01

Errors in calculating the disadvantage factor are caused by applying the diffusion approximation and one-group method. This paper describes the method for calculating the fuel disadvantage factor by applying a non-diffusion method taking into account neutron thermalization

Neutrons in basic and applied nuclear research - a review

International Nuclear Information System (INIS)

Bhattacharya, Sailajananda

2013-01-01

Energetic neutron sources, both white and mono-energetic, are widely used In basic nuclear physics as well as various multidisciplinary research. Precise measurement of various neutron induced reaction cross-sections are crucial for the design and development of new generation of reactors, like accelerator driven subcritical systems, nuclear incinerators, etc. A review of some recent trends in neutron induced basic and applied nuclear research will be presented in this talk. (author)

Investigation of the 232Th neutron cross-sections in resonance energy range

International Nuclear Information System (INIS)

Grigoriev, Yu.V.; Kitaev, V.Ya.; Sinitsa, V.V.; Zhuravlev, B.V.; Borzakov, S.B.; Faikov-Stanchik, H.; Ilchev, G.L.; Panteleev, Ts.Ts.; Kim, G.N.

2001-01-01

The alternative path in the development of atomic energy is the uranium-thorium cycle. In connection with this, the measurements of the 232 Th neutron capture and total cross-sections and its resonance self-shielding coefficients in resonance energy range are necessary because of their low accuracy. In this work, the results of the investigations of the thorium-232 neutron cross-sections are presented. The measurements have been carried out on the gamma-ray multisection liquid detector and neutron detector as a battery of boron counters on the 120 m flight path of the pulsed fast reactor IBR-30. As the filter samples were used the metallic disks of various thickness and diameter of 45 mm. Two plates from metallic thorium with thickness of 0.2 mm and with the square of 4.5x4.5 cm 2 were used as the radiator samples. The group neutron total and capture cross-sections within the accuracy of 2-7% in the energy range of (10 eV-10 keV) were obtained from the transmissions and the sum spectra of g-rays from the fourth multiplicity to the seventh one. The neutron capture group cross-sections of 238 U were used as the standard for obtaining of thorium ones. Analogous values were calculated on the GRUCON code with the ENDF/B-6, JENDL-3 evaluated data libraries. Within the limits of experimental errors an agreement between the experiment and calculation is observed, but in some groups the experimental values are larger than the calculated ones. (author)

Neutron generators and their uses in research and applied fields. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

Asfour, F I [Division of Basic Nuclear Sciences, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

The development of the low voltage neutron generators (NGS) has contributed considerably to the scope of nuclear research and the economical application of nuclear methods. Such simple instruments are used to produce 14 MeV and 3 MeV neutrons via the 3{sup H}(d,n)4{sup H}e and 2{sup H}(d,n)3{sup H}e reactions,respectively. The neutrons are very widely used and are inexpensive, easy to install and operate, therefore, in addition to nuclear physicists, there are a number of groups of scientists who use low voltage accelerators as tools for pure and applied research, service and education. The aim of this work is to review shortly those problems and methods of science and technology where the neutrons produced in the D-T and D-D reactions play the main role. A wide range of experiments with the detection of neutrons and charged particles is available including the study of shielding and the generator technology itself. N.G. are recently widely used for the determination of neutron data needed for fast reactor and thermonuclear devices. The principles and techniques of the possible uses of neutron generators in technology and research are summarized. The review is devoted to:- Give a short review of the most important operational characteristics of the neutron generators and the necessary instruments needed for application. Outline the main applications of the neutron generators in neutron activation and prompt radiation analysis in various fields(metallurgy, chemistry, biology, meteoritic and lunar studies, geology and mining, etc...) fast neutron therapy, and radiation effects. 2 figs.

An analytical solution for the two-group kinetic neutron diffusion equation in cylindrical geometry

International Nuclear Information System (INIS)

Fernandes, Julio Cesar L.; Vilhena, Marco Tullio; Bodmann, Bardo Ernst

2011-01-01

Recently the two-group Kinetic Neutron Diffusion Equation with six groups of delay neutron precursor in a rectangle was solved by the Laplace Transform Technique. In this work, we report on an analytical solution for this sort of problem but in cylindrical geometry, assuming a homogeneous and infinite height cylinder. The solution is obtained applying the Hankel Transform to the Kinetic Diffusion equation and solving the transformed problem by the same procedure used in the rectangle. We also present numerical simulations and comparisons against results available in literature. (author)

Production of low energy gamma rays by neutron interactions with fluorine for incident neutron energies between 0.1 and 20 MeV

International Nuclear Information System (INIS)

Morgan, G.L.; Dickens, J.K.

1975-06-01

Differential cross sections for the production of low-energy gamma rays (less than 240 keV) by neutron interactions in fluorine have been measured for neutron energies between 0.1 and 20 MeV. The Oak Ridge Electron Linear Accelerator was used as the neutron source. Gamma rays were detected at 92 0 using an intrinsic germanium detector. Incident neutron energies were determined by time-of-flight techniques. Tables are presented for the production cross sections of three gamma rays having energies of 96, 110, and 197 keV. (14 figures, 3 tables) (U.S.)

Neutron-capture cross-section measurement for 163Dy In the neutron energy range from 15 to 75 keV

International Nuclear Information System (INIS)

Kim, Hyun Duk; Jung, Eui Jung; Ahn, Jung Keun; Lee, Dae Won; Kim, Guin Yun; Ro, Tae Ik; Min, Young Ki; Igashira, Masayuki; Ohsaki, Toshiro; Mizuno, Satoshi

2002-01-01

The neutron-capture cross-section of 163 Dy were measured in the neutron energy range from 15 to 75 keV at the 3-MV Pelletron accelerator of the Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology. Pulsed neutrons were produced from the 7 Li(p,n) 7 Be reaction by bombarding a metallic lithium target with the 1.903-MeV proton beam. The incident neutron spectra were measured by means of a neutron time-of-flight method with a 6 Li-glass detector. Capture γ-rays were detected with a large anti-Compton NaI(Tl) spectrometer. A pulse-height weighting technique was applied to the capture γ-ray pulse-height spectra to obtain capture yields. The neutron capture cross-section were determined relative to the standard capture cross-sections of 197 Au. The present results were compared with the previous measurements and the evaluated values of ENDF/B-VI

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.

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

Inter-comparison of High Energy Files (neutron-induced, from 20 to 150 MeV)

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Ouk; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-11-01

Recent new applications using accelerator-driven system require well-tested nuclear data when modeling the interaction of neutrons above 20 MeV. This work is aimed to review evaluation methods applied in currently available neutron high energy files above 20 to 150 MeV, to inter-compare their evaluated cross sections on some important isotopes, and to analyze resulting discrepancies. Through out these, integrities and consistencies of the high energy files are checked, applicability of physics models and evaluation methodologies are assessed, and some directions are derived to improve and expand current JENDL High Energy File. (author)

Intermediate-energy neutron beams from reactors for NCT

International Nuclear Information System (INIS)

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

1986-01-01

This paper discusses ways that a beam of intermediate-energy neutrons might be extracted from a nuclear reactor. The challenge is to suppress the fast-neutron component and the gamma-ray component of the flux while leaving enough of the intermediate-energy neutrons in the beam to be able to perform neutron capture therapy in less than an hour exposure time. Moderators, filters, and reflectors are considered. 11 references, 7 figures, 3 tables

CERN-group conceptual design of a fast neutron operated high power energy amplifier

International Nuclear Information System (INIS)

Rubbia, C.; Rubio, J.A.; Buono, S.

1997-01-01

The practical feasibility of an Energy Amplifier (EA) with power and power density which are comparable to the ones of the present generation of large PWR is discussed in this paper. This is only possible with fast neutrons. Schemes are described which offer a high gain, a large maximum power density and an extended burn-up, well in excess of 100 GW x d/t corresponding to about five years at full power operation with no intervention on the fuel core. The following topics are discussed: physics considerations and parameter definition, the accelerator complex, the energy amplifier unit, computer simulated operation, and fuel cycle closing

CERN-group conceptual design of a fast neutron operated high power energy amplifier

Energy Technology Data Exchange (ETDEWEB)

Rubbia, C; Rubio, J A [European Organization for Nuclear Research, CERN, Geneva (Switzerland); Buono, S [Laboratoire du Cyclotron, Nice (France); and others

1997-11-01

The practical feasibility of an Energy Amplifier (EA) with power and power density which are comparable to the ones of the present generation of large PWR is discussed in this paper. This is only possible with fast neutrons. Schemes are described which offer a high gain, a large maximum power density and an extended burn-up, well in excess of 100 GW x d/t corresponding to about five years at full power operation with no intervention on the fuel core. The following topics are discussed: physics considerations and parameter definition, the accelerator complex, the energy amplifier unit, computer simulated operation, and fuel cycle closing. 84 refs, figs, tabs.

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-08-01

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

Neutrons for global energy solutions. Book of abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-07-01

The book of abstracts of the conference on neutrons for global energy solutions include contributions to the following topics: Views from politics: What do we need in European energy research: cooperation, large facilities, more science? Fundamental research for energy supply. View from the United States. View from industry: Neutrons for nuclear reactor development in transition stage between generation III and generation IV. Toyotas's expectations for neutron analysis. Instrumentation and cross cutting issues. Energy sources. Waste management and environment. Li ion batteries. Photovoltaics. Savings and catalysis. Fuel cells. Hydrogen storage.

Neutrons for global energy solutions. Book of abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-07-01

The book of abstracts of the conference on neutrons for global energy solutions include contributions to the following topics: Views from politics: What do we need in European energy research: cooperation, large facilities, more science? Fundamental research for energy supply. View from the United States. View from industry: Neutrons for nuclear reactor development in transition stage between generation III and generation IV. Toyotas's expectations for neutron analysis. Instrumentation and cross cutting issues. Energy sources. Waste management and environment. Li ion batteries. Photovoltaics. Savings and catalysis. Fuel cells. Hydrogen storage.

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.

A general formula considering one group delayed neutron under nonequilibrium condition

International Nuclear Information System (INIS)

Li Haofeng; Chen Wenzhen; Zhu Qian; Luo Lei

2008-01-01

A general neutron breeder formula is developed when the reactor does not reach the steady state and the reactivity changes in phase. This formula can be used to calculate the results of six groups delayed neutron model through a way of amending λ in one group delayed neutron model. The analysis shows that the solution of amended single group delayed neutron model is approximately equal to that of six-group delayed neutron model, and the amended model meets the engineering accuracy. (authors)

Experience in developing and using the VITAMIN-C 171-neutron, 36-gamma-ray group cross-section library

International Nuclear Information System (INIS)

Roussin, R.W.; Weisbin, C.R.; White, J.E.; Wright, R.Q.; Greene, N.M.; Ford, W.E. III; Wright, J.B.; Diggs, B.R.

1978-01-01

The Department of Energy (DOE) Division of Magnetic Fusion Energy (DMFE) and Reactor Research and Technology (DRRT) jointly sponsored the development of a coupled, fine-group cross-section library. The 171-neutron, 36-gamma-ray group library is intended to be applicable to fusion reactor neutronics and LMFBR core and shield analysis. Versions of the library are available from the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory in both AMPX and CCCC formats. Computer codes for energy group collapsing, interpolation on Bondarenko factors for resonance self-shielding and temperature corrections, and various other useful data manipulations are available. The experience gained in the utilization of this library is discussed. Indications are that this venture, which is designed to allow users to derive problem-dependent cross sections from a fine-group master library, has been a success

A Kinematically Beamed, Low Energy Pulsed Neutron Source for Active Interrogation

International Nuclear Information System (INIS)

Dietrich, D.; Hagmann, C.; Kerr, P.; Nakae, L.; Rowland, M.; Snyderman, N.; Stoeffl, W.; Hamm, R.

2004-01-01

We are developing a new active interrogation system based on a kinematically focused low energy neutron beam. The key idea is that one of the defining characteristics of SNM (Special Nuclear Materials) is the ability for low energy or thermal neutrons to induce fission. Thus by using low energy neutrons for the interrogation source we can accomplish three goals, (1) Energy discrimination allows us to measure the prompt fast fission neutrons produced while the interrogation beam is on; (2) Neutrons with an energy of approximately 60 to 100 keV do not fission 238U and Thorium, but penetrate bulk material nearly as far as high energy neutrons do and (3) below about 100keV neutrons lose their energy by kinematical collisions rather than via the nuclear (n,2n) or (n,n') processes thus further simplifying the prompt neutron induced background. 60 keV neutrons create a low radiation dose and readily thermal capture in normal materials, thus providing a clean spectroscopic signature of the intervening materials. The kinematically beamed source also eliminates the need for heavy backward and sideway neutron shielding. We have designed and built a very compact pulsed neutron source, based on an RFQ proton accelerator and a lithium target. We are developing fast neutron detectors that are nearly insensitive to the ever-present thermal neutron and neutron capture induced gamma ray background. The detection of only a few high energy fission neutrons in time correlation with the linac pulse will be a clear indication of the presence of SNM

Thermal neutron diffusion parameters dependent on the flux energy distribution in finite hydrogenous media

International Nuclear Information System (INIS)

Drozdowicz, K.

1999-01-01

Macroscopic parameters for a description of the thermal neutron transport in finite volumes are considered. A very good correspondence between the theoretical and experimental parameters of hydrogenous media is attained. Thermal neutrons in the medium possess an energy distribution, which is dependent on the size (characterized by the geometric buckling) and on the neutron transport properties of the medium. In a hydrogenous material the thermal neutron transport is dominated by the scattering cross section which is strongly dependent on energy. A monoenergetic treatment of the thermal neutron group (admissible for other materials) leads in this case to a discrepancy between theoretical and experimental results. In the present paper the theoretical definitions of the pulsed thermal neutron parameters (the absorption rate, the diffusion coefficient, and the diffusion cooling coefficient) are based on Nelkin's analysis of the decay of a neutron pulse. Problems of the experimental determination of these parameters for a hydrogenous medium are discussed. A theoretical calculation of the pulsed parameters requires knowledge of the scattering kernel. For thermal neutrons it is individual for each hydrogenous material because neutron scattering on hydrogen nuclei bound in a molecule is affected by the molecular dynamics (characterized with internal energy modes which are comparable to the incident neutron energy). Granada's synthetic model for slow-neutron scattering is used. The complete up-dated formalism of calculation of the energy transfer scattering kernel after this model is presented in the paper. An influence of some minor variants within the model on the calculated differential and integral neutron parameters is shown. The theoretical energy-dependent scattering cross section (of Plexiglas) is compared to experimental results. A particular attention is paid to the calculation of the diffusion cooling coefficient. A solution of an equation, which determines the

Symmetry energy, unstable nuclei and neutron star crusts

Energy Technology Data Exchange (ETDEWEB)

Iida, Kei [Kochi University, Department of Natural Science, Kochi (Japan); RIKEN Nishina Center, Saitama (Japan); Oyamatsu, Kazuhiro [RIKEN Nishina Center, Saitama (Japan); Aichi Shukutoku University, Department of Human Informatics, Aichi (Japan)

2014-02-15

The phenomenological approach to inhomogeneous nuclear matter is useful to describe fundamental properties of atomic nuclei and neutron star crusts in terms of the equation of state of uniform nuclear matter. We review a series of researches that we have developed by following this approach. We start with more than 200 equations of state that are consistent with empirical masses and charge radii of stable nuclei and then apply them to describe matter radii and masses of unstable nuclei, proton elastic scattering and total reaction cross sections off unstable nuclei, and nuclei in neutron star crusts including nuclear pasta. We finally discuss the possibility of constraining the density dependence of the symmetry energy from experiments on unstable nuclei and even observations of quasi-periodic oscillations in giant flares of soft gamma-ray repeaters. (orig.)

Interpretation of active neutron measurements by the heterogeneous theory

International Nuclear Information System (INIS)

Birkhoff, G.; Depraz, J.; Descieux, J.P.

1979-01-01

In this paper are presented results from a study on the application of the heterogeneous method for the interpretation of active neutron measurements. The considered apparatus consists out of a cylindrical lead pile, which is provided with two axial channels: a central channel incorporates an antimony beryllium photoneutron source and an excentric channel serves for the insertion of the sample to be assayed for fissionable materials contents. The mathematical model of this apparatus is the heterogeneous group diffusion theory. Sample and source channel are described by multigroup monopolar and dipolar sources and sinks. Monopolar sources take account of neutron production within energy group and in-scatter from upper groups. Monopolar sinks represent neutron removal by absorption within energy group and outscatter to lower groups. Dipol sources describe radial streaming of neutrons across the sample channel. Multigroup diffusion theory is applied throughout the lead pile. The strengths of the monopolar and dipolar sources and sinks are determined by linear extrapolation distances of azimuthal mean and first harmonic flux values at the channels' surface. In an experiment we may measure the neutrons leaking out of the lead pile and linear extrapolation distances at the channels' surface. Such informations are utilized for interpretation in terms of fission neutron source strengh and mean neutron flux values in the sample. In this paper we summarized the theoretical work in course

On the origin of low energy tail for monoenergetic neutron sources

International Nuclear Information System (INIS)

Kornilov, N.V.; Kagalenko, A.B.

1995-01-01

The problems of data processing when measuring inelastic neutron scattering cross sections for separated nuclei levels are studied. The model describing the neutron energy distribution for monoenergetic neutron sources is developed. The factors which make the major contributions into spectrometer response function formation are discussed. It is shown that the model considered predicts well neutron energy distribution from metal Li-target. The model parameters should be estimated on the basis of the experimental data. The neutron scattering on target environment contributes much into the low energy region of the neutron spectrum. An additional neutron source is introduced into the model in order to describe the low energy peak asymmetry (the so-called low energy tail). The tail neutron contribution dependence on incident energy and angle turns out to be rather unexpected. The conclusion is made that it is difficult to explain the origin and the properties of the tail neutron source by slit proton scattering or some Li-nuclei distribution regularities. 3 refs., 6 figs

Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

1986-02-01

Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs

Energy response study of modified CR-39 neutron personnel dosimeter

International Nuclear Information System (INIS)

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

2018-01-01

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

Measurements of the energy spectrum of backscattered fast neutrons

International Nuclear Information System (INIS)

Segal, Y.

1976-03-01

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

Estimation of neutron energy distributions from prompt gamma emissions

Science.gov (United States)

Panikkath, Priyada; Udupi, Ashwini; Sarkar, P. K.

2017-11-01

A technique of estimating the incident neutron energy distribution from emitted prompt gamma intensities from a system exposed to neutrons is presented. The emitted prompt gamma intensities or the measured photo peaks in a gamma detector are related to the incident neutron energy distribution through a convolution of the response of the system generating the prompt gammas to mono-energetic neutrons. Presently, the system studied is a cylinder of high density polyethylene (HDPE) placed inside another cylinder of borated HDPE (BHDPE) having an outer Pb-cover and exposed to neutrons. The emitted five prompt gamma peaks from hydrogen, boron, carbon and lead can be utilized to unfold the incident neutron energy distribution as an under-determined deconvolution problem. Such an under-determined set of equations are solved using the genetic algorithm based Monte Carlo de-convolution code GAMCD. Feasibility of the proposed technique is demonstrated theoretically using the Monte Carlo calculated response matrix and intensities of emitted prompt gammas from the Pb-covered BHDPE-HDPE system in the case of several incident neutron spectra spanning different energy ranges.

Solution of the neutron point kinetics equations with temperature feedback effects applying the polynomial approach method

International Nuclear Information System (INIS)

Tumelero, Fernanda; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana

2015-01-01

In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)

Solution of the neutron point kinetics equations with temperature feedback effects applying the polynomial approach method

Energy Technology Data Exchange (ETDEWEB)

Tumelero, Fernanda, E-mail: fernanda.tumelero@yahoo.com.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana, E-mail: claudiopeteren@yahoo.com.br, E-mail: gleniogoncalves@yahoo.com.br, E-mail: luana-lazzari@hotmail.com [Universidade Federal de Pelotas (DME/UFPEL), Capao do Leao, RS (Brazil). Instituto de Fisica e Matematica

2015-07-01

In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)

Energy measurement of prompt fission neutrons in 239Pu(n,f) for incident neutron energies from 1 to 200 MeV

CERN Document Server

Chatillon, A; Granier, Th; Laurent, B; Taïeb, J; Noda, S; Haight, R C; Devlin, M; Nelson, R O; O’Donnell, J M

2010-01-01

Prompt fission neutron spectra in the neutron-induced fission of 239Pu have been measured for incident neutron energies from 1 to 200 MeV at the Los Alamos Neutron Science Center. Preliminary results are discussed and compared to theoretical model calculation.

New neutron physics using spallation sources

International Nuclear Information System (INIS)

Bowman, C.D.

1988-01-01

The extraordinary neutron intensities available from the new spallation pulsed neutron sources open up exciting opportunities for basic and applied research in neutron nuclear physics. The energy range of neutron research which is being explored with these sources extends from thermal energies to almost 800 MeV. The emphasis here is on prospective experiments below 100 keV neutron energy using the intense neutron bursts produced by the Proton Storage Ring (PSR) at Los Alamos. 30 refs., 10 figs

The universal library of fission products and delayed neutron group yields

International Nuclear Information System (INIS)

Koldobskiy, A.B.; Zhivun, V.M.

1997-01-01

A new fission product yield library based on the Semiempirical method for the estimation of their mass and charge distribution is described. Contrary to other compilations, this library can be used with all possible excitation energies of fissionable actinides. The library of delayed neutron group yields, based on the fission product yield compilation, is described as well. (author). 15 refs, 4 tabs

On the idea of low-energy nuclear reactions in metallic lattices by producing neutrons from protons capturing "heavy" electrons

Science.gov (United States)

Tennfors, Einar

2013-02-01

The present article is a critical comment on Widom and Larsens speculations concerning low-energy nuclear reactions (LENR) based on spontaneous collective motion of protons in a room temperature metallic hydride lattice producing oscillating electric fields that renormalize the electron self-energy, adding significantly to the effective electron mass and enabling production of low-energy neutrons. The frequency and mean proton displacement estimated on the basis of neutron scattering from protons in palladium and applied to the Widom and Larsens model of the proton oscillations yield an electron mass enhancement less than one percent, far below the threshold for the proposed neutron production and even farther below the mass enhancement obtained by Widom and Larsen assuming a high charge density. Neutrons are not stopped by the Coulomb barrier, but the energy required for the neutron production is not low.

Neutron resonance transmission spectroscopy with high spatial and energy resolution at the J-PARC pulsed neutron source

Energy Technology Data Exchange (ETDEWEB)

Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California at Berkeley, 7 Gauss Way, Berkeley, CA 94720 (United States); Shinohara, T.; Kai, T.; Ooi, M. [Japan Atomic Energy Agency, 2–4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kamiyama, T.; Kiyanagi, Y.; Shiota, Y. [Hokkaido University, Kita 13 Nishi 8 Kita-ku, Sapporo-shi, Hokkaido 060-8628 (Japan); McPhate, J.B.; Vallerga, J.V.; Siegmund, O.H.W. [University of California at Berkeley, 7 Gauss Way, Berkeley, CA 94720 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

2014-05-11

The sharp variation of neutron attenuation at certain energies specific to particular nuclides (the lower range being from ∼1 eV up to ∼1 keV), can be exploited for the remote mapping of element and/or isotope distributions, as well as temperature probing, within relatively thick samples. Intense pulsed neutron beam-lines at spallation sources combined with a high spatial, high-timing resolution neutron counting detector, provide a unique opportunity to measure neutron transmission spectra through the time-of-flight technique. We present the results of experiments where spatially resolved neutron resonances were measured, at energies up to 50 keV. These experiments were performed with the intense flux low background NOBORU neutron beamline at the J-PARC neutron source and the high timing resolution (∼20 ns at epithermal neutron energies) and spatial resolution (∼55 µm) neutron counting detector using microchannel plates coupled to a Timepix electronic readout. Simultaneous element-specific imaging was carried out for several materials, at a spatial resolution of ∼150 µm. The high timing resolution of our detector combined with the low background beamline, also enabled characterization of the neutron pulse itself – specifically its pulse width, which varies with neutron energy. The results of our measurements are in good agreement with the predicted results for the double pulse structure of the J-PARC facility, which provides two 100 ns-wide proton pulses separated by 600 ns, broadened by the neutron energy moderation process. Thermal neutron radiography can be conducted simultaneously with resonance transmission spectroscopy, and can reveal the internal structure of the samples. The transmission spectra measured in our experiments demonstrate the feasibility of mapping elemental distributions using this non-destructive technique, for those elements (and in certain cases, specific isotopes), which have resonance energies below a few keV, and with lower

Dipole strength in 80Se below the neutron-separation energy for the nuclear transmutation of 79Se

Directory of Open Access Journals (Sweden)

Makinaga Ayano

2017-01-01

Full Text Available The γ-ray strength function (γSF in 80Se is an important parameter to estimate the neutron-capture cross section of 79Se which is one of the long-lived fission products (LLFPs. Until now, the γSF method was applied for 80Se only above the neutron-separation energy (Sn and the evaluated 79Se(n,γ cross section has an instability caused by the GSF below Sn. We studied the dipole-strength distribution of 80Se in a photon-scattering experiment using bremsstrahlung produced by an electron beam of an energy of 11.5 MeV at the linear accelerator ELBE at HZDR. The present photoabsorption cross section of 80Se was combined with results of (γ,n experiments and are compared with predictions usinmg the TALYS code. We also estimated the 79Se(n,γ cross sections and compare them with TALYS predictionms and earlier work by other groups.

Dipole strength in 80Se below the neutron-separation energy for the nuclear transmutation of 79Se

Science.gov (United States)

Makinaga, Ayano; Massarczyk, Ralph; Beard, Mary; Schwengner, Ronald; Otsu, Hideaki; Müller, Stefan; Röder, Marko; Schmidt, Konrad; Wagner, Andreas

2017-09-01

The γ-ray strength function (γSF) in 80Se is an important parameter to estimate the neutron-capture cross section of 79Se which is one of the long-lived fission products (LLFPs). Until now, the γSF method was applied for 80Se only above the neutron-separation energy (Sn) and the evaluated 79Se(n,γ) cross section has an instability caused by the GSF below Sn. We studied the dipole-strength distribution of 80Se in a photon-scattering experiment using bremsstrahlung produced by an electron beam of an energy of 11.5 MeV at the linear accelerator ELBE at HZDR. The present photoabsorption cross section of 80Se was combined with results of (γ,n) experiments and are compared with predictions usinmg the TALYS code. We also estimated the 79Se(n,γ) cross sections and compare them with TALYS predictionms and earlier work by other groups.

Self-energy dispersion effects on neutron matter superfluidity

International Nuclear Information System (INIS)

Zuo Wei

2001-01-01

The effects of the dispersion and ground state correlation of the single particle self-energy on neutron matter superfluidity have been investigated in the framework of the Extended Brueckner-Hartree-Fock and the generalized BCS approaches. A sizable reduction of the energy gap is found due to the energy dependence of the self-energy. And the inclusion of the ground state correlations in the self-energy suppresses further the neutron matter superfluidity

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

CERN Document Server

Zhang Qi; Wang Qun; Xie Zhong Shen

2003-01-01

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

Use of one delayed-neutron precursor group in transient analysis

International Nuclear Information System (INIS)

Diamond, D.J.

1983-01-01

In most reactor dynamics calculations six groups of delayed-neutron precursors are usually accounted for. However, under certain circumstances it may be advantageous to simplify the calculation and utilize a single delayed-neutron group. The motivation for going to one precursor group is economy. For LWR transient codes that use point kinetics the equations are solved very rapidly and six precursor groups should always be used. However, codes with spatially dependent neutron kinetics are very long running and the use of one precursor group may save computer costs and not impair the accuracy of the results significantly. Furthermore, in some codes, the elimation of five presursor groups makes additional memory available which may be used to give a net increase in the accuracy of the calculations, e.g., by allowing for an increase in mesh density. In order to use one delayed neutron precursor group it is necessary to derive a single decay constant, 6 lambda-, which, along with the total (or one group) delayed neutron fraction β = Σ/sub i = 1/β/sub i/, will adequately describe the transeint precursor behavior. The present summary explains how a recommendation for lambda- was derived

Neutron energy response measurement of scintillation detectors

International Nuclear Information System (INIS)

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

2004-01-01

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

Response functions of the Andersson-Braun and extended range rem counters for neutron energies from thermal to 10 GeV

CERN Document Server

Mares, V; Schraube, H

2002-01-01

This work is devoted to the calculation of responses as functions of neutron energy for a paired set of Andersson-Braun rem counters, which is commercially available. Different Monte Carlo codes such as MCNP, LAHET, HADRON and MCNPX were applied in the calculations. The study extended to frontal, lateral and isotropic neutron incidence. For an estimation of the contribution of charged high-energy particles to the reading, the responses to protons and pions were also determined. The results obtained give good bases for the practical use of the new instrument in high-energy neutron fields.

Applied Energy Program

Science.gov (United States)

Science Programs Applied Energy Programs Civilian Nuclear Energy Programs Laboratory Directed Research » Applied Energy Program Applied Energy Program Los Alamos is using its world-class scientific capabilities to enhance national energy security by developing energy sources with limited environmental impact

Symmetry Parameter Constraints from a Lower Bound on Neutron-matter Energy

Energy Technology Data Exchange (ETDEWEB)

Tews, Ingo [Institute for Nuclear Theory, University of Washington, Seattle, WA 98195-1550 (United States); Lattimer, James M. [Department of Physics and Astronomy, Stony Brook University, Stony Brook, NY 11794-3800 (United States); Ohnishi, Akira [Yukawa Institute for Theoretical Physics, Kyoto University, Kyoto 606-8502 (Japan); Kolomeitsev, Evgeni E., E-mail: itews@uw.edu, E-mail: james.lattimer@stonybrook.edu, E-mail: ohnishi@yukawa.kyoto-u.ac.jp, E-mail: e.kolomeitsev@gsi.de [Faculty of Natural Sciences, Matej Bel University, Tajovskeho 40, SK-97401 Banska Bystrica (Slovakia)

2017-10-20

We propose the existence of a lower bound on the energy of pure neutron matter (PNM) on the basis of unitary-gas considerations. We discuss its justification from experimental studies of cold atoms as well as from theoretical studies of neutron matter. We demonstrate that this bound results in limits to the density-dependent symmetry energy, which is the difference between the energies of symmetric nuclear matter and PNM. In particular, this bound leads to a lower limit to the volume symmetry energy parameter S {sub 0}. In addition, for assumed values of S {sub 0} above this minimum, this bound implies both upper and lower limits to the symmetry energy slope parameter L , which describes the lowest-order density dependence of the symmetry energy. A lower bound on neutron-matter incompressibility is also obtained. These bounds are found to be consistent with both recent calculations of the energies of PNM and constraints from nuclear experiments. Our results are significant because several equations of state that are currently used in astrophysical simulations of supernovae and neutron star mergers, as well as in nuclear physics simulations of heavy-ion collisions, have symmetry energy parameters that violate these bounds. Furthermore, below the nuclear saturation density, the bound on neutron-matter energies leads to a lower limit to the density-dependent symmetry energy, which leads to upper limits to the nuclear surface symmetry parameter and the neutron-star crust–core boundary. We also obtain a lower limit to the neutron-skin thicknesses of neutron-rich nuclei. Above the nuclear saturation density, the bound on neutron-matter energies also leads to an upper limit to the symmetry energy, with implications for neutron-star cooling via the direct Urca process.

A proposal to order the neutron data set in neutron spectrometry using the RDANN methodology

Energy Technology Data Exchange (ETDEWEB)

Ortiz R, J.M.; Martinez B, M.R.; Vega C, H.R. [UAZ, Av. Ramon Lopez Velarde No. 801, 98000 Zacatecas (Mexico)

2006-07-01

A new proposal to order a neutron data set in the design process of artificial neural networks in the neutron spectrometry field is presented for first time. The robust design of artificial neural networks methodology was applied to 187 neutron spectra data set compiled by the International Atomic Energy Agency. Four cases of grouping the neutron spectra were considered and around 1000 different neural networks were designed, trained and tested with different net topologies each one. After carrying out the systematic methodology for all the cases, it was determined that the best neural network topology that produced the best reconstructed neutron spectra was case with 187 neutron spectra data set, determining that the best neural network topology is: 7 entrance neurons, 14 neurons in a hidden layer and 31 neurons in the exit layer, with a value of 0.1 in the learning rate and 0.1 in the moment. (Author)

A proposal to order the neutron data set in neutron spectrometry using the RDANN methodology

International Nuclear Information System (INIS)

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

2006-01-01

A new proposal to order a neutron data set in the design process of artificial neural networks in the neutron spectrometry field is presented for first time. The robust design of artificial neural networks methodology was applied to 187 neutron spectra data set compiled by the International Atomic Energy Agency. Four cases of grouping the neutron spectra were considered and around 1000 different neural networks were designed, trained and tested with different net topologies each one. After carrying out the systematic methodology for all the cases, it was determined that the best neural network topology that produced the best reconstructed neutron spectra was case with 187 neutron spectra data set, determining that the best neural network topology is: 7 entrance neurons, 14 neurons in a hidden layer and 31 neurons in the exit layer, with a value of 0.1 in the learning rate and 0.1 in the moment. (Author)

Kaon condensates, nuclear symmetry energy and cooling of neutron stars

Energy Technology Data Exchange (ETDEWEB)

Kubis, S. E-mail: kubis@alf.ifj.edu.pl; Kutschera, M

2003-06-02

The cooling of neutron stars by URCA processes in the kaon-condensed neutron star matter for various forms of nuclear symmetry energy is investigated. The kaon-nucleon interactions are described by a chiral Lagrangian. Nuclear matter energy is parametrized in terms of the isoscalar contribution and the nuclear symmetry energy in the isovector sector. High density behaviour of nuclear symmetry energy plays an essential role in determining the composition of the kaon-condensed neutron star matter which in turn affects the cooling properties. We find that the symmetry energy which decreases at higher densities makes the kaon-condensed neutron star matter fully protonized. This effect inhibits strongly direct URCA processes resulting in slower cooling of neutron stars as only kaon-induced URCA cycles are present. In contrast, for increasing symmetry energy direct URCA processes are allowed in the almost whole density range where the kaon condensation exists.

Kaon condensates, nuclear symmetry energy and cooling of neutron stars

International Nuclear Information System (INIS)

Kubis, S.; Kutschera, M.

2003-01-01

The cooling of neutron stars by URCA processes in the kaon-condensed neutron star matter for various forms of nuclear symmetry energy is investigated. The kaon-nucleon interactions are described by a chiral Lagrangian. Nuclear matter energy is parametrized in terms of the isoscalar contribution and the nuclear symmetry energy in the isovector sector. High density behaviour of nuclear symmetry energy plays an essential role in determining the composition of the kaon-condensed neutron star matter which in turn affects the cooling properties. We find that the symmetry energy which decreases at higher densities makes the kaon-condensed neutron star matter fully protonized. This effect inhibits strongly direct URCA processes resulting in slower cooling of neutron stars as only kaon-induced URCA cycles are present. In contrast, for increasing symmetry energy direct URCA processes are allowed in the almost whole density range where the kaon condensation exists

NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

International Nuclear Information System (INIS)

OH, S.Y.; CHANG, J.; MUGHABGHAB, S.

2000-01-01

Neutron cross section evaluations of the fission-product isotopes, 95 Mo, 99 Tc, 101 Ru, 103 Rh, 105 Pd, 109 Ag, 131 Xe, 133 Cs, 141 Pr, 141 Nd, 147 Sm, 149 Sm, 150 Sm, 151 Sm, 152 Sm, 153 Eu, 155 Gd, and 157 Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of 155 Gd and 157 Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations

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

Scoping studies - photon and low energy neutron interrogation

Energy Technology Data Exchange (ETDEWEB)

Becker, G.; Harker, Y.; Jones, J. [LMITCo, Idaho Falls, ID (United States); Harmon, F. [Idaho State Univ., Pocatello, ID (United States)

1997-11-01

High energy photon interrogation of waste containers, with the aim of producing photo nuclear reactions, in specific materials, holds the potential of good penetration and rapid analysis. Compact high energy ({le} 10 MeV) photon sources in the form of electron linacs producing bremstrahlung radiation are readily available. Work with the Varitron variable energy accelerator at ISU will be described. Advantages and limitations of the technique will be discussed. Using positive ion induced neutron producing reactions, it is possible to generate neutrons in a specific energy range. By this means, variable penetration and specific reactions can be excited in the assayed material. Examples using the {sup 3}H(p,n) and {sup 7}Li(p,n) reactions as neutron sources will be discussed. 4 refs., 7 figs.

Intermediate-energy neutron beam for NCT at MURR

International Nuclear Information System (INIS)

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

1986-01-01

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

Neutron production enhancements for the Intense Pulsed Neutron Source.

Energy Technology Data Exchange (ETDEWEB)

Iverson, E. B.

1999-01-04

The Intense Pulsed Neutron Source (IPNS) was the first high energy spallation neutron source in the US dedicated to materials research. It has operated for sixteen years, and in that time has had a very prolific record concerning the development of new target and moderator systems for pulsed spallation sources. IPNS supports a very productive user program on its thirteen instruments, which are oversubscribed by more than two times, meanwhile having an excellent overall reliability of 95%. Although the proton beam power is relatively low at 7 kW, the target and moderator systems are very efficient. The typical beam power which gives an equivalent flux for long-wavelength neutrons is about 60 kW, due to the use of a uranium target and liquid and solid methane moderators, precluded at some sources due to a higher accelerator power. The development of new target and moderator systems is by no means stagnant at IPNS. They are presently considering numerous enhancements to the target and moderators that offer prospects for increasing the useful neutron production by substantial factors. Many of these enhancements could be combined, although their combined benefit has not yet been well established. Meanwhile, IPNS is embarking on a coherent program of study concerning these improvements and their possible combination and implementation. Moreover, any improvements accomplished at IPNS would immediately increase the performance of IPNS instruments.

Neutron production enhancements for the Intense Pulsed Neutron Source

International Nuclear Information System (INIS)

Iverson, E. B.

1999-01-01

The Intense Pulsed Neutron Source (IPNS) was the first high energy spallation neutron source in the US dedicated to materials research. It has operated for sixteen years, and in that time has had a very prolific record concerning the development of new target and moderator systems for pulsed spallation sources. IPNS supports a very productive user program on its thirteen instruments, which are oversubscribed by more than two times, meanwhile having an excellent overall reliability of 95%. Although the proton beam power is relatively low at 7 kW, the target and moderator systems are very efficient. The typical beam power which gives an equivalent flux for long-wavelength neutrons is about 60 kW, due to the use of a uranium target and liquid and solid methane moderators, precluded at some sources due to a higher accelerator power. The development of new target and moderator systems is by no means stagnant at IPNS. They are presently considering numerous enhancements to the target and moderators that offer prospects for increasing the useful neutron production by substantial factors. Many of these enhancements could be combined, although their combined benefit has not yet been well established. Meanwhile, IPNS is embarking on a coherent program of study concerning these improvements and their possible combination and implementation. Moreover, any improvements accomplished at IPNS would immediately increase the performance of IPNS instruments

RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

International Nuclear Information System (INIS)

Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

1977-07-01

The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

Extension of the AUS reactor neutronics system for application to fusion blanket neutronics

International Nuclear Information System (INIS)

Robinson, G.S.

1984-03-01

The AUS modular code scheme for reactor neutronics computations has been extended to apply to fusion blanket neutronics. A new group cross-section library with 200 neutron groups, 37 photon groups and kerma factor data has been generated from ENDF/B-IV. The library includes neutron resonance subgroup parameters and temperature-dependent data for thermal neutron scattering matrices. The validity of the overall calculation system for fusion applications has been checked by comparison with a number of published conceptual design studies

Recent research on nuclear reaction using high-energy proton and neutron

Energy Technology Data Exchange (ETDEWEB)

Shibata, Tokushi [Tokyo Univ., Tanashi (Japan). Inst. for Nuclear Study

1997-11-01

The presently available high-energy neutron beam facilities are introduced. Then some interesting research on nuclear reaction using high-energy protons are reported such as the intermediate mass fragments emission and neutron spectrum measurements on various targets. As the important research using high-energy neutron, the (p,n) reactions on Mn, Fe, and Ni, the elastic scattering of neutrons, and the shielding experiments are discussed. (author)

An efficient methodology of two groups spatial calculation for neutronic state and sensisivity coefficients in fast reactors

International Nuclear Information System (INIS)

Jachic, J.

1985-01-01

It is presented the ONEDM neutronic simulator for RZ spatial calculation, two energy groups, aiming at researching and optimization of a low power fast reactor design. The simulator's methodology is based in RZ calculation from radial and axial calculation iteractively coupled and in macroscopic cross sections corrected by power density and asymmetry of the spectrum in the feedback process with phase library for reference neutronic state. The transversal area which are determined by energy groups and material region in the iteration are introduced in the spatial calculation. The simulator efficiency is tested and compared with the CITATION and 2DB codes. The cross sections are generated by 1DX code. (M.C.K.) [pt

Development of a medium energy polarized neutron facility

International Nuclear Information System (INIS)

Burzynski, S.; Gysin, C.; Henneck, R.; Jourdan, J.; Kohler, D.; Pickar, M.A.; Plattner, G.R.; Sick, I.; Berdoz, A.; Foroughi, F.; Nussbaum, Ch.; Stammbach, Th.

1984-01-01

By the end of 1983 the major construction work for the new polarized neutron source was completed. The source will provide an essentially monoenergetic beam of both polarized and unpolarized neutrons in the energy range from 20 MeV to 70 MeV. Intensities are expected to be approx. 2 x 10 5 neutrons/s.cm 2 per μA of incident proton beam. The polarization is expected to be approx. 0.2 and can be chosen to be either longitudinal or transverse. Protons from the Philips injector cyclotron are focussed onto a liquid deuterium target and produce neutrons via the 2 H(p,n)2p reaction at 0 0 . This process provides essentially monoenergetic neutrons of almost the same energy as the incoming protons. The zero production angle implies that the neutron polarization comes from the polarization of the proton beam only. This allows an easy and fast change of the neutron spin direction by selecting proton spin states in the polarized ion source (atomic beam type). (Auth.)

Portable instrument for measuring neutron energy spectra and neutron dose in a mixed n-γ field

International Nuclear Information System (INIS)

Daniels, C. J.; Silberberg, J. L.

1980-01-01

A portable high-speed neutron spectrometer consists of an organic scintillator, a true zero-crossing pulse shape discriminator, a 1 MHZ conversion-rate multichannel analyzer, an 8-bit microcomputer, and appropriate displays. The device can be used to measure neutron energy spectra and kerma rate in intense n- gamma radiation fields in which the neutron energy is from 5 to 15 MEV

The Lattice Boltzmann Method applied to neutron transport

International Nuclear Information System (INIS)

Erasmus, B.; Van Heerden, F. A.

2013-01-01

In this paper the applicability of the Lattice Boltzmann Method to neutron transport is investigated. One of the main features of the Lattice Boltzmann method is the simultaneous discretization of the phase space of the problem, whereby particles are restricted to move on a lattice. An iterative solution of the operator form of the neutron transport equation is presented here, with the first collision source as the starting point of the iteration scheme. A full description of the discretization scheme is given, along with the quadrature set used for the angular discretization. An angular refinement scheme is introduced to increase the angular coverage of the problem phase space and to mitigate lattice ray effects. The method is applied to a model problem to investigate its applicability to neutron transport and the results are compared to a reference solution calculated, using MCNP. (authors)

Matrix-type multiple reciprocity boundary element method for solving three-dimensional two-group neutron diffusion equations

International Nuclear Information System (INIS)

Itagaki, Masafumi; Sahashi, Naoki.

1997-01-01

The multiple reciprocity boundary element method has been applied to three-dimensional two-group neutron diffusion problems. A matrix-type boundary integral equation has been derived to solve the first and the second group neutron diffusion equations simultaneously. The matrix-type fundamental solutions used here satisfy the equation which has a point source term and is adjoint to the neutron diffusion equations. A multiple reciprocity method has been employed to transform the matrix-type domain integral related to the fission source into an equivalent boundary one. The higher order fundamental solutions required for this formulation are composed of a series of two types of analytic functions. The eigenvalue itself is also calculated using only boundary integrals. Three-dimensional test calculations indicate that the present method provides stable and accurate solutions for criticality problems. (author)

The neutronic design and performance of the Indiana University Cyclotron Facility (IUCF) Low Energy Neutron Source (LENS)

Science.gov (United States)

Lavelle, Christopher M.

Neutron scattering research is performed primarily at large-scale facilities. However, history has shown that smaller scale neutron scattering facilities can play a useful role in education and innovation while performing valuable materials research. This dissertation details the design and experimental validation of the LENS TMR as an example for a small scale accelerator driven neutron source. LENS achieves competitive long wavelength neutron intensities by employing a novel long pulse mode of operation, where the neutron production target is irradiated on a time scale comparable to the emission time of neutrons from the system. Monte Carlo methods have been employed to develop a design for optimal production of long wavelength neutrons from the 9Be(p,n) reaction at proton energies ranging from 7 to 13 MeV proton energy. The neutron spectrum was experimentally measured using time of flight, where it is found that the impact of the long pulse mode on energy resolution can be eliminated at sub-eV neutron energies if the emission time distribution of neutron from the system is known. The emission time distribution from the TMR system is measured using a time focussed crystal analyzer. Emission time of the fundamental cold neutron mode is found to be consistent with Monte Carlo results. The measured thermal neutron spectrum from the water reflector is found to be in agreement with Monte Carlo predictions if the scattering kernels employed are well established. It was found that the scattering kernels currently employed for cryogenic methane are inadequate for accurate prediction of the cold neutron intensity from the system. The TMR and neutronic modeling have been well characterized and the source design is flexible, such that it is possible for LENS to serve as an effective test bed for future work in neutronic development. Suggestions for improvements to the design that would allow increased neutron flux into the instruments are provided.

The Group Neutron Data Library (GNDL)

International Nuclear Information System (INIS)

Voronkov, A.V.; Zhuravlev, V.I.; Natrusova, E.G.

1987-01-01

The paper describes the structure, organization and basic data representation formats of the GNDL, which was developed at the M.V. Keldysh Institute of Applied Mathematics of the USSR Academy of Sciences for the purpose of neutron data storage and retrieval. A simple method for linking up applications programs with the library is proposed. (author)

Feasibility of the Precise Energy Calibration for Fast Neutron Spectrometers

Science.gov (United States)

Gaganov, V. V.; Usenko, P. L.; Kryzhanovskaja, M. A.

2017-12-01

Computational studies aimed at improving the accuracy of measurements performed using neutron generators with a tritium target were performed. A measurement design yielding an extremely narrow peak in the energy spectrum of DT neutrons was found. The presence of such a peak establishes the conditions for precise energy calibration of fast-neutron spectrometers.

Continuous energy Monte Carlo method based homogenization multi-group constants calculation

International Nuclear Information System (INIS)

Li Mancang; Wang Kan; Yao Dong

2012-01-01

The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P 1 cross sections were used to calculate the diffusion coefficients for diffusion reactor simulator codes. B N theory is applied to take the leakage effect into account when the infinite lattice of identical symmetric motives is assumed. The MCMC code was developed and the code was applied in four assembly configurations to assess the accuracy and the applicability. At core-level, A PWR prototype core is examined. The results show that the Monte Carlo based multi-group constants behave well in average. The method could be applied to complicated configuration nuclear reactor core to gain higher accuracy. (authors)

Neutron radiography working group test programme

International Nuclear Information System (INIS)

Domanus, J.C.

1989-03-01

Scope and results of the Euratom Neutron Radiography Working Group Test Program are described. Seven NR centers from six European Community countries have performed this investigation using eleven NR facilities. Four test items were neutron radiographed using 30 different film/converter combinations. From film density measurements neutron beam components were determined. Radiographic sensitivity was assessed from visual examinations of the radiographs. About 25,000 dimensional measurements were made and were used for the assessment of accuracies of dimensional measurements from neutron radiographs. The report gives a description of the test items used for the Test Program, the film density and dimensional measurements, and concentrates on the assessment of the measuring results. The usefulness of the beam purity and sensitivity indicators was assessed with the conclusion that they are not suitable for neutron radiography of nuclear reactor fuel. Ample information is included in the report about measuring accuracies which can be reached in dimensional measurements of fuel pins. After a general comparison of measuring accuracies is discussed. Results from different NR facilities are treated separately as are the different kinds of dimensions of the fuel pins. Finally human and instrument factors are discussed. After presenting final conclusions (which take into account the above-mentioned factors) results of other investigations about dimensional measurements are shortly reviewed

Determining space-energy distribution of thermal neutrons in heterogeneous cylindrically symmetric reactor cell, Master Thesis; Odredjivanje prostorno energetske raspodele fluksa termalnih neutrona u heterogenoj cilindricno simetricnoj reaktorskoj celiji, Magistarski rad

Energy Technology Data Exchange (ETDEWEB)

Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1966-06-15

A combination of multigroup method and P{sub 3} approximation of spherical harmonics method was chosen for calculating space-energy distribution of thermal neutron flux in elementary reactor cell. Application of these methods reduced solution of complicated transport equation to the problem of solving an inhomogeneous system of six ordinary firs-order differential equations. A procedure is proposed which avoids numerical solution and enables analytical solution when applying certain approximations. Based on this approach, computer codes were written for ZUSE-Z-23 computer: SIGMA code for calculating group constants for a given material; MULTI code which uses results of SIGMA code as input and calculates spatial ana energy distribution of thermal neutron flux in a reactor cell. Calculations of thermal neutron spectra for a number of reactor cells were compared to results available from literature. Agreement was satisfactory in all the cases, which proved the correctness of the applied method. Some possibilities for improving the precision and acceleration of the calculation process were found during calculation. (author)

Monitor units are not predictive of neutron dose for high-energy IMRT

Directory of Open Access Journals (Sweden)

Hälg Roger A

2012-08-01

Full Text Available Abstract Background Due to the substantial increase in beam-on time of high energy intensity-modulated radiotherapy (>10 MV techniques to deliver the same target dose compared to conventional treatment techniques, an increased dose of scatter radiation, including neutrons, is delivered to the patient. As a consequence, an increase in second malignancies may be expected in the future with the application of intensity-modulated radiotherapy. It is commonly assumed that the neutron dose equivalent scales with the number of monitor units. Methods Measurements of neutron dose equivalent were performed for an open and an intensity-modulated field at four positions: inside and outside of the treatment field at 0.2 cm and 15 cm depth, respectively. Results It was shown that the neutron dose equivalent, which a patient receives during an intensity-modulated radiotherapy treatment, does not scale with the ratio of applied monitor units relative to an open field irradiation. Outside the treatment volume at larger depth 35% less neutron dose equivalent is delivered than expected. Conclusions The predicted increase of second cancer induction rates from intensity-modulated treatment techniques can be overestimated when the neutron dose is simply scaled with monitor units.

Measurements of angular and energy distributions of gamma-rays resulting from neutron interactions in shielding barriers

International Nuclear Information System (INIS)

Makarious, A.S.; Maayouf, R.M.A.; Megahid, R.

1978-01-01

Measurements of both angular and energy distributions of secondary gamma resulting from interactions of neutrons emerging from one of the ET-RR-1 reactor beam holes, in barriers from iron, lead and water are reported. The measurements were carried out, both with a bare neutron beam and with the beam being transmitted through a B4C. Filter, using a stilbene crystal gamma spectrometer. The spectrometer applies discrimination between neutrons and gammas according to the difference in decay times of the scintillations produced by them in stilbene. The described angular distributions resulted from measurements made at different angles of neutron incidence and with three different thicknesses of each sample

Characterization of Monoenergetic Low Energy Neutron Fields with the {mu}TPC Detector

Energy Technology Data Exchange (ETDEWEB)

Golabek, C.; Lebreton, L.; Petit, M. [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Billard, J.; Grignon, C.; Bosson, G.; Bourrion, O.; Guillaudin, O.; Mayet, F.; Richer, J.-P.; Santos, D. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph (France)

2011-12-13

The AMANDE facility produces monoenergetic neutron fields from 2 keV to 20 MeV for metrological purposes. To be considered as a reference facility, fluence and energy distributions of neutron fields have to be determined by primary measurement standards. For this purpose, a micro Time Projection Chamber is being developed to be dedicated to measure neutron fields with energy ranging from 2 keV up to 1 MeV. We present simulations showing that such a detector, which allows the measurement of the ionization energy and the 3D reconstruction of the recoil nucleus, provides the determination of neutron energy and fluence of such low energy neutron fields.

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

Optimization of a neutron detector design using adjoint transport simulation

International Nuclear Information System (INIS)

Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

2012-01-01

A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

Analysis of incident-energy dependence of delayed neutron yields in actinides

Energy Technology Data Exchange (ETDEWEB)

Nasir, Mohamad Nasrun bin Mohd, E-mail: monasr211@gmail.com; Metorima, Kouhei, E-mail: kohei.m2420@hotmail.co.jp; Ohsawa, Takaaki, E-mail: ohsawa@mvg.biglobe.ne.jp; Hashimoto, Kengo, E-mail: kengoh@pp.iij4u.or.jp [Graduate School of Science and Engineering, Kindai University, Kowakae, Higashi-Osaka, 577-8502 (Japan)

2015-04-29

The changes of delayed neutron yields (ν{sub d}) of Actinides have been analyzed for incident energy up to 20MeV using realized data of precursor after prompt neutron emission, from semi-empirical model, and delayed neutron emission probability data (P{sub n}) to carry out a summation method. The evaluated nuclear data of the delayed neutron yields of actinide nuclides are still uncertain at the present and the cause of the energy dependence has not been fully understood. In this study, the fission yields of precursor were calculated considering the change of the fission fragment mass yield based on the superposition of fives Gaussian distribution; and the change of the prompt neutrons number associated with the incident energy dependence. Thus, the incident energy dependent behavior of delayed neutron was analyzed.The total number of delayed neutron is expressed as ν{sub d}=∑Y{sub i} • P{sub ni} in the summation method, where Y{sub i} is the mass yields of precursor i and P{sub ni} is the delayed neutron emission probability of precursor i. The value of Y{sub i} is derived from calculation of post neutron emission mass distribution using 5 Gaussian equations with the consideration of large distribution of the fission fragments. The prompt neutron emission ν{sub p} increases at higher incident-energy but there are two different models; one model says that the fission fragment mass dependence that prompt neutron emission increases uniformly regardless of the fission fragments mass; and the other says that the major increases occur at heavy fission fragments area. In this study, the changes of delayed neutron yields by the two models have been investigated.

Continuous energy adjoint Monte Carlo for coupled neutron-photon transport

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.

2001-07-01

Although the theory for adjoint Monte Carlo calculations with continuous energy treatment for neutrons as well as for photons is known, coupled neutron-photon transport problems present fundamental difficulties because of the discrete energies of the photons produced by neutron reactions. This problem was solved by forcing the energy of the adjoint photon to the required discrete value by an adjoint Compton scattering reaction or an adjoint pair production reaction. A mathematical derivation shows the exact procedures to follow for the generation of an adjoint neutron and its statistical weight. A numerical example demonstrates that correct detector responses are obtained compared to a standard forward Monte Carlo calculation. (orig.)

NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

Energy Technology Data Exchange (ETDEWEB)

OH,S.Y.; CHANG,J.; MUGHABGHAB,S.

2000-05-11

Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.

High-Energy Neutron Backgrounds for Underground Dark Matter Experiments

Energy Technology Data Exchange (ETDEWEB)

Chen, Yu [Syracuse Univ., NY (United States)

2016-01-01

Direct dark matter detection experiments usually have excellent capability to distinguish nuclear recoils, expected interactions with Weakly Interacting Massive Particle (WIMP) dark matter, and electronic recoils, so that they can efficiently reject background events such as gamma-rays and charged particles. However, both WIMPs and neutrons can induce nuclear recoils. Neutrons are then the most crucial background for direct dark matter detection. It is important to understand and account for all sources of neutron backgrounds when claiming a discovery of dark matter detection or reporting limits on the WIMP-nucleon cross section. One type of neutron background that is not well understood is the cosmogenic neutrons from muons interacting with the underground cavern rock and materials surrounding a dark matter detector. The Neutron Multiplicity Meter (NMM) is a water Cherenkov detector capable of measuring the cosmogenic neutron flux at the Soudan Underground Laboratory, which has an overburden of 2090 meters water equivalent. The NMM consists of two 2.2-tonne gadolinium-doped water tanks situated atop a 20-tonne lead target. It detects a high-energy (>~ 50 MeV) neutron via moderation and capture of the multiple secondary neutrons released when the former interacts in the lead target. The multiplicity of secondary neutrons for the high-energy neutron provides a benchmark for comparison to the current Monte Carlo predictions. Combining with the Monte Carlo simulation, the muon-induced high-energy neutron flux above 50 MeV is measured to be (1.3 ± 0.2) ~ 10-9 cm-2s-1, in reasonable agreement with the model prediction. The measured multiplicity spectrum agrees well with that of Monte Carlo simulation for multiplicity below 10, but shows an excess of approximately a factor of three over Monte Carlo prediction for multiplicities ~ 10 - 20. In an effort to reduce neutron backgrounds for the dark matter experiment SuperCDMS SNO- LAB, an active neutron veto was developed

On the exact solution for the multi-group kinetic neutron diffusion equation in a rectangle

International Nuclear Information System (INIS)

Petersen, C.Z.; Vilhena, M.T.M.B. de; Bodmann, B.E.J.

2011-01-01

In this work we consider the two-group bi-dimensional kinetic neutron diffusion equation. The solution procedure formalism is general with respect to the number of energy groups, neutron precursor families and regions with different chemical compositions. The fast and thermal flux and the delayed neutron precursor yields are expanded in a truncated double series in terms of eigenfunctions that, upon insertion into the kinetic equation and upon taking moments, results in a first order linear differential matrix equation with source terms. We split the matrix appearing in the transformed problem into a sum of a diagonal matrix plus the matrix containing the remaining terms and recast the transformed problem into a form that can be solved in the spirit of Adomian's recursive decomposition formalism. Convergence of the solution is guaranteed by the Cardinal Interpolation Theorem. We give numerical simulations and comparisons with available results in the literature. (author)

Symmetry energy and surface properties of neutron-rich exotic nuclei

Energy Technology Data Exchange (ETDEWEB)

Gaidarov, M. K.; Antonov, A. N. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Sofia 1784 (Bulgaria); Sarriguren, P. [Instituto de Estructura de la Materia, IEM-CSIC, Serrano 123, E-28006 Madrid (Spain); Moya de Guerra, E. [Departamento de Fisica Atomica, Molecular y Nuclear, Facultad de Ciencias Fisicas, Universidad Complutense de Madrid, E-28040 Madrid (Spain)

2014-07-23

The symmetry energy, the neutron pressure and the asymmetric compressibility of spherical Ni, Sn, and Pb and deformed Kr and Sm neutron-rich even-even nuclei are calculated within the coherent density fluctuation model using the symmetry energy as a function of density within the Brueckner energy-density functional. The correlation between the thickness of the neutron skin and the characteristics related with the density dependence of the nuclear symmetry energy is investigated for isotopic chains of these nuclei in the framework of the deformed self-consistent mean-field Skyrme HF+BCS method. The mass dependence of the nuclear symmetry energy and the neutron skin thickness are also studied together with the role of the neutron-proton asymmetry. The studied correlations reveal a smoother behavior in the case of spherical nuclei than for deformed ones. We also notice that the neutron skin thickness obtained for {sup 208}Pb with SLy4 force is found to be in a good agreement with the recent data. In addition to the interest that this study may have by itself, we give some numerical arguments in proof of the existence of peculiarities of the studied quantities in Ni and Sn isotopic chains that are not present in the Pb chain.

Experimental evaluation of the primary damage process: neutron energy effects

Energy Technology Data Exchange (ETDEWEB)

Goland, A.N.

1979-01-01

Experimental evaluation of the neutron energy dependnece of the primary damage stage depends upon a number of theoretical concepts. This state can only be observed after low- or perhaps ambient-temperature, low-fluence irradiations. The primary recoil energy spectrum, which determines the character of the displacement cascades, can be calculated if dosimetry has provided an accurate neutron spectrum. A review of experimental results relating neutron-energy effects shows that damage energy or damage energy cross section has often been a reliable correlation parameter for primary damage state experiments. However, the forthcoming emphasis on higher irradiation temperatures, more complex alloys and microstructural evolution has fostered a search for additional meaningful correlation parameters.

Oncogenic transformation in C3H10T1/2 cells by low-energy neutrons.

Science.gov (United States)

Miller, R C; Marino, S A; Napoli, J; Shah, H; Hall, E J; Geard, C R; Brenner, D J

2000-03-01

Occupational exposure to neutrons typically includes significant doses of low-energy neutrons, with energies below 100 keV. In addition, the normal-tissue dose from boron neutron capture therapy will largely be from low-energy neutrons. Microdosimetric theory predicts decreasing biological effectiveness for neutrons with energies below about 350 keV compared with that for higher-energy neutrons; based on such considerations, and limited biological data, the current radiation weighting factor (quality factor) for neutrons with energies from 10 keV to 100 keV is less than that for higher-energy neutrons. By contrast, some reports have suggested that the biological effectiveness of low-energy neutrons is similar to that of fast neutrons. The purpose of the current work is to assess the relative biological effectiveness of low-energy neutrons for an endpoint of relevance to carcinogenesis: in vitro oncogenic transformation. Oncogenic transformation induction frequencies were determined for C3H10T1/2 cells exposed to two low-energy neutron beams, respectively, with dose-averaged energies of 40 and 70 keV, and the results were compared with those for higher-energy neutrons and X-rays. These results for oncogenic transformation provide evidence for a significant decrease in biological effectiveness for 40 keV neutrons compared with 350 keV neutrons. The 70 keV neutrons were intermediate in effectiveness between the 70 and 350 keV beams. A decrease in biological effectiveness for low-energy neutrons is in agreement with most (but not all) earlier biological studies, as well as microdosimetric considerations. The results for oncogenic transformation were consistent with the currently recommended decreased values for low-energy neutron radiation weighting factors compared with fast neutrons.

D-D neutron energy-spectra measurements in Alcator C

International Nuclear Information System (INIS)

Pappas, D.S.; Wysocki, F.J.; Furnstahl, R.J.

1982-08-01

Measurements of energy spectra of neutrons produced during high density (anti n/sub e/ > 2 x 10 14 cm -3 ) deuterium discharges have been performed using a proton-recoil (NE 213) spectrometer. A two foot section of light pipe (coupling the scintillator and photomultiplier) was used to extend the scintillator into a diagnostic viewing port to maximize the neutron detection efficiency while not imposing excessive magnetic shielding requirements. A derivative unfolding technique was used to deduce the energy spectra. The results showed a well defined peak at 2.5 MeV which was consistent with earlier neutron flux measurements on Alcator C that indicated the neutrons were of thermonuclear origin

Neutron scattering in applied research: possibilities for medium-flux research reactors

International Nuclear Information System (INIS)

Reijonen, H.

1977-01-01

The possibilities for neutron scattering techniques in applied research are reviewed. The areas covered are magnetism, determination of hydrogen self-diffusion constants and ionic mobility in superionic conductors, liquid crystals, molecular solids, polymers, surface chemistry and catalysts, colloids biology, physical metallurgy and neutron diagnostics. Established applications and new proposals for research projects are presented in various degrees of detail. Small angle scattering techniques, physical metallurgy and biology were identified as having particular potential. Prospective users should be realistically informed of the possibilities of neutron scattering techniques. (author)

High energy resolution measurement of the sup 238 U neutron capture yield from 1 to 100 keV

Energy Technology Data Exchange (ETDEWEB)

Macklin, R.L. (Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering); Perez, R.B. (Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering Oak Ridge National Lab., TN (United States)); De Saussure, G.; Ingle, R.W. (Oak Ridge National Lab., TN (United States))

1991-01-01

The purpose of this work is the precise determination of the {sup 238}U neutron capture yield (i.e. the probability of neutron capture) as a function of neutron energy with the highest available neutron energy resolution. The motivation for this undertaking arises from the central role played by the {sup 238}U neutron capture process in the neutron balance of both thermal reactors and fast breeder reactors. The present measurement was performed using the Oak Ridge Electron Linear Accelerator (ORELA) facility. The pulsed beam of neutrons from the ORELA facility is collimated on a sample of {sup 238}U. The neutron capture rate in the sample is measured, as a function of neutron time-of-flight (TOF) by detecting the {gamma}-rays from the {sup 238}U(n, {gamma}){sup 239}U reaction with a large {gamma}-ray detector surrounding the {sup 238}U sample. At each energy, the capture yield is proportional to the observed capture rate divided by the measured intensity of the neutron beam. The constant of proportionality (the normalization constant) is obtained from the ratio of theoretical to experimentally measured areas under small {sup 238}U resonances where the resonance parameters have been determined from high-resolution {sup 238}U transmission measurements. The cross section for the reaction {sup 238}U(n,{gamma}){sup 239}U can be derived from the measured capture yield if one applies appropriate corrections for multiple scattering and resonance self-shielding. Some 200 {sup 238}U neutron resonances in the energy range from 250 eV to 10 keV have been observed which had not been detected in previous measurements. (author).

Use of Neutron Beams for Materials Research Relevant to the Nuclear Energy Sector

International Nuclear Information System (INIS)

2015-10-01

Nuclear technologies such as fission and fusion reactors, including associated waste storage and disposal, rely on the availability of not only nuclear fuels but also advanced structural materials. In 2010–2013, the IAEA organized and implemented the Coordinated Research Project (CRP) on Development, Characterization and Testing of Materials of Relevance to Nuclear Energy Sector Using Neutron Beams. A total of 19 institutions from 18 Member States (Argentina, Australia, Brazil, China, Czech Republic, France, Germany, Hungary, Indonesia, Italy, Japan, Netherlands, Republic of Korea, Romania, Russian Federation (two institutions), South Africa, Switzerland and United States of America) cooperated with the main objective to address the use of various neutron beam techniques for characterization, testing and qualification of materials and components produced or under development for applications in the nuclear energy sector. This CRP aimed to bring stakeholders and end users of research reactors and accelerator based neutron sources together for the enhanced use of available facilities and development of new infrastructures for applied materials research. Work envisioned under this CRP was related to the optimization and validation of neutron beam techniques, including facility and instrument modifications/optimizations as well as improved data acquisition, processing and analysis systems. Particular emphasis was placed on variable environments during material characterization and testing as required by some applications such as intensive irradiation load, high temperature and high pressure conditions, and the presence of strong magnetic fields. Targeted neutron beam techniques were neutron diffraction, small angle neutron scattering and digital neutron radiography/tomography. This publication is a compilation of the main results and findings of the CRP, and the CD-ROM accompanying this publication contains 19 reports with additional relevant technical details

Monte Carlo electron-transport calculations for clinical beams using energy grouping

Energy Technology Data Exchange (ETDEWEB)

Teng, S P; Anderson, D W; Lindstrom, D G

1986-01-01

A Monte Carlo program has been utilized to study the penetration of broad electron beams into a water phantom. The MORSE-E code, originally developed for neutron and photon transport, was chosen for adaptation to electrons because of its versatility. The electron energy degradation model employed logarithmic spacing of electron energy groups and included effects of elastic scattering, inelastic-moderate-energy-loss-processes and inelastic-large-energy-loss-processes (catastrophic). Energy straggling and angular deflections were modeled from group to group, using the Moeller cross section for energy loss, and Goudsmit-Saunderson theory to describe angular deflections. The resulting energy- and electron-deposition distributions in depth were obtained at 10 and 20 MeV and are compared with ETRAN results and broad beam experimental data from clinical accelerators.

The energy spectrum of delayed neutrons from thermal neutron induced fission of 235U and its analytical approximation

International Nuclear Information System (INIS)

Doroshenko, A.Yu.; Tarasko, M.Z.; Piksaikin, V.M.

2002-01-01

The energy spectrum of the delayed neutrons is the poorest known of all input data required in the calculation of the effective delayed neutron fractions. In addition to delayed neutron spectra based on the aggregate spectrum measurements there are two different approaches for deriving the delayed neutron energy spectra. Both of them are based on the data related to the delayed neutron spectra from individual precursors of delayed neutrons. In present work these two different data sets were compared with the help of an approximation by gamma-function. The choice of this approximation function instead of the Maxwellian or evaporation type of distribution is substantiated. (author)

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)

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

International Nuclear Information System (INIS)

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

2015-10-01

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

Energy spectra of neutrons accompanying the emission fission of 238U

International Nuclear Information System (INIS)

Smirenkin, G.N.; Lovchikova, G.N.; Trufanov, A.M.; Svirin, M.I.; Polyakov, A.V.; Vinogradov, V.A.; Dmitriev, V.D.; Boykov, G.S.

1996-01-01

The spectra of fission neutrons emitted from 238U are measured for the first time by the time-of-flight method at incident-neutron energies of 16.0 and 17.7 MeV. Analysis of the neutron spectra shows that experimental results at incident-neutron energies of 14.7, 16.0, and 17.7 MeV (above the threshold of chance fission) differ significantly from those obtained at a neutron energy of 2.9 MeV (below the threshold of chance fission). Owing to the prefission emission of neutrons, the observed spectra of neutrons from emission fission exhibit a characteristic growth of the neutron yield in both hard and soft sections of the spectrum of secondary neutrons. This growth manifests itself as a step in the first case and as a rise in the second case, where it results in a noticeable excess of neutrons over the statistical-model predictions for E<2 MeV. The first feature in the spectra of neutrons from emission fission can be associated with the nonequilibrium decay of an excited fissile nucleus. On the contrary, the origin of the second feature has yet to be clarified. Additional measurements of angular distributions of secondary neutrons may prove helpful in this respect

Investigations of safety-related parameters applying a new multi-group diffusion code for HTR transients

International Nuclear Information System (INIS)

Kasselmann, S.; Druska, C.; Lauer, A.

2010-01-01

The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)

Transport calculation of medium-energy protons and neutrons by Monte Carlo method

International Nuclear Information System (INIS)

Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.

1978-09-01

A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)

A {mu}TPC detector for the characterization of low energy neutron fields

Energy Technology Data Exchange (ETDEWEB)

Golabek, C., E-mail: cedric.golabek@irsn.fr [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Billard, J. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France); Allaoua, A. [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Bosson, G.; Bourrion, O.; Grignon, C.; Guillaudin, O. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France); Lebreton, L., E-mail: lena.lebreton@irsn.fr [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Mayet, F. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France); Petit, M. [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Richer, J.-P.; Santos, D. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France)

2012-06-21

The AMANDE facility produces monoenergetic neutron fields from 2 keV to 20 MeV for metrological purposes. To be considered as a reference facility, fluence and energy distributions of neutron fields have to be determined by primary measurement standards. For this purpose, a micro Time Projection Chamber is being developed to be dedicated to measure neutron fields with energy ranging from 8 keV up to 1 MeV. In this work we present simulations showing that such a detector, which allows the measurement of the ionization energy and the 3D reconstruction of the recoil nucleus, provides the determination of neutron energy and fluence of these neutron fields.

Simulations of the neutron energy-spectra at the Olympus Gate Environmental Monitoring Station due to historical Bevatron operations

International Nuclear Information System (INIS)

Donahue, R.J.; Thomas, R.H.; Zeman, G.H.

2001-01-01

Offsite neutron fluences resulting from Bevatron operations reached a maximum in 1959, prior to the addition of a permanent concrete roof shield, which was constructed in 1962. From the first operation of the Bevatron measurements of neutron fluence were made at locations around the perimeter of the Lawrence Berkeley National Laboratory (LBNL) campus. Since the late 1950's measurements made at several locations, and particularly at the site of what is now called the Olympus Gate Environmental Monitoring Station, have been routinely reported and published. Early measurements were used to establish the shape of the neutron-energy spectrum from which an energy-averaged fluence-to-dose equivalent conversion coefficient could be derived. This conversion coefficient was then applied to a measured total neutron fluence to obtain the appropriate dose equivalent quantity required by regulation. Recent work by Thomas et al. (2000) have compared the early conversion coefficients used in the sixties with those accepted today and suggest suggested that ''the dose equivalents reported in the late fifties and early sixties were conservative by factors between two and four. In any current review of the historical data, therefore it would be prudent to reduce the reported dose equivalents by at least a factor of two.'' However, that analysis was based on the ''state of the art'' neutron energy-spectra of the '60s. This paper provides a detailed knowledge of the neutron energy spectrum at the site boundary paper thus removing any uncertainty in the analysis of Thomas et al., which might be caused by the use of the early neutron energy-spectra. Detailed Monte Carlo analyses of the interactions of 6.2 GeV protons in thick, medium-A targets are described. In the computer simulations, neutrons produced were allowed to scatter in the atmosphere. Detailed neutron energy spectra were calculated at a distance and elevation corresponding to the location of the Olympus Gate EMS. Both older

Performance of Large Neutron Detectors Containing Lithium-Gadolinium-Borate Scintillator

Energy Technology Data Exchange (ETDEWEB)

Slaughter, David M.; Stuart, Cory R.; Klaass, R. Fred; Merrill, David B. [MSI/Photogenics Division, Orem, Utah (United States)

2015-07-01

This paper describes the development and testing of a neutron counter, spectrometer, and dosimeter that is compact, efficient, and accurate. A self-contained neutron detection instrument has wide applications in health physics, scientific research, and programs to detect, monitor, and control strategic nuclear materials (SNM). The 1.3 liter detector head for this instrument is a composite detector with an organic scintillator containing uniformly distributed {sup 6}Li{sub 6}{sup nat}Gd{sup 10}B{sub 3}O{sub 9}:Ce (LGB:Ce) microcrystals. The plastic scintillator acts to slow impinging neutrons and emits light proportional to the energy lost by the neutrons as they moderate in the detector body. Moderating neutrons that have slowed sufficiently capture in one of the Lithium-6, Boron-10, or Gadolinium-157 atoms in the LGB:Ce scintillator, which then releases the capture energy in a characteristic cerium emission pulse. The measured captured pulses indicate the presence of neutrons. When a scintillating fluor is present in the plastic, the light pulse resulting from the neutron moderating in the plastic is paired with the LGB:Ce capture pulse to identify the energy of the neutron. About 2% of the impinging neutrons lose all of their energy in a single collision with the detector. There is a linear relationship between the pulse areas of this group of neutrons and energy. The other 98% of neutrons have a wide range of collision histories within the detector body. When these neutrons are 'binned' into energy groups, each group contains a distribution of pulse areas. This data was used to assist in the unfolding of the neutron spectra. The unfolded spectra were then validated with known spectra, at both neutron emitting isotopes and fission/accelerator facilities. Having validated spectra, the dose equivalent and dose rate are determined by applying standard, regulatory damage coefficients to the measured neutron counts for each energy bin of the spectra. Testing

Nuclear symmetry energy and the neutron skin in neutron-rich nuclei

NARCIS (Netherlands)

Dieperink, AEL; Dewulf, Y; Van Neck, D; Waroquier, M; Rodin, [No Value

2003-01-01

The symmetry energy for nuclear matter and its relation to the neutron. skin in finite nuclei is discussed. The symmetry energy as a function of density obtained in a self-consistent Green function approach is presented and compared to the results of other recent theoretical approaches. A partial

Experimental investigations of the neutron contamination in high-energy photon fields at medical linear accelerators

Energy Technology Data Exchange (ETDEWEB)

Brunckhorst, Elin

2009-02-26

The scope of this thesis was to develop a device for the detection of the photoneutron dose inside the high-energy photon field. The photoneutron contamination of a Siemens PRIMUS linear accelerator was investigated in detail in its 15 MV photon mode. The experimental examinations were performed with three ionisation chambers (a tissue equivalent chamber, a magnesium chamber and a {sup 10}B-coated magnesium chamber) and two types of thermoluminescence detectors (enriched with {sup 6}Li and {sup 7}Li, respectively). The detectors have different sensitivities to photons and neutrons and their combination allows the dose separation in a mixed neutron/photon field. The application of the ionisation chamber system, as well as the present TLD system for photoneutron detection in high-energy photon beams is a new approach. The TLD neutron sensitivity was found to be too low for a measurement inside the open photon field and the further investigation focused on the ionisation chambers. The three ionisation chambers were calibrated at different photon and neutron sources and a the borated magnesium chamber showed a very high response to thermal neutrons. For a cross check of the calibration, the three chambers were also used for dose separation of a boron neutron capture therapy beam where the exact determination of the thermal neutron dose is essential. Very accurate results were achieved for the thermal neutron dose component. At the linear accelerator the chamber system was reduced to a paired chamber system utilising the two magnesium chambers, since the fast neutron component was to small to be separated. The neutron calibration of the three chambers could not be applied, instead a conversion of measured thermal neutron signal by the borated chamber to Monte Carlo simulated total neutron dose was performed. Measurements for open fields in solid water and liquid water were performed with the paired chamber system. In larger depths the neutron dose could be determined

Experimental investigations of the neutron contamination in high-energy photon fields at medical linear accelerators

International Nuclear Information System (INIS)

Brunckhorst, Elin

2009-01-01

The scope of this thesis was to develop a device for the detection of the photoneutron dose inside the high-energy photon field. The photoneutron contamination of a Siemens PRIMUS linear accelerator was investigated in detail in its 15 MV photon mode. The experimental examinations were performed with three ionisation chambers (a tissue equivalent chamber, a magnesium chamber and a 10 B-coated magnesium chamber) and two types of thermoluminescence detectors (enriched with 6 Li and 7 Li, respectively). The detectors have different sensitivities to photons and neutrons and their combination allows the dose separation in a mixed neutron/photon field. The application of the ionisation chamber system, as well as the present TLD system for photoneutron detection in high-energy photon beams is a new approach. The TLD neutron sensitivity was found to be too low for a measurement inside the open photon field and the further investigation focused on the ionisation chambers. The three ionisation chambers were calibrated at different photon and neutron sources and a the borated magnesium chamber showed a very high response to thermal neutrons. For a cross check of the calibration, the three chambers were also used for dose separation of a boron neutron capture therapy beam where the exact determination of the thermal neutron dose is essential. Very accurate results were achieved for the thermal neutron dose component. At the linear accelerator the chamber system was reduced to a paired chamber system utilising the two magnesium chambers, since the fast neutron component was to small to be separated. The neutron calibration of the three chambers could not be applied, instead a conversion of measured thermal neutron signal by the borated chamber to Monte Carlo simulated total neutron dose was performed. Measurements for open fields in solid water and liquid water were performed with the paired chamber system. In larger depths the neutron dose could be determined with an

Experimental research of plastic scintillation detector loaded 6Li neutron energy response

International Nuclear Information System (INIS)

Wang Lizong; Zhang Chuanfei; Peng Taiping; Guo Cun; Yang Hongqiong; Zhang Jianhua

2005-01-01

A new plastic scintillator, plastic scintillator loaded 6 Li, is brought forward and developed in this paper in order to increase low energy neutron sensitivity. Neutron sensitivity of several plastic scintillation detectors loaded 6 Li new developed in neutron energy range 0.2 MeV-5.0 MeV are calibrated by direct current at serial accelerator. Energy response curves of the detectors are obtained in this experiment. It is shown that this new plastic scintillation detector can increase low energy neutron sensitivity in experimental results. (authors)

Mirror displacement energies and neutron skins

International Nuclear Information System (INIS)

Duflo, J.; Zuker, A.P.

2002-01-01

A gross estimate of the neutron skin [0.80(5)(N-Z)/A fm] is extracted from experimental proton radii, represented by a four parameter fit, and observed mirror displacement energies (CDE). The calculation of the latter relies on an accurately derived Coulomb energy and smooth averages of the charge symmetry breaking potentials constrained to state of the art values. The only free parameter is the neutron skin itself. The Nolen Schiffer anomaly is reduced to small deviations (rms=127 keV) that exhibit a secular trend. It is argued that with state of the art shell model calculations the anomaly should disappear. Highly accurate fits to proton radii emerge as a fringe benefit

Linear extended neutron diffusion theory for semi-in finites homogeneous means

International Nuclear Information System (INIS)

Vazquez R, R.; Vazquez R, A.; Espinosa P, G.

2009-10-01

Originally developed for heterogeneous means, the linear extended neutron diffusion theory is applied to the limit case of monoenergetic neutron diffusion in a semi-infinite homogeneous mean with a neutron source, located in the coordinate origin situated in the frontier of dispersive material. The monoenergetic neutron diffusion is studied taking into account the spatial deviations in the neutron flux to the interfacial current caused by the neutron source, as well as the influence of the spatial deviations in the absorption rate. The developed pattern is an unidimensional model for an energy group obtained of application of volumetric average diffusion equation in the moderator. The obtained results are compared against the classic diffusion theory and qualitatively against the neutron transport theory. (Author)

Continuous neutron slowing down theory applied to resonances

International Nuclear Information System (INIS)

Segev, M.

1977-01-01

Neutronic formalisms that discretize the neutron slowing down equations in large numerical intervals currently account for the bulk effect of resonances in a given interval by the narrow resonance approximation (NRA). The NRA reduces the original problem to an efficient numerical formalism through two assumptions: resonance narrowness with respect to the scattering bands in the slowing down equations and resonance narrowness with respect to the numerical intervals. Resonances at low energies are narrow neither with respect to the slowing down ranges nor with respect to the numerical intervals, which are usually of a fixed lethargy width. Thus, there are resonances to which the NRA is not applicable. To stay away from the NRA, the continuous slowing down (CSD) theory of Stacey was invoked. The theory is based on a linear expansion in lethargy of the collision density in integrals of the slowing down equations and had notable success in various problems. Applying CSD theory to the assessment of bulk resonance effects raises the problem of obtaining efficient quadratures for integrals involved in the definition of the so-called ''moderating parameter.'' The problem was solved by two approximations: (a) the integrals were simplified through a rationale, such that the correct integrals were reproduced for very narrow or very wide resonances, and (b) the temperature-broadened resonant line shapes were replaced by nonbroadened line shapes to enable analytical integration. The replacement was made in such a way that the integrated capture and scattering probabilities in each resonance were preserved. The resulting formalism is more accurate than the narrow-resonance formalisms and is equally as efficient

Transport of accelerator produced high energy neutrons though concrete

International Nuclear Information System (INIS)

Prabhakar Rao, G.; Sarkar, P.K.

1996-01-01

Development of a computational system for estimating the production and transport of high energy neutrons in particle accelerators is reported. The energy-angle distribution of neutrons from accelerated ions bombarding thick targets is calculated by a hybrid nuclear reaction model code, ALICE-91, modified to suit the purpose. Subsequent transmission of these neutrons through concrete slabs is treated using the anisotropic source-flux iteration technique (ASFIT) in the framework of a coupled neutron-gamma transport. Several parameters of both the codes have been optimized to obtain the transmitted dose through concrete. The calculations are found to be accurate and at the same time faster compared to the detailed Monte Carlo calculations. (author). 8 refs., 2 figs

Preliminary investigations of Monte Carlo Simulations of neutron energy and LET spectra for fast neutron therapy facilities

International Nuclear Information System (INIS)

Kroc, T.K.

2009-01-01

No fast neutron therapy facility has been built with optimized beam quality based on a thorough understanding of the neutron spectrum and its resulting biological effectiveness. A study has been initiated to provide the information necessary for such an optimization. Monte Carlo studies will be used to simulate neutron energy spectra and LET spectra. These studies will be bench-marked with data taken at existing fast neutron therapy facilities. Results will also be compared with radiobiological studies to further support beam quality ptimization. These simulations, anchored by this data, will then be used to determine what parameters might be optimized to take full advantage of the unique LET properties of fast neutron beams. This paper will present preliminary work in generating energy and LET spectra for the Fermilab fast neutron therapy facility.

Improved treatment for determining the group cross section for elastic down-scattering into the adjacent group

International Nuclear Information System (INIS)

Woll, D.

1985-04-01

In the group cross section libraries usually applied for reactor calculations, the energy dependent probabilities of interactions between neutrons and the materials existing in the reactor are represented by weighted average values over certain energy ranges with a neutron energy spectrum regarded as representative. The influence of the resonance structure of the cross sections via the neutron spectrum and the resultant effect on the averaged group cross sections is taken into account in an approximate way by so-called resonance self-shielding factors. The approximations indicated are of considerable importance for the elastic down scattering. They can be improved by the so-called REMO correction, which takes into account the neutron energy distribution existing in the reactor model. Because such detailed neutron distributions are very expensive to prepare, especially in multi-dimensional models, automatic program runs were established which, in some cases by simplifications of the model, allow collision densities to be made available at relatively little expenditure which permit many nuclear quantities to be calculated with a sufficient degree of accuracy. This report describes the program runs set up and the experience acquired in testing them by the examples of the MASURCA 3B experiment and the SNEAK 11B2 assembly. This report deals especially with the influence of the collision density used for the REMO correction on the ksub(eff) value and other parameters of the reactor models considered. (orig.) [de

Fusion materials high energy-neutron studies. A status report

International Nuclear Information System (INIS)

Doran, D.G.; Guinan, M.W.

1980-01-01

The objectives of this paper are (1) to provide background information on the US Magnetic Fusion Reactor Materials Program, (2) to provide a framework for evaluating nuclear data needs associated with high energy neutron irradiations, and (3) to show the current status of relevant high energy neutron studies. Since the last symposium, the greatest strides in cross section development have been taken in those areas providing FMIT design data, e.g., source description, shielding, and activation. In addition, many dosimetry cross sections have been tentatively extrapolated to 40 MeV and integral testing begun. Extensive total helium measurements have been made in a variety of neutron spectra. Additional calculations are needed to assist in determining energy dependent cross sections

Generation of broad-group neutron/photon cross-section libraries for shielding applications

International Nuclear Information System (INIS)

Ingersoll, D.T.; Roussin, R.W.; Fu, C.Y.; White, J.E.

1989-01-01

The generation and use of multigroup cross-section libraries with broad energy group structures is primarily for the economy of computer resources. Also, the establishment of reference broad-group libraries is desirable in order to avoid duplication of effort, both in terms of the data generation and verification, and to assure a common data base for all participants in a specific project. Uncertainties are inevitably introduced into the broad-group cross sections due to approximations in the grouping procedure. The dominant uncertainty is generally with regard to the energy weighting function used to average the pointwise or fine-group data within a single broad group. Intelligent choice of the weighting functions can reduce such uncertainties. Also, judicious selection of the energy group structure can help to reduce the sensitivity of the computed responses to the weighting function, at least for a selected set of problems. Two new multigroup cross section libraries have been recently generated from ENDF/B-V data for two specific shielding applications. The first library was prepared for use in sodium-cooled reactor systems and is available in both broad-group structures. The second library, just recently completed, was prepared for use in air-over-ground environments and is available in a broad-group (46-neutron, 23-photon) energy structure. The selection of the specific group structures and weighting functions was an important part of the generation of both libraries

Computed neutron coincidence counting applied to passive waste assay

Energy Technology Data Exchange (ETDEWEB)

Bruggeman, M.; Baeten, P.; De Boeck, W.; Carchon, R. [Nuclear Research Centre, Mol (Belgium)

1997-11-01

Neutron coincidence counting applied for the passive assay of fissile material is generally realised with dedicated electronic circuits. This paper presents a software based neutron coincidence counting method with data acquisition via a commercial PC-based Time Interval Analyser (TIA). The TIA is used to measure and record all time intervals between successive pulses in the pulse train up to count-rates of 2 Mpulses/s. Software modules are then used to compute the coincidence count-rates and multiplicity related data. This computed neutron coincidence counting (CNCC) offers full access to all the time information contained in the pulse train. This paper will mainly concentrate on the application and advantages of CNCC for the non-destructive assay of waste. An advanced multiplicity selective Rossi-alpha method is presented and its implementation via CNCC demonstrated. 13 refs., 4 figs., 2 tabs.

Computed neutron coincidence counting applied to passive waste assay

International Nuclear Information System (INIS)

Bruggeman, M.; Baeten, P.; De Boeck, W.; Carchon, R.

1997-01-01

Neutron coincidence counting applied for the passive assay of fissile material is generally realised with dedicated electronic circuits. This paper presents a software based neutron coincidence counting method with data acquisition via a commercial PC-based Time Interval Analyser (TIA). The TIA is used to measure and record all time intervals between successive pulses in the pulse train up to count-rates of 2 Mpulses/s. Software modules are then used to compute the coincidence count-rates and multiplicity related data. This computed neutron coincidence counting (CNCC) offers full access to all the time information contained in the pulse train. This paper will mainly concentrate on the application and advantages of CNCC for the non-destructive assay of waste. An advanced multiplicity selective Rossi-alpha method is presented and its implementation via CNCC demonstrated. 13 refs., 4 figs., 2 tabs

Neutron activation: an invaluable technique for teaching applied radiation

International Nuclear Information System (INIS)

Trainer, Matthew

2002-01-01

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

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)

Energy response of neutron area monitor with silicon semiconductor detector

International Nuclear Information System (INIS)

Kitaguchi, Hiroshi; Izumi, Sigeru; Kobayashi, Kaoru; Kaihara, Akihisa; Nakamura, Takashi.

1993-01-01

A prototype neutron area monitor with a silicon semiconductor detector has been developed which has the energy response of 1 cm dose equivalent recommended by the ICRP-26. Boron and proton radiators are coated on the surface of the silicon semiconductor detector. The detector is set at the center of a cylindrical polyethylene moderator. This moderator is covered by a porous cadmium board which serves as the thermal neutron absorber. Neutrons are detected as α-particles generated by the nuclear reaction 10 B(n,α) 7 Li and as recoil protons generated by the interaction of fast neutrons with hydrogen. The neutron energy response of the monitor was measured using thermal neutrons and monoenergetic fast neutrons generated by an accelerator. The response was consistent with the 1 cm dose equivalent response required for the monitor within ±34% in the range of 0.025 - 15 Mev. (author)

Hot neutron stars at birth and energy release

International Nuclear Information System (INIS)

Takatsuka, Tatsuyuki

1994-01-01

For the discussion of hot neutron stars at birth, it is necessary to calculate the equation of state for a so-called 'supernova matter' consisting of a neutron-rich nuclear matter and degenerated leptons. One of the aims of this paper is to obtain the realistic results for the equation of state. In 10-20s after the birth, new born hot neutron stars are cooled down by neutrino diffusion process, and gradually contract to usual cold neutron starts. It is another aim of this paper to determine how much energy is released during this cooling stage. The points to which attention was paid are explained. A three-nucleon interaction was introduced phenomenologically, as a two-nucleon interaction is insufficient to satisfy the empirical saturation property of symmetric nuclear matters. The separation of uncertain part from well-known part has the merit to clarify the dependence of the results on the present theoretical uncertainties. The validity of the simplified calculation as an approximation for the exact calculation is discussed. The results by both calculations were compared for the case of hot symmetric nuclear matters. The comparison of the density profiles for a hot neutron star and a cold neutron star is shown. The binding energy for hot and cold neutron stars was plotted. These results are examined. (K.I.)

Influence of media size on energy distribution of pulsed thermal neutrons

International Nuclear Information System (INIS)

Dabrowska, J.

2007-01-01

The work is devoted to the investigation of the diffusion cooling phenomenon of pulsed thermalized neutron fields in bounded media. It is aimed at the examination of the validity of the neutron temperature model that involves the assumption that an asymptotic energy distribution of neutrons in bounded media can be described by the Maxwell distribution but with a shifted temperature, lower than a temperature of medium. The research carried out entirely by means of Monte Carlo simulation of the neutron transport was preceded by a measurement of the time decay constants obtained in all variants of Monte Carlo simulations of the experiment and the measured one was stated. The form of asymptotic energy distribution of neutrons and its dependence on the size of medium was investigated in three kinds of materials of different thermal neutron transport properties: energy independent scatterer with negligible absorption (silica), energy dependent scatterer with 1/v absorption (borated silica) and energy dependent scatterer with 1/v absorption (water). As it was expected, in the case of large media, which can be treated as infinite, neutrons attained the Maxwell energy distribution at the temperature of the medium. For all materials under investigation the average and the most probable values of the energy distribution steadily decreased with decreasing geometric dimensions of the media. At the same time a growing distortion from the pure Maxwellian energy distribution was observed, which means that the concept of the neutron temperature fails in the case of small media. Although the spectra under investigation in general did not have the Maxwellian shape, the most probable velocity in a neutron density distribution decreased linearly with the increasing geometric buckling of the medium. This dependence manifested a stronger cooling than the one predicted by a certain approximate formula. The neutron spectrum in a small medium of pure silica was cooler than the spectrum in

High energy neutron source for materials research and development

International Nuclear Information System (INIS)

Odera, M.

1989-01-01

Requirements for neutron source for nuclear materials research are reviewed and ESNIT, Energy Selective Neutron Irradiation Test facility proposed by JAERI is discussed. Its principal aims of a wide neutron energy tunability and spectra peaking at each energy to enable characterization of material damage process are demanding but attractive goals which deserve detailed study. It is also to be noted that the requirements make a difference in facility design from those of FMIT, IFMIF and other high energy intense neutron sources built or planned to date. Areas of technologies to be addressed to realize the ESNIT facility are defined and discussed. In order to get neutron source having desired spectral characteristics keeping moderate intensity, projectile and target combinations must be examined including experimentation if necessary. It is also desired to minimize change of flux density and energy spectrum according to location inside irradiation chamber. Extended target or multiple targets configuration might be a solution as well as specimen rotation and choice of combination of projectile and target which has minimum velocity of the center of mass. Though relevant accelerator technology exists, it is to be stressed that considerable efforts must be paid, especially in the area of target and irradiation devices to get ESNIT goal. Design considerations to allow hands-on maintenance and future upgrading possibility are important either, in order to exploit the facility fully for nuclear materials research and development. (author)

Applied research and development of neutron activation analysis

International Nuclear Information System (INIS)

Chung, Yong Sam; Moon, Jong Hwa; Kim, Sun Ha; Baek, Sung Ryel; Kim, Young Gi; Jung, Hwan Sung; Park, Kwang Won; Kang, Sang Hun; Lim, Jong Myoung

2003-05-01

The aims of this project are to establish the quality control system of Neutron Activation Analysis(NAA) due to increase of industrial needs for standard analytical method and to prepare and identify the standard operation procedure of NAA through practical testing for different analytical items. R and D implementations of analytical quality system using neutron irradiation facility and gamma-ray measurement system and automation of NAA facility in HANARO research reactor are as following ; 1) Establishment of NAA quality control system for the maintenance of best measurement capability and the promotion of utilization of HANARO research reactor 2) Improvement of analytical sensitivity for industrial applied technologies and establishment of certified standard procedures 3) Standardization and development of Prompt Gamma-ray Activation Analysis (PGAA) technology

Applied research of environmental monitoring using instrumental neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Chung, Young Sam; Moon, Jong Hwa; Chung, Young Ju

1997-08-01

This technical report is written as a guide book for applied research of environmental monitoring using Instrumental Neutron Activation Analysis. The contents are as followings; sampling and sample preparation as a airborne particulate matter, analytical methodologies, data evaluation and interpretation, basic statistical methods of data analysis applied in environmental pollution studies. (author). 23 refs., 7 tabs., 9 figs.

Nuclear symmetry energy and stability of matter in neutron stars

International Nuclear Information System (INIS)

Kubis, Sebastian

2007-01-01

It is shown that the nuclear symmetry energy is the key quantity in the stability consideration in neutron star matter. The symmetry energy controls the position of crust-core transition and also may lead to new effects in the inner core of neutron star

Training of reverse propagation neural networks applied to neutron dosimetry

International Nuclear Information System (INIS)

Hernandez P, C. F.; Martinez B, M. R.; Leon P, A. A.; Espinoza G, J. G.; Castaneda M, V. H.; Solis S, L. O.; Castaneda M, R.; Ortiz R, M.; Vega C, H. R.; Mendez V, R.; Gallego, E.; De Sousa L, M. A.

2016-10-01

Neutron dosimetry is of great importance in radiation protection as aims to provide dosimetric quantities to assess the magnitude of detrimental health effects due to exposure of neutron radiation. To quantify detriment to health is necessary to evaluate the dose received by the occupationally exposed personnel using different detection systems called dosimeters, which have very dependent responses to the energy distribution of neutrons. The neutron detection is a much more complex problem than the detection of charged particles, since it does not carry an electric charge, does not cause direct ionization and has a greater penetration power giving the possibility of interacting with matter in a different way. Because of this, various neutron detection systems have been developed, among which the Bonner spheres spectrometric system stands out due to the advantages that possesses, such as a wide range of energy, high sensitivity and easy operation. However, once obtained the counting rates, the problem lies in the neutron spectrum deconvolution, necessary for the calculation of the doses, using different mathematical methods such as Monte Carlo, maximum entropy, iterative methods among others, which present various difficulties that have motivated the development of new technologies. Nowadays, methods based on artificial intelligence technologies are being used to perform neutron dosimetry, mainly using the theory of artificial neural networks. In these new methods the need for spectrum reconstruction can be eliminated for the calculation of the doses. In this work an artificial neural network or reverse propagation was trained for the calculation of 15 equivalent doses from the counting rates of the Bonner spheres spectrometric system using a set of 7 spheres, one of 2 spheres and two of a single sphere of different sizes, testing different error values until finding the most appropriate. The optimum network topology was obtained through the robust design

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

CERN Document Server

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

2002-01-01

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

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)

Simulation of neutron rich nuclei production through 239U fission at intermediates energies

International Nuclear Information System (INIS)

Mirea, M.; Clapier, F.; Pauwels, N.; Proust, J.

1997-01-01

The theoretical part and some results obtained from a model realised for fission processes in wide range of mass-asymmetries are presented. The fission barriers are computed in a tridimensional configuration space using the Yukawa - plus - exponential macroscopic energies corrected within the Strutinsky procedure. It is assumed that channel probabilities are proportional with Gamow penetrabilities. The model is applied for the disintegration of the 239 U in order to determine the relative yields for the production of neutron rich nuclei at diverse intermediate energies. (author)

A coaxial double cylindrical TEPC for the microdosimetry of selected neutron energy bands in mixed fields of fast neutrons

Energy Technology Data Exchange (ETDEWEB)

Saion, E.B.; Watt, D.E. (Saint Andrews Univ. (UK). Dept. of Physics); East, B.W. (Scottish Universities Research and Reactor Centre, Glasgow (UK)); Colautti, P. (Istituto Nazionale di Fisica Nucleare, Padua (Italy))

1990-01-01

A new low pressure tissue-equivalent proportional counter (TEPC) in a coaxial double cylindrical form has been developed to measure separately the microdose spectrum from any desired energy band of neutrons in the presence of mixed fields of faster neutrons, by selecting the thickness of the common TE dividing wall to be equivalent to the corresponding maximum proton ranges and by appropriate use of coincidence/anti-coincidence pulse arrangements. This thickness ensures charged particle equilibrium for the relevant neutron energy. Event spectra due to recoils generated by faster neutrons which interact with both the counters are removed completely by anti-coincidence techniques, thereby optimising the sensitivity of the inner microdosemeter to the event spectra of interest. The ability of this counter to discriminate in favour of events due to neutrons of energy <850 keV was achieved in microdosimetric measurements from mixed fields of a nuclear reactor. Mean values of lineal energy and quality factor for neutrons of energy <850 keV from a nuclear reactor were determined from the anti-coincidence spectrum. Good discrimination against {gamma} ray induced events is also achieved for the spectrum recorded in the anti-coincidence mode. This is an advantageous feature for other applications and requires further investigation. (author).

A coaxial double cylindrical TEPC for the microdosimetry of selected neutron energy bands in mixed fields of fast neutrons

International Nuclear Information System (INIS)

Saion, E.B.; Watt, D.E.; Colautti, P.

1990-01-01

A new low pressure tissue-equivalent proportional counter (TEPC) in a coaxial double cylindrical form has been developed to measure separately the microdose spectrum from any desired energy band of neutrons in the presence of mixed fields of faster neutrons, by selecting the thickness of the common TE dividing wall to be equivalent to the corresponding maximum proton ranges and by appropriate use of coincidence/anti-coincidence pulse arrangements. This thickness ensures charged particle equilibrium for the relevant neutron energy. Event spectra due to recoils generated by faster neutrons which interact with both the counters are removed completely by anti-coincidence techniques, thereby optimising the sensitivity of the inner microdosemeter to the event spectra of interest. The ability of this counter to discriminate in favour of events due to neutrons of energy <850 keV was achieved in microdosimetric measurements from mixed fields of a nuclear reactor. Mean values of lineal energy and quality factor for neutrons of energy <850 keV from a nuclear reactor were determined from the anti-coincidence spectrum. Good discrimination against γ ray induced events is also achieved for the spectrum recorded in the anti-coincidence mode. This is an advantageous feature for other applications and requires further investigation. (author)

A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy

International Nuclear Information System (INIS)

Camargo, Dayana Queiroz de

2011-01-01

This thesis has developed a stochastic model to simulate the neutrons transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using Monte Carlo method for the propagation of neutrons in different environment. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational processing time introduced the variable control volume along the (pseudo-) periodic boundary conditions in order to overcome this problem. The choice of class physical Monte Carlo is due to the fact that it can decompose into simpler constituents the problem of solve a transport equation. The components may be treated separately, these are the propagation and interaction while respecting the laws of energy conservation and momentum, and the relationships that determine the probability of their interaction. We are aware of the fact that the problem approached in this thesis is far from being comparable to building a nuclear reactor, but this discussion the main target was to develop the Monte Carlo model, implement the code in a computer language that allows extensions of modular way. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude the need to consider the energy dependence, i.e. an spectral effective multiplication factor should be introduced each energy group separately. (author)

Neutron cross section libraries for analysis of fusion neutronics experiments

International Nuclear Information System (INIS)

Kosako, Kazuaki; Oyama, Yukio; Maekawa, Hiroshi; Nakamura, Tomoo

1988-03-01

We have prepared two computer code systems producing neutron cross section libraries to analyse fusion neutronics experiments. First system produces the neutron cross section library in ANISN format, i.e., the multi-group constants in group independent format. This library can be obtained by using the multi-group constant processing code system MACS-N and the ANISN format cross section compiling code CROKAS. Second system is for the continuous energy cross section library for the MCNP code. This library can be obtained by the nuclear data processing system NJOY which generates pointwise energy cross sections and the cross section compiling code MACROS for the MCNP library. In this report, we describe the production procedures for both types of the cross section libraries, and show six libraries with different conditions in ANISN format and a library for the MCNP code. (author)

Detecting neutrons by forward recoil protons at the Energy & Transmutation facility: Detector development and calibration with 14.1-MeV neutrons

Science.gov (United States)

Afanasev, S.; Vishnevskiy, A.; Vishnevskiy, D.; Rogachev, A.; Tyutyunnikov, S.

2017-05-01

As part of the Energy & Transmutation project, we are developing a detector for neutrons with energies in the 10-100 MeV range emitted from the target irradiated by a charged-particle beam. The neutron is detected by measuring the time-of-flight and total kinetic energy of the forward-going recoil proton [1] knocked out at a small angle from a thin layer of plastic scintillator, which has to be selected against an intense background created by γ quanta, scattered neutrons, and charged particles. On the other hand, neutron energy has to be measured over the full range with no extra tuning of the detector operation regime. Initial measurements with a source of 14.1-MeV neutrons are reported.

Simulations and developments of the Low Energy Neutron detector Array LENA

International Nuclear Information System (INIS)

Langer, C.; Algora, A.; Couture, A.; Csatlós, M.; Gulyás, J.; Heil, M.; Krasznahorkay, A.; O'Donnell, J.M.; Plag, R.; Reifarth, R.; Stuhl, L.; Sonnabend, K.; Tornyi, T.; Tovesson, F.

2011-01-01

Prototypes of the Low Energy Neutron detector Array (LENA) have been tested and compared with detailed GEANT simulations. LENA will consist of plastic scintillation bars with the dimensions 1000×45×10 mm 3 . The tests have been performed with γ-ray sources and neutrons originating from the neutron-induced fission of 235 U. The simulations agreed very well with the measured response and were therefore used to simulate the response to mono-energetic neutrons with different detection thresholds. LENA will be used to detect low-energy neutrons from (p,n)-type reactions with low momentum transfer foreseen at the R 3 B and EXL setups at FAIR, Darmstadt.

Simulation of a high energy neutron irradiation facility at beamline 11 of the China Spallation Neutron Source

Energy Technology Data Exchange (ETDEWEB)

Tairan, Liang [School of Physics and Electronic Information Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Zhiduo, Li [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Wen, Yin, E-mail: wenyin@aphy.iphy.ac.cn [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Fei, Shen [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Quanzhi, Yu [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Tianjiao, Liang [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China)

2017-07-11

The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm{sup 2}/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.

Basic and Applied Research at the Los Alamos Neutron Science Center

International Nuclear Information System (INIS)

Lisowski, P.W.

2003-01-01

The Los Alamos Neutron Science Center, or LANSCE, is an accelerator-based national user facility for research in basic and applied science. At present LANSCE has two experimental areas primarily using neutrons generated by 800-MeV protons striking tungsten target systems. A third area uses the proton beam for radiography. This paper describes the three LANSCE experimental areas, gives highlights of the past operating period, and discusses plans for the future

Ionizing Energy Depositions After Fast Neutron Interactions in Silicon

CERN Document Server

Bergmann, Benedikt; Caicedo, Ivan; Kierstead, James; Takai, Helio; Frojdh, Erik

2016-01-01

In this study we present the ionizing energy depositions in a 300 μm thick silicon layer after fast neutron impact. With the Time-of-Flight (ToF) technique, the ionizing energy deposition spectra of recoil silicons and secondary charged particles were assigned to (quasi-)monoenergetic neutron energies in the range from 180 keV to hundreds of MeV. We show and interpret representative measured energy spectra. By separating the ionizing energy losses of the recoil silicon from energy depositions by products of nuclear reactions, the competition of ionizing (IEL) and non-ionizing energy losses (NIEL) of a recoil silicon within the silicon lattice was investigated. The data give supplementary information to the results of a previous measurement and are compared with different theoretical predictions.

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...

Applicability of the two-angle differential method to response measurement of neutron-sensitive devices at the RCNP high-energy neutron facility

Energy Technology Data Exchange (ETDEWEB)

Masuda, Akihiko, E-mail: aki-masuda@aist.go.jp [National Institute of Advanced Industrial Science and Technology (AIST), 1-1-1 Umezono, Tsukuba, Ibaraki 305-8568 (Japan); Matsumoto, Tetsuro [National Institute of Advanced Industrial Science and Technology (AIST), 1-1-1 Umezono, Tsukuba, Ibaraki 305-8568 (Japan); Iwamoto, Yosuke [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai, Naka, Ibaraki 319-1195 (Japan); Hagiwara, Masayuki [High Energy Accelerator Research Organization (KEK), 1-1 Oho, Tsukuba, Ibaraki 305-0801 (Japan); Satoh, Daiki; Sato, Tatsuhiko [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai, Naka, Ibaraki 319-1195 (Japan); Iwase, Hiroshi [High Energy Accelerator Research Organization (KEK), 1-1 Oho, Tsukuba, Ibaraki 305-0801 (Japan); Yashima, Hiroshi [Research Reactor Institute, Kyoto University, 2-1010 Asashiro-nishi, Kumatori, Sennan, Osaka 590-0494 (Japan); Nakane, Yoshihiro [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai, Naka, Ibaraki 319-1195 (Japan); Nishiyama, Jun [Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo 152-8550 (Japan); Shima, Tatsushi; Tamii, Atsushi; Hatanaka, Kichiji [Research Center for Nuclear Physics (RCNP), Osaka University, 10-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan); Harano, Hideki [National Institute of Advanced Industrial Science and Technology (AIST), 1-1-1 Umezono, Tsukuba, Ibaraki 305-8568 (Japan); Nakamura, Takashi [Cyclotron and Radioisotope Center (CYRIC), Tohoku University, 6-3 Aramaki, Aoba, Sendai, Miyagi 980-8578 (Japan)

2017-03-21

Quasi-monoenergetic high-energy neutron fields induced by {sup 7}Li(p,n) reactions are used for the response evaluation of neutron-sensitive devices. The quasi-monoenergetic high-energy field consists of high-energy monoenergetic peak neutrons and unwanted continuum neutrons down to the low-energy region. A two-angle differential method has been developed to compensate for the effect of the continuum neutrons in the response measurements. In this study, the two-angle differential method was demonstrated for Bonner sphere detectors, which are typical examples of moderator-based neutron-sensitive detectors, to investigate the method's applicability and its dependence on detector characteristics. Experiments were performed under 96–387 MeV quasi-monoenergetic high-energy neutron fields at the Research Center for Nuclear Physics (RCNP), Osaka University. The measurement results for large high-density polyethylene (HDPE) sphere detectors agreed well with Monte Carlo calculations, which verified the adequacy of the two-angle differential method. By contrast, discrepancies were observed in the results for small HDPE sphere detectors and metal-induced sphere detectors. The former indicated that detectors that are particularly sensitive to low-energy neutrons may be affected by penetrating neutrons owing to the geometrical features of the RCNP facility. The latter discrepancy could be consistently explained by a problem in the evaluated cross-section data for the metals used in the calculation. Through those discussions, the adequacy of the two-angle differential method was experimentally verified, and practical suggestions were made pertaining to this method.

Fundamental physics with low-energy neutrons

International Nuclear Information System (INIS)

Barrón-Palos, Libertad

2016-01-01

Low-energy neutrons are playing a prominent role in a growing number of fundamental physics studies. This paper provides a brief description of the physics that some of the experiments in the area are addressing. (paper)

Criticality problems in energy dependent neutron transport theory

International Nuclear Information System (INIS)

Victory, H.D. Jr.

1979-01-01

The criticality problem is considered for energy dependent neutron transport in an isotropically scattering, homogeneous slab. Under a positivity assumption on the scattering kernel, an expression can be found relating the thickness of the slab to a parameter characterizing production by fission. This is accomplished by exploiting the Perron-Frobenius-Jentsch characterization of positive operators (i.e. those leaving invariant a normal, reproducing cone in a Banach space). It is pointed out that those techniques work for classes of multigroup problems were the Case singular eigenfunction approach is not as feasible as in the one-group theory, which is also analyzed

The generation, validation and testing of a coupled 219-group neutron 36-group gamma ray AMPX-II library

International Nuclear Information System (INIS)

Panini, G.C.; Siciliano, F.; Lioi, A.

1987-01-01

The main characteristics of a P 3 coupled 219-group neutron 36-group gamma-ray library in the AMPX-II Master Interface Format obtained processing ENDF/B-IV data by means of various AMPX-II System modules are presented in this note both for the more reprocessing aspects and features of the generated component files-neutrons, photon and secondary gamma-ray production cross sections. As far as the neutron data are concerned there is the avaibility of 186 data sets regarding most significant fission products. Results of the additional validation of the neutron data pertaining to eighteen benchmark experiments are also given. Some calculational tests on both neutron and coupled data emphasize the important role of the secondary gamma-ray data in nuclear criticality safety calculations

Verification of the DUCT-III for calculation of high energy neutron streaming

Energy Technology Data Exchange (ETDEWEB)

Masukawa, Fumihiro; Nakano, Hideo; Nakashima, Hiroshi; Sasamoto, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tayama, Ryu-ichi; Handa, Hiroyuki; Hayashi, Katsumi [Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan); Hirayama, Hideo [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Shin, Kazuo [Kyoto Univ., Kyoto (Japan)

2003-03-01

A large number of radiation streaming calculations under a variety of conditions are required as a part of shielding design for a high energy proton accelerator facility. Since sophisticated methods are very time consuming, simplified methods are employed in many cases. For accuracy evaluation of a simplified code DUCT-III for high energy neutron streaming calculations, two kinds of benchmark problems based on the experiments were analyzed. Through comparison of the DUCT-III calculations with both the measurements and the sophisticated Monte Carlo calculations, DUCT-III was seen reliable enough for applying to the shielding design for the Intense Proton Accelerator Facility. (author)

Verification of the DUCT-III for calculation of high energy neutron streaming

CERN Document Server

Masukawa, F; Hayashi, K; Hirayama, H; Nakano, H; Nakashima, H; Sasamoto, N; Shin, K; Tayama, R I

2003-01-01

A large number of radiation streaming calculations under a variety of conditions are required as a part of shielding design for a high energy proton accelerator facility. Since sophisticated methods are very time consuming, simplified methods are employed in many cases. For accuracy evaluation of a simplified code DUCT-III for high energy neutron streaming calculations, two kinds of benchmark problems based on the experiments were analyzed. Through comparison of the DUCT-III calculations with both the measurements and the sophisticated Monte Carlo calculations, DUCT-III was seen reliable enough for applying to the shielding design for the Intense Proton Accelerator Facility.

Epithermal neutron activation analysis in applied microbiology

International Nuclear Information System (INIS)

Marina Frontasyeva

2012-01-01

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

A compound parabolic concentrator as an ultracold neutron spectrometer

Energy Technology Data Exchange (ETDEWEB)

Hickerson, K.P., E-mail: hickerson@gmail.com; Filippone, B.W., E-mail: bradf@caltech.edu

2013-09-01

The design principles of nonimaging optics are applied to ultracold neutrons (UCN). In particular a vertical compound parabolic concentrator (CPC) that efficiently redirects UCN vertically into a bounded spatial volume where they have a maximum energy mga that depends only on the initial phase space cross sectional area πa{sup 2} creates a spectrometer which can be applied to neutron lifetime and gravitational quantum state experiments. -- Highlights: • Nonimaging optics is applied to ultracold neutrons. • A novel ultracold neutron spectrometer is discussed. • New uses may include a neutron lifetime experiment.

A compound parabolic concentrator as an ultracold neutron spectrometer

International Nuclear Information System (INIS)

Hickerson, K.P.; Filippone, B.W.

2013-01-01

The design principles of nonimaging optics are applied to ultracold neutrons (UCN). In particular a vertical compound parabolic concentrator (CPC) that efficiently redirects UCN vertically into a bounded spatial volume where they have a maximum energy mga that depends only on the initial phase space cross sectional area πa 2 creates a spectrometer which can be applied to neutron lifetime and gravitational quantum state experiments. -- Highlights: • Nonimaging optics is applied to ultracold neutrons. • A novel ultracold neutron spectrometer is discussed. • New uses may include a neutron lifetime experiment

The Martin Marietta Energy Systems personnel neutron dosimetry program

International Nuclear Information System (INIS)

McMahan, K.L.

1991-01-01

Martin Marietta Energy Systems, Inc. (Energy Systems), manages five sites for the US Department of Energy. Personnel dosimetry for four of the five sites is coordinated through a Centralized External Dosimetry System (CEDS). These four sites are the Oak Ridge National Laboratory (ORNL), the Oak Ridge Y-12 Plant (Y-12), the Oak Ridge K-25 Site (K-25), and the Paducah Gaseous Diffusion Plant (PGDP). The fifth Energy Systems site, Portsmouth Gaseous Diffusion Plant, has an independent personnel dosimetry program. The current CEDS personnel neutron dosimeter was first issued in January 1989, after an evaluation and characterization of the dosimeters' response in the workplaces was performed. For the workplace characterization, Energy Systems contracted with Pacific Northwest Laboratory (PNL) to perform neutron measurements at selected locations at ORNL and Y-12. K-25 and PGDP were not included because their neutron radiation fields were similar to others already planned for characterization at ORNL and Y-12. Since the initial characterization, PNL has returned to Oak Ridge twice to perform follow up measurements, and another visit is planned in the near future

Analysis of cavity effect on space- and time-dependent fast and thermal neutron energy spectra

International Nuclear Information System (INIS)

Kudo, Katsuhisa; Narita, Masakuni; Ozawa, Yasutomo.

1975-01-01

The effects of the presence of a central cavity on the space- and time-dependent neutron energy spectra in both thermal and fast neutron systems are analyzed theoretically with use made of the multi-group one-dimensional time-dependent Ssub(n) method. The thermal neutron field is also analyzed for the case of a fundamental time eigenvalue problem with the time-dependent P 1 approximation. The cavity radius is variable, and the system radius for graphite is 120 cm and for the other materials 7 cm. From the analysis of the time-dependent Ssub(n) calculations in the non-multiplying systems of polythene, light water and graphite, cavity heating is the dominant effect for the slowing-down spectrum in the initial period following fast neutron burst, and when the slowing-down spectrum comes into the thermal energy region, cavity heating shifts to cavity cooling. In the multiplying system of 235 U, cavity cooling also takes place as the spectrum approaches equilibrium after the fast neutron burst is injected. The mechanism of cavity cooling is explained analytically for the case of thermal neutron field to illustrate its physical aspects, using the time-dependent P 1 approximation. An example is given for the case of light water. (auth.)

Implosion anisotropy of neutron kinetic energy distributions as measured with the neutron time-of-flight diagnostics at the National Ignition Facility

Science.gov (United States)

Hartouni, Edward; Eckart, Mark; Field, John; Grim, Gary; Hatarik, Robert; Moore, Alastair; Munro, David; Sayer, Daniel; Schlossberg, David

2017-10-01

Neutron kinetic energy distributions from fusion reactions are characterized predominantly by the excess energy, Q, of the fusion reaction and the variance of kinetic energy which is related to the thermal temperature of the plasma as shown by e.g. Brysk. High statistics, high quality neutron time-of-flight spectra obtained at the National Ignition Facility provide a means of measuring small changes to the neutron kinetic energy due to the spatial and temporal distribution of plasma temperature, density and velocity. The modifications to the neutron kinetic energy distribution as described by Munro include plasma velocity terms with spatial orientation, suggesting that the neutron kinetic energy distributions could be anisotropic when viewed by multiple lines-of-sight. These anisotropies provide a diagnostic of burn averaged plasma velocity distributions. We present the results of measurements made for a variety of DT implosions and discuss their possible physical interpretations. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344. Lawrence Livermore National Security, LLC.

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

NESTLE: Few-group neutron diffusion equation solver utilizing the nodal expansion method for eigenvalue, adjoint, fixed-source steady-state and transient problems

International Nuclear Information System (INIS)

Turinsky, P.J.; Al-Chalabi, R.M.K.; Engrand, P.; Sarsour, H.N.; Faure, F.X.; Guo, W.

1994-06-01

NESTLE is a FORTRAN77 code that solves the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM). NESTLE can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source steady-state; or external fixed-source. or eigenvalue initiated transient problems. The code name NESTLE originates from the multi-problem solution capability, abbreviating Nodal Eigenvalue, Steady-state, Transient, Le core Evaluator. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two or four energy groups can be utilized, with all energy groups being thermal groups (i.e. upscatter exits) if desired. Core geometries modelled include Cartesian and Hexagonal. Three, two and one dimensional models can be utilized with various symmetries. The non-linear iterative strategy associated with the NEM method is employed. An advantage of the non-linear iterative strategy is that NSTLE can be utilized to solve either the nodal or Finite Difference Method representation of the few-group neutron diffusion equation

Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

Science.gov (United States)

Colonna, N; Altieri, S

2002-06-01

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

Surface energy of very neutron rich nuclei

CERN Document Server

Von Groote, H

1976-01-01

For a microscopic model calculation of the nuclear surface-energy coefficient sigma the surface energy is defined as the energy loss of an uncharged, semiinfinite (inhomogeneous) two-component system compared to an infinite (homogeneous) system with the same particle asymmetry delta . Using the Thomas-Fermi model the calculations are performed for a series of systems with increasing delta , starting from symmetric matter ( delta =0) and extending beyond the drip line of the neutrons, until the system undergoes a phase transition to a homogeneous system. The results for the surface energy as well as for the neutron skin and for the surface diffuseness are compared to the macroscopic approach of the Droplet Model (DM), which turns out to be a good approximation for small asymmetries typical for the region of the valley of beta -stability. For larger asymmetries, close to the drip lines, terms of higher order than contained in the DM approach are no longer negligible. Beyond the drip lines the pressure of the ou...

METHODS OF ASSESSMENT OF THE RELATIVE BIOLOGICAL EFFECTIVENESS OF NEUTRONS IN NEUTRON THERAPY

Directory of Open Access Journals (Sweden)

V. A. Lisin

2017-01-01

Full Text Available The relative biological effectiveness (RBE of fast neutrons is an important factor influencing the quality of neutron therapy therefore, the assessment of RBE is of great importance. Experimental and clinical studies as well as different mathematical and radiobiological models are used for assessing RBE. Research is conducted for neutron sources differing in the method of producing particles, energy and energy spectrum. Purpose: to find and analyze the dose-dependence of fast neutron RBE in neutron therapy using the U-120 cyclotron and NG-12I generator. Material and methods: The optimal method for assessing the relative biological effectiveness of neutrons for neutron therapy was described. To analyze the dependence of the RBE on neutron dose, the multi-target model of cell survival was applied. Results: The dependence of the RBE of neutrons produced from the U-120 cyclotron and NG-120 generator on the dose level was found for a single irradiation of biological objects. It was shown that the function of neutron dose was consistent with similar dependencies found by other authors in the experimental and clinical studies.

Development of neutron-monitor detectors applicable for energies up to 100 MeV

Energy Technology Data Exchange (ETDEWEB)

Sato, Tatsuhiko; Endo, Akira; Yamaguchi, Yasuhiro; Kim, Eunjoo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakamura, Takashi [Tohoku Univ., Sendai, Miyagi (Japan)

2003-03-01

For the purpose of monitoring of neutron doses in high energy accelerator facilities, we have been developing neutron detectors which are applicable for neutron energies up to 100 MeV. The present paper reports characteristics of a phoswitch-type neutron detector which is composed of a liquid organic scintillator and {sup 6}Li+ZnS(Ag) sheets. (author)

Bench mark spectra for high-energy neutron dosimetry

International Nuclear Information System (INIS)

Dierckx, R.

1986-01-01

To monitor radiation damage experiments, activation detectors are commonly used. The precision of the results obtained by the multiple foil analysis is largely increased by the intercalibration in bench-mark spectra. This technique is already used in dosimetry measurements for fission reactors. To produce neutron spectra similar to fusion reactor and high-energy high-intensity neutron sources (d-Li or spallation), accelerators can be used. Some possible solutions as p-Be and d-D 2 O neutron sources, useful as bench-mark spectra are described. (author)

Energy spectra of fast neutrons by nuclear emulsion method

International Nuclear Information System (INIS)

Quaresma, A.A.

1977-01-01

An experimental method which uses nuclear emulsion plates to determine the energy spectrum of fission neutrons is described. By using this technique, we have obtained the energy distribution of neutrons from spontaneous fission of Cf 2 5 2 . The results are in good agreement with whose obtained previously by others authors who have used different detection techniques, and they are consistent with a Maxwellian distribution as expected by Weisskopf's nuclear evaporation theory. (author)

Recent developments in very low energy neutron technology

International Nuclear Information System (INIS)

Utsuro, Masahiko; Kawabata, Yuji; Yamaguchi, Akira; Yoshiki, Hajime.

1993-01-01

In this report, the recent state of the research and technical development of the neutrons in the energy region below 0.5 meV is introduced. The neutrons in this region are further divided into very cold neutrons (VCN) and ultracold neutrons (UCN). The UCNs are known by such characteristic behavior that they can be confined in a neutron bottle for long time. The attempt to verify the break of T conversion symmetry using neutrons is carried out. The experiment to show the break of T conversion symmetry by grasping the asymmetry of particle emission accompanying the beta decay of polarized neutrons is conceivable. In these cases, the use of UCNs in neutron bottles is effective. The optical properties of VCNs and UCNs are peculiar and resemble to those of light. The only VCN source in Japan is installed in the liquid deuterium CN source in the graphite facility of the KUR. VCNs are taken out from the reactor, and are converted to UCNs using a neutron turbine. The characteristics of an UCN bottle were measured, and the life of neutrons was determined as 887.6 ± 3s. The UCN experiment using superfluid helium was carried out, and the application of gravity to UCN spectrometry was developed as NESSIE. (K.I.)

Neutron dose measurements with the GSI ball at high energy accelerators

International Nuclear Information System (INIS)

Fehrenbacher, G.; Gutermuth, F.; Radon, T.; Kozlova, E.

2005-01-01

Full text: At high energy particle accelerators the production of neutron radiation dominates radiation protection. For the radiation survey at accelerators there is a need for reliable detection systems (passive radiation monitors), which can measure the dose for a wide range of neutron energies independently on the beam pulse structure of the produced radiation. In this work a passive neutron dosemeter for the measurement of the ambient dose equivalent is presented. The dosemeter is suitable for measurements of the emerging neutron radiation at accelerators for the whole energy range up to about 10 GeV. The dosemeter consists of a polyethylene sphere, TL elements (pairs of TLD600/700) and an additional lead layer (PE/Pb) in neutron fields at high energy accelerators is investigated in this work. Results of dose measurements which were performed in realistic neutron fields at the high energy accelerator SPS at CERN (CERF facility) and in Cave A at the heavy ion synchrotron SIS at GSI are presented. The results of these measurements are compared with the expected dose values from the neutron spectra determined for the measurement positions at CERF and in Cave A (FLUKA) and with the dosemeter response derived by the calculated response functions (FLUKA) folded with the neutron spectra. The comparisons show that the additional lead layer in the PE/Pb-sphere improves significantly the response of the dosemeter. The response of the PE/Pb-sphere is 40 to 50 % higher at CERF and Cave A in comparison to the bare PE-sphere. At CERF the dose values of the PE/Pb-sphere is about 25 % lower than the expected dose value, whilst for Cave A, a rather good agreement was found (2 % deviation). (author)

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.

Experimental verification of a method to create a variable energy neutron beam from a monoenergetic, isotropic source using neutron elastic scatter and time of flight

Energy Technology Data Exchange (ETDEWEB)

Whetstone, Zachary D., E-mail: zacwhets@umich.edu; Flaska, Marek, E-mail: mflaska@umich.edu; Kearfott, Kimberlee J., E-mail: kearfott@umich.edu

2016-08-11

An experiment was performed to determine the neutron energy of near-monoergetic deuterium–deuterium (D–D) neutrons that elastically scatter in a hydrogenous target. The experiment used two liquid scintillators to perform time of flight (TOF) measurements to determine neutron energy, with the start detector also serving as the scatter target. The stop detector was placed 1.0 m away and at scatter angles of π/6, π/4, and π/3 rad, and 1.5 m at a scatter angle of π/4 rad. When discrete 1 ns increments were implemented, the TOF peaks had estimated errors between −21.2 and 3.6% relative to their expected locations. Full widths at half-maximum (FWHM) ranged between 9.6 and 20.9 ns, or approximately 0.56–0.66 MeV. Monte Carlo simulations were also conducted that approximated the experimental setup and had both D–D and deuterium–tritium (DT) neutrons. The simulated results had errors between −17.2 and 0.0% relative to their expected TOF peaks when 1 ns increments were applied. The largest D–D and D–T FWHMs were 26.7 and 13.7 ns, or approximately 0.85 and 4.98 MeV, respectively. These values, however, can be reduced through manipulation of the dimensions of the system components. The results encourage further study of the neutron elastic scatter TOF system with particular interest in application to active neutron interrogation to search for conventional explosives.

Compact neutron generator with nanotube ion source

Science.gov (United States)

Chepurnov, A. S.; Ionidi, V. Y.; Ivashchuk, O. O.; Kirsanov, M. A.; Kitsyuk, E. P.; Klenin, A. A.; Kubankin, A. S.; Nazhmudinov, R. M.; Nikulin, I. S.; Oleinik, A. N.; Pavlov, A. A.; Shchagin, A. V.; Zhukova, P. N.

2018-02-01

In this letter, we report the observation of fast neutrons generated when a positive acceleration potential is applied to an array of orientated carbon nanotubes, which are used as an ion source. The neutrons with energy of 2.45 MeV are generated as a result of D-D fusion reaction. The dependencies of the neutron yield on the value of the applied potential and residual pressure of deuterium are measured. The proposed approach is planned to be used for the development of compact neutron generators.

Thermal neutron moderating device

International Nuclear Information System (INIS)

Takigami, Hiroyuki.

1995-01-01

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

The neutron electric dipole moment in left-right symmetric low energy supergravity

International Nuclear Information System (INIS)

Ahn, Y.J.

1984-01-01

We compute the neutron electric dipole moment in low energy supergravity based on the gauge group SU(2)sub(L) x SU(2)sub(R) x U(1)sub(B-L). We find the electric dipole moment dsub(n) -25 e cm x (CP violating phase) provided the left-right symmetry breaking scale > or approx. 10 3 GeV. (orig.)

Development and characterization of real-time wide-energy range personal neutron dosimeter

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Takashi; Tsujimura, Norio (Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center); Yamano, Toshiya; Suzuki, Toshikazu; Okamoto, Eisuke

1994-04-01

The authors developed a real-time personal neutron dosimeter which could give neutron dose equivalent over wide energy region from thermal to 10 odd MeV by using 2 silicon detectors, fast neutron sensor and slow neutron sensor. The energy response of this dosimeter was evaluated under thermal neutron field, monoenergetic neutron field between 200 keV and 15 MeV, and moderated [sup 252]Cf neutron field. The neutron dose equivalent was estimated by adding neutron dose equivalent below 1 MeV given by slow neutron sensor and that above 1 MeV by fast neutron sensor. It was verified from various field tests that this dosimeter is able to give neutron dose equivalent within a factor of 2 margin of accuracy in reactor, accelerator, fusion research and nuclear fuel handling facilities. This dosimeter has more than one order higher sensitivity than conventional personal neutron dosimeters and is insensitive to [gamma]-rays up to about 500 mSv/h. This dosimeter will soon be commercially available as a personal dosimeter which gives neutron and [gamma]-ray dose equivalents simultaneously by installing [gamma]-ray silicon sensor. (author).

Comparative analysis between P1 and B1 equations for neutron moderation

International Nuclear Information System (INIS)

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

2000-01-01

In order to calculate the neutron flux in nuclear reactors, B1 or P1 equations are solved by numerical methods for several groups of energy. The neutron fluxes obtained from the solutions of the B1 and P1 equations are similar when they are applied to large nuclear power reactors. However, an important difference between the two fluxes is that the system of P1 equations uses one more approximation than the B1 system and then, its flux is less precise. The present work shows the relations between both equations and analyzes for what conditions the two equations systems are equivalent. Furthermore, this equations are numerically solved in 54 groups of energy for a quadrangular arrange. (author)

Energy-resolved fast neutron resonance radiography at CSNS

Science.gov (United States)

Tan, Zhixin; Tang, Jingyu; Jing, Hantao; Fan, Ruirui; Li, Qiang; Ning, Changjun; Bao, Jie; Ruan, Xichao; Luan, Guangyuan; Feng, Changqin; Zhang, Xianpeng

2018-05-01

The white neutron beamline at the China Spallation Neutron Source will be used mainly for nuclear data measurements. It will be characterized by high flux and broad energy spectra. To exploit the beamline as a neutron imaging source, we propose a liquid scintillator fiber array for fast neutron resonance radiography. The fiber detector unit has a small exposed area, which will limit the event counts and separate the events in time, thus satisfying the requirements for single-event time-of-flight (SEToF) measurement. The current study addresses the physical design criteria for ToF measurement, including flux estimation and detector response. Future development and potential application of the technology are also discussed.

Single event upset and charge collection measurements using high energy protons and neutrons

International Nuclear Information System (INIS)

Normand, E.; Oberg, D.L.; Wert, J.L.; Ness, J.D.; Majewski, P.P.; Wender, S.; Gavron, A.

1994-01-01

RAMs, microcontrollers and surface barrier detectors were exposed to beams of high energy protons and neutrons to measure the induced number of upsets as well as energy deposition. The WNR facility at Los Alamos provided a neutron spectrum similar to that of the atmospheric neutrons. Its effect on devices was compared to that of protons with energies of 200, 400, 500, and 800 MeV. Measurements indicate that SEU cross sections for 400 MeV protons are similar to those induced by the atmospheric neutron spectrum

Response function measurement of plastic scintillator for high energy neutrons

International Nuclear Information System (INIS)

Sanami, Toshiya; Ban, Syuichi; Takahashi, Kazutoshi; Takada, Masashi

2003-01-01

The response function and detection efficiency of 2''φ x 2''L plastic (PilotU) and NE213 liquid (2''NE213) scintillators, which were used for the measurement of secondary neutrons from high energy electron induced reactions, were measured at Heavy Ion Medical Accelerator in Chiba (HIMAC). High energy neutrons were produced via 400 MeV/n C beam bombardment on a thick graphite target. The detectors were placed at 15 deg with respect to C beam axis, 5 m away from the target. As standard, a 5''φ x 5''L NE213 liquid scintillator (5''NE213) was also placed at same position. Neutron energy was determined by the time-of-flight method with the beam pickup scintillator in front of the target. In front of the detectors, veto scintillators were placed to remove charged particle events. All detector signals were corrected with list mode event by event. We deduce neutron spectrum for each detectors. The efficiency curves for pilotU and 2''NE213 were determined on the bases of 5 N E213 neutron spectrum and its efficiency calculated by CECIL code. (author)

CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident

International Nuclear Information System (INIS)

Bessis, J.

1976-01-01

Nature of physical problem solved: Calculation of flux and neutron spectra in the case of a criticality accident. The method is unsophisticated but fast. The program is divided into two parts: (1) The code CRITIC is based on the Fermi age equation and evaluates the neutron number per fission emitted from a moderate critical system and its energy spectrum. (2) The code NARCISSE uses concrete current albedo, evaluates the product of neutron reflection on walls of the source containment and calculates the resulting flux at any point, and its energy distribution into 21 groups. The results obtained seem satisfactory, if compared with a Monte Carlo program

Reevaluation of the case, de Hoffman, and Placzek one-group neutron transport benchmark solution in plane geometry

International Nuclear Information System (INIS)

Ganapol, B.D.

1986-01-01

In a course on neutron transport theory and also in the analytical neutron transport theory literature, the pioneering work of Case et al. (CdHP) is often referenced. This work was truly a monumental effort in that it treated the fundamental mathematical properties of the one-group neutron Boltzmann equation in detail as well as the numerical evaluation of most of the resulting solutions. Many mathematically and numerically oriented dissertations were based on this classic monograph. In light of the considerable advances made both in numerical methods and computer technology since 1953, when the historic CdHP monograph first appeared, it seems appropriate to reevaluate the numerical benchmark solutions found therein with present-day computational technology. In most transport theory courses, the subject of proper benchmarking of numerical algorithms and transport codes is seldom addressed at any great length. This may be the reason that the benchmarking procedure is so rarely practiced in the nuclear community and when practiced is improperly applied. In this presentation, the development of a new benchmark for the one-group neutron flux in an infinite medium will be detailed with emphasis placed on the educational aspects of the benchmarking activity

Neutron yield from thick lead target by the action of high-energy electrons

International Nuclear Information System (INIS)

Noga, V.I.; Ranyuk, Yu.N.; Telegin, Yu.N.; Sorokin, P.V.

1978-01-01

The results are presented of studying the complete neutron yield from a lead target bombarded by high-energy electrons. Neutrons were recorded by the method of radio-active indicators. The dependence of the neutron yield on the target thickness varying from 0.2 to 8 cm was obtained at the energies of electrons of 230 and 1200 MeV. The neutron yield for the given energies with the target of 6 cm in thickness is in the range of saturation and is 0.1 +-0.03 and 0.65+-0.22 (neutr./MeV.el.), respectively. The neutron angular distributions were measured for different thicknesses of targets at the 201, 230 and 1200 MeV electrons. Within the error limits the angular distributions are isotropic. The dependence of neutron yield on the electron energy was examined for a 3 cm thick target. In the energy range of 100-1200 MeV these values are related by a linear dependence with the proportionality coefficient C=3x10 -4 (neutr./MeV.el.)

Energy behaviour of neutrons generated by Witch-type distributed axi-symmetrical deuteron beams accelerated onto plane tritium targets

International Nuclear Information System (INIS)

Timus, D.M.; Bradley, D.A.; Timus, B.D.; Kalla, S.L.; Srivastava, H.M.

2000-01-01

This paper is an analytical study of the spatial dependency of the d-T neutron energy in the vicinity of a homogeneous tritium-occluded plane target. Close to the target, and along the path of incidence of axially symmetric deuteron beams, the transverse density of accelerated deuterons is assumed to be governed by a law approximated by the 'Witch' function. In particular circumstances, the elementary neutron emission process in non-dispersive media can be considered to be omni-directional (due consideration being paid to collision kinetics, depending upon mass and kinetic energy of particles involved in the nuclear collision, nuclear reaction energy, etc.). Consequently, analytical expressions can be considerably simplified. By applying the classical kinetic energy and momentum conservation laws to nuclear processes, a theoretical description is obtained, taking into account the exoergic character of d-T fusion reaction. A number of expressions for energetic prediction of the fast neutron field are proposed. The associated relations, involving elementary functions, can be investigated using a desk-top computer. Computationally tractable tools are of importance in the study of diverse situations such as induced reactions and activation analysis using 14 MeV neutron generators, investigations in health-physics, radiation dose measurements, nuclear medicine, damage effects, and simulation studies

Genetic Algorithms: A New Method for Neutron Beam Spectral Characterization

International Nuclear Information System (INIS)

David W. Freeman

2000-01-01

A revolutionary new concept for solving the neutron spectrum unfolding problem using genetic algorithms (GAs) has recently been introduced. GAs are part of a new field of evolutionary solution techniques that mimic living systems with computer-simulated chromosome solutions that mate, mutate, and evolve to create improved solutions. The original motivation for the research was to improve spectral characterization of neutron beams associated with boron neutron capture therapy (BNCT). The GA unfolding technique has been successfully applied to problems with moderate energy resolution (up to 47 energy groups). Initial research indicates that the GA unfolding technique may well be superior to popular unfolding methods in common use. Research now under way at Kansas State University is focused on optimizing the unfolding algorithm and expanding its energy resolution to unfold detailed beam spectra based on multiple foil measurements. Indications are that the final code will significantly outperform current, state-of-the-art codes in use by the scientific community

Calculations of nuclear data for the reactions of neutrons and protons with heavy nuclei at energy from 1 MeV up to 2 GeV

International Nuclear Information System (INIS)

Konshin, V.A.

1995-01-01

Several nuclear model codes were applied to calculations of nuclear data in the energy region from 1 MeV to 2 GeV. At energies from 1 to 20 MeV the statistical model code STAPRE was used for calculations of the neutron cross-sections for fission, (n,2n) and (n,3n) reaction cross-sections for 71 actinide isotopes. In the energy region from 10 to 100 MeV the nuclear theory code GNASH was used to calculate the neutron fission and (n,xn) cross-sections for 238 U, 235 U, 239 Pu, 232 Th, 237 Np, 238 Pu, 241 Am, 243 Am, 245 Cm and 246 Cm. At energies from 100 MeV to 2 GeV the intranuclear cascade-exciton model including the fission process was applied to calculations of the interactions of protons and neutrons with actinides and the calculated results are compared with experimental data. (author)

Spatial distribution of moderated neutrons along a Pb target irradiated by high-energy protons

International Nuclear Information System (INIS)

Fragopoulou, M.; Manolopoulou, M.; Stoulos, S.; Brandt, R.; Westmeier, W.; Kulakov, B.A.; Krivopustov, M.I.; Sosnin, A.N.; Debeauvais, M.; Adloff, J.C.; Zamani Valasiadou, M.

2006-01-01

High-energy protons in the range of 0.5-7.4 GeV have irradiated an extended Pb target covered with a paraffin moderator. The moderator was used in order to shift the hard Pb spallation neutron spectrum to lower energies and to increase the transmutation efficiency via (n,γ) reactions. Neutron distributions along and inside the paraffin moderator were measured. An analysis of the experimental results was performed based on particle production by high-energy interactions with heavy targets and neutron spectrum shifting by the paraffin. Conclusions about the spallation neutron production in the target and moderation through the paraffin are presented. The study of the total neutron fluence on the moderator surface as a function of the proton beam energy shows that neutron cost is improved up to 1 GeV. For higher proton beam energies it remains constant with a tendency to decline

First observations of power MOSFET burnout with high energy neutrons

International Nuclear Information System (INIS)

Oberg, D.L.; Wert, J.L.; Normand, E.; Majewski, P.P.; Wender, S.A.

1996-01-01

Single event burnout was seen in power MOSFETs exposed to high energy neutrons. Devices with rated voltage ≥400 volts exhibited burnout at substantially less than the rated voltage. Tests with high energy protons gave similar results. Burnout was also seen in limited tests with lower energy protons and neutrons. Correlations with heavy-ion data are discussed. Accelerator proton data gave favorable comparisons with burnout rates measured on the APEX spacecraft. Implications for burnout at lower altitudes are also discussed

Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

Constraints on the symmetry energy from neutron star observations

International Nuclear Information System (INIS)

Newton, W G; Gearheart, M; Wen, De-Hua; Li, Bao-An

2013-01-01

The modeling of many neutron star observables incorporates the microphysics of both the stellar crust and core, which is tied intimately to the properties of the nuclear matter equation of state (EoS). We explore the predictions of such models over the range of experimentally constrained nuclear matter parameters, focusing on the slope of the symmetry energy at nuclear saturation density L. We use a consistent model of the composition and EoS of neutron star crust and core matter to model the binding energy of pulsar B of the double pulsar system J0737-3039, the frequencies of torsional oscillations of the neutron star crust and the instability region for r-modes in the neutron star core damped by electron-electron viscosity at the crust-core interface. By confronting these models with observations, we illustrate the potential of astrophysical observables to offer constraints on poorly known nuclear matter parameters complementary to terrestrial experiments, and demonstrate that our models consistently predict L < 70 MeV.

A large neutron missing mass spectrometer using long plastic scintillators with electronical determination of the neutron interaction point

International Nuclear Information System (INIS)

Apel, W.D.; Mueller, H.; Schinzel, D.; Sigurdsson, G.; Staudenmaier, H.M.; Stier, U.

1975-11-01

A large acceptance and high efficiency neutron detector is described. The sensitive area and volume of the detector is 3.07 m 2 and 1.03 m 3 respectively. The detector consists of sixteen elements of plastic scintillator, each having the dimensions (16 x 16 x 240) cm 3 . The mean detection efficiency is about 33% for neutrons of 100-500 MeV/c momentum. An interesting feature of this instrument is the accuracy achieved in locating incident particles, which is +-3 cm for charged particles and +-4 cm for neutrons. The accuracy achieved for the time-of-flight measurement is +-1 nsec for fast neutrons. (orig.) [de

The dependence of radiation damage analysis on neutron dosimetry

International Nuclear Information System (INIS)

Goland, A.N.; Parkin, D.M.

1977-01-01

The characteristics of defect production in neutron spectra can be determined by utilizing neutron cross section data (e.g. ENDF/B), detailed neutron spectral data and radiation damage models. The combination of neutron cross section and spectral data is a fundamental starting point in applying damage models. Calculations using these data and damage models show that there are significant differences in the way defects are produced in various neutron spectra. Nonelastic events dominate the recoil energy distribution in high-energy neutron sources such as those based upon fusion and deuteron-breakup reactions. Therefore, high-energy neutron cross sections must be measured or calculated to supplement existing data files. Radiation damage models can then be used to further characterize the diverse neutron spectra

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

A portable and wide energy range semiconductor-based neutron spectrometer

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-11

Hand-held instruments that can be used to passively detect and identify sources of neutron radiation—either bare or obscured by neutron moderating and/or absorbing material(s)—in real time are of interest in a variety of nuclear non-proliferation and health physics applications. Such an instrument must provide a means to high intrinsic detection efficiency and energy-sensitive measurements of free neutron fields, for neutrons ranging from thermal energies to the top end of the evaporation spectrum. To address and overcome the challenges inherent to the aforementioned applications, four solid-state moderating-type neutron spectrometers of varying cost, weight, and complexity have been designed, fabricated, and tested. The motivation of this work is to introduce these novel human-portable instruments by discussing the fundamental theory of their operation, investigating and analyzing the principal considerations for optimal instrument design, and evaluating the capability of each of the four fabricated spectrometers to meet the application needs.

A portable and wide energy range semiconductor-based neutron spectrometer

International Nuclear Information System (INIS)

Hoshor, C.B.; Oakes, T.M.; Myers, E.R.; Rogers, B.J.; Currie, J.E.; Young, S.M.; Crow, J.A.; Scott, P.R.; Miller, W.H.; Bellinger, S.L.; Sobering, T.J.; Fronk, R.G.; Shultis, J.K.; McGregor, D.S.; Caruso, A.N.

2015-01-01

Hand-held instruments that can be used to passively detect and identify sources of neutron radiation—either bare or obscured by neutron moderating and/or absorbing material(s)—in real time are of interest in a variety of nuclear non-proliferation and health physics applications. Such an instrument must provide a means to high intrinsic detection efficiency and energy-sensitive measurements of free neutron fields, for neutrons ranging from thermal energies to the top end of the evaporation spectrum. To address and overcome the challenges inherent to the aforementioned applications, four solid-state moderating-type neutron spectrometers of varying cost, weight, and complexity have been designed, fabricated, and tested. The motivation of this work is to introduce these novel human-portable instruments by discussing the fundamental theory of their operation, investigating and analyzing the principal considerations for optimal instrument design, and evaluating the capability of each of the four fabricated spectrometers to meet the application needs.

Photon and neutron energy response of Thermoluminescent (TL) dosimeters

International Nuclear Information System (INIS)

Thilagam, L.; Priya, M.R.; Mohapatra, D.K.

2018-01-01

Theoretical Monte Carlo (MC) simulations are carried out to investigate the relative thermoluminesence (TL) response of the most commonly used TLD materials to a wide range of photon energy. The effect of polytetrafluoroethylene (PTFE) on TL response of CaSO 4 :Dy is also studied. Additionally, the neutron response of LiF:Mg,Ti TL materials with different concentrations of 6 Li is estimated in terms of the number of 6 Li(n, t) 4 He capture reactions for a wider neutron energy

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Neutron spectroscopy measurements of 14 MeV neutrons at unprecedented energy resolution and implications for deuterium-tritium fusion plasma diagnostics

Science.gov (United States)

Rigamonti, D.; Giacomelli, L.; Gorini, G.; Nocente, M.; Rebai, M.; Tardocchi, M.; Angelone, M.; Batistoni, P.; Cufar, A.; Ghani, Z.; Jednorog, S.; Klix, A.; Laszynska, E.; Loreti, S.; Pillon, M.; Popovichev, S.; Roberts, N.; Thomas, D.; Contributors, JET

2018-04-01

An accurate calibration of the JET neutron diagnostics with a 14 MeV neutron generator was performed in the first half of 2017 in order to provide a reliable measurement of the fusion power during the next JET deuterium-tritium (DT) campaign. In order to meet the target accuracy, the chosen neutron generator has been fully characterized at the Neutron Metrology Laboratory of the National Physical Laboratory (NPL), Teddington, United Kingdom. The present paper describes the measurements of the neutron energy spectra obtained using a high-resolution single-crystal diamond detector (SCD). The measurements, together with a new neutron source routine ‘ad hoc’ developed for the MCNP code, allowed the complex features of the neutron energy spectra resulting from the mixed D/T beam ions interacting with the T/D target nuclei to be resolved for the first time. From the spectral analysis a quantitative estimation of the beam ion composition has been made. The unprecedented intrinsic energy resolution (<1% full width at half maximum (FWHM) at 14 MeV) of diamond detectors opens up new prospects for diagnosing DT plasmas, such as, for instance, the possibility to study non-classical slowing down of the beam ions by neutron spectroscopy on ITER.

Improved fission neutron energy discrimination with {sup 4}He detectors through pulse filtering

Energy Technology Data Exchange (ETDEWEB)

Zhu, Ting, E-mail: ting.zhu@ufl.edu [University of Florida, Gainesville, FL (United States); Liang, Yinong; Rolison, Lucas; Barker, Cathleen; Lewis, Jason; Gokhale, Sasmit [University of Florida, Gainesville, FL (United States); Chandra, Rico [Arktis Radiation Detectors Ltd., Räffelstrasse 11, Zürich (Switzerland); Kiff, Scott [Sandia National Laboratories, CA (United States); Chung, Heejun [Korean Institute for Nuclear Nonproliferation and Control, 1534 Yuseong-daero, Yuseong-gu, Daejeon (Korea, Republic of); Ray, Heather; Baciak, James E.; Enqvist, Andreas; Jordan, Kelly A. [University of Florida, Gainesville, FL (United States)

2017-03-11

This paper presents experimental and computational techniques implemented for {sup 4}He gas scintillation detectors for induced fission neutron detection. Fission neutrons are produced when natural uranium samples are actively interrogated by 2.45 MeV deuterium-deuterium fusion reaction neutrons. Fission neutrons of energies greater than 2.45 MeV can be distinguished by their different scintillation pulse height spectra since {sup 4}He detectors retain incident fast neutron energy information. To enable the preferential detection of fast neutrons up to 10 MeV and suppress low-energy event counts, the detector photomultiplier gain is lowered and trigger threshold is increased. Pile-up and other unreliable events due to the interrogating neutron flux and background radiation are filtered out prior to the evaluation of pulse height spectra. With these problem-specific calibrations and data processing, the {sup 4}He detector's accuracy at discriminating fission neutrons up to 10 MeV is improved and verified with {sup 252}Cf spontaneous fission neutrons. Given the {sup 4}He detector's ability to differentiate fast neutron sources, this proof-of-concept active-interrogation measurement demonstrates the potential of special nuclear materials detection using a {sup 4}He fast neutron detection system.

Dose determination of Neutron contamination in radiothrapy rooms equiped with high energy linear accelerators

International Nuclear Information System (INIS)

Shweikani, R.; Anjak, O.

2014-03-01

Radiotherapy represents the most widely spread technique to control and treat cancer. To increase the treatment efficiency, high-energy linear accelerators are used. However, applying high energy photon beams leads to a non-negligible dose of neutrons contaminating therapeutic beams. A high-energy (23 MV) linear accelerator (Varian 21EX) was studied. The CR-39 nuclear track detectors (NTDs) were used to study the variation of fast neutron relative intensities around a linear accelerator high energy photon beam and to determined the its variation on the patient plane at 0, 50, 100, 150 and 200 cm from the center of the photon beam was. By increasing the distance from the center of the X-ray beam towards the periphery, the photoneutron dose equivalent decreased rapidly for the fields. Photoneutron intensity and distributions at isocenter level with the field sizes of 40*40 cm'2 at SSD=100cm around 23 MV photon beam using Nuclear Track Detectors were determined. The advantages of CR-39 NTD s over active detectors: 1- there is no pulse pileup problem. 2- no photon interference with neutron measurement. 3- no electronics are required. 4 - less prone to noise and interference. The photoneutron intensities were rapidly decreased as we move away from the isocenter of linear accelerators. As the use of simulation software MCNP match in the results we have obtained through direct measurements and the modeling results using the code MCNP (author).

Semiempirical formulas for single-particle energies of neutrons and protons

International Nuclear Information System (INIS)

Lodhi, M.A.K.; Waak, B.T.

1978-01-01

The stepwise multiple linear regression technique has been used to analyze the single-particle energies of neutrons and protons in nuclei along the line of beta stability. Their regular and systematic trends lead to semiempirical model-independent formulas for single-particle energies of neutrons and protons in the bound nuclei as functions of nuclear parameters A and Z for given states specified by nl/sub j/. These formulas are almost as convenient as the harmonic oscillator energy formulas to use. The single-particle energies computed from these formulas have been compared with the experimental data and are found in reasonable agreement

Oxygen enhancement ratio as a function of neutron energy with mammalian cells in culture

International Nuclear Information System (INIS)

Rini, F.J.; Hall, E.J.; Marino, S.A.

1979-01-01

Chinese hamster cells (V79) in culture under oxic and hypoxic conditions were irradiated with several neutron beams spanning a wide energy range to determine the oxygen enhancement ratio (OER). Eight essentially monoenergetic neutron beams, ranging from 0.22 to 13.6 MeV and a 0.11-MeV neutron spectrum, were produced at the Radiological Research Accelerator Facility (RARAF) at Brookhaven National Laboratory. Additional experiments were performed at the Naval Research Laboratory, Washington, DC, where neutrons are produced for radiotherapy by bombarding a beryllium target with 35-MeV deuterons. This beam has a broad energy spectrum with a mean energy of about 15 MeV. A maximum OER of about 1.9 was observed for 13.6-MeV neutrons. The OER values of the monoenergetic neutrons decreased with energy, plateaued at about 1.45 for the energy range from 0.22 to 2.0 MeV and increased slightly to about 1.55 for lower energy spectrum. In the light of microdosimetric data obtained for the neutron beams at RARAF, the OER appears to depend primarily on the intermediate-LET secondaries produced by neutrons in tissue, such as protons, while in contrast high LET-secondaries, such as α-particles and recoil ions, play a minor role. The studies using the NRL neutron beam resulted in a lower OER of about 1.67 as compared to the monoenergetic 13.6-MeV beam. This is a consequence of the fact that more of the dose is deposited by intermediate LET secondaries for the NRL neutron beam

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Analysis of a neutron scattering integral experiment on iron for neutron energies from 1 to 15 MeV

International Nuclear Information System (INIS)

Cramer, S.N.; Oblow, E.M.

1976-11-01

Monte Carlo calculations were made to analyze the results of an integral experiment with an iron sample to determine the adequacy of neutron scattering cross section data for iron. The experimental results analyzed included energy-dependent NE-213 detector count rates at a scattering angle of 90 deg and pulse-height spectra for scattered neutrons produced in an iron ring pulsed with a 1- to 20-MeV neutron source. The pulse-height data were unfolded to generate secondary neutron spectra at 90 deg as a function of incident neutron energy. Multigroup Monte Carlo calculations using the MORSE code and ENDF/B-IV cross sections were made to analyze all reported results. Discrepancies between calculated and measured responses were found for inelastic scattering reactions in the range from 1 to 4 MeV. These results were related to deficiencies in ENDF/B-IV iron cross section data

Neutron quality parameters versus energy below 4 MeV from microdosimetric calculations

International Nuclear Information System (INIS)

Stinchcomb, T.G.; Borak, T.B.

1983-01-01

Charged-particle production by neutrons and the resulting energy-deposition spectra in micron-sized spheres of tissue of varying diameters were calculated from thermal energies to 4 MeV. These data were used to obtain dose-average values of several quality-indicating parameters as functions of neutron energy and of tissue sphere diameter. The contrast among the parameters is shown and discussed. Applications are made to two neutron spectra, one a fission spectrum in air and the other a moderated spectrum at the center of an irradiated cube of water

Methodology of Continuous-Energy Adjoint Monte Carlo for Neutron, Photon, and Coupled Neutron-Photon Transport

International Nuclear Information System (INIS)

Hoogenboom, J. Eduard

2003-01-01

Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems

Some principal problems in physics and low-energy neutron physics

International Nuclear Information System (INIS)

Aleksandrov, Yu.A.

2004-01-01

The questions connected with internal particle (e.g. neutron) structure obtained at low-energy neutron physics are discussed. The first question deals with the charge neutron radius E 2 > 1/2 connected with the value of neutron-electron scattering length a ne determined at low neutron energies. At present, the obtained accuracy allows us to speak not only about the value of E 2 > but also on the segmentation of E 2 > into Dirac and Foldy addenda. The sign of the Dirac addendum is connected directly with the fundamental Yukawa theory explaining the origin of nuclear forces. One of the popular experimental values of the Dirac addendum (from a ne =(-1.32±0.03)·10 -16 cm) contradicts the Yukawa theory. The second question also concerns the subject of the structure of the neutron, namely its deformation. The notion of deformation (polarizability) of the nucleon in electromagnetic field was introduced in the mid-1950s. The reasons are given in favor of the opinion that the neutron polarizability was observed for the first time in neutron experiments as far back as 1957, i.e. earlier than proton polarizability was detected (1960). Finally, the third question deals with the search for a magnetic charge of the neutron. A beautiful experiment (Finkelstein, Shull, Zeilinger, 1986) testifying with high accuracy the absence of a magnetic charge of the neutron is discussed. This diffraction experiment was based on the concept of anomalously small effective mass of the neutron providing greatly enhanced sensitivity. The existence of an isolated magnetic charge in the nature would explain the quantization of electric and magnetic charges (Dirac, 1931)

Design of calibration method in neutron and individual dosimeter

International Nuclear Information System (INIS)

Belkhodia, M.

1984-12-01

Usually albedo dosemeters are calibrated with beam of monoenergetic neutrons. Since neutron energy around neutron sources varies greatly, we applied the calibration method to a mixed field whose energy spectrum lies between 0.025 ev and 10 Mev. The method is based on a mathematical model that deals with the dosimeter response as a function at the neutron energy. The measurements carried out with solid state nuclear track detectors show the dosimeter practical aspect. The albedo dosimeter calibration gave results on good agreement with the international institution recommendations

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-01

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

Status of the low energy neutron source at Indiana University

International Nuclear Information System (INIS)

Baxter, D.V.; Cameron, J.M.; Derenchuk, V.P.; Lavelle, C.M.; Leuschner, M.B.; Lone, M.A.; Meyer, H.O.; Rinckel, T.; Snow, W.M.

2005-01-01

The National Science Foundation has recently approved funding for LENS (the low energy neutron source) at Indiana University and construction of this facility has begun. LENS represents a new paradigm for economically introducing neutron scattering into a university or industrial setting. In this design, neutrons are produced in a long-pulse (1 ms) mode through (p,n) reactions on a water-cooled Be target and the target is tightly coupled to a cryogenic moderator with a water reflector. This design gives a facility suitable for materials research, the development of new neutron instrumentation, and the education of new neutron scientists

Pure Neutron Matter Constraints and Nuclear Symmetry Energy

International Nuclear Information System (INIS)

Fattoyev, F J; Newton, W G; Xu, Jun; Li, Bao-An

2013-01-01

In this review, we will discuss the results of our recent work [1] to study the general optimization of the pure isovector parameters of the popular relativistic mean-field (RMF) and Skyrme-Hartree-Fock (SHF) nuclear energy-density functionals (EDFs), using constraints on the pure neutron matter (PNM) equation of state (EoS) from recent ab initio calculations. By using RMF and SHF parameterizations that give equivalent predictions for ground-state properties of doubly magic nuclei and properties of symmetric nuclear matter (SNM) and PNM, we found that such optimization leads to broadly consistent symmetry energy J and its slope parameter L at saturation density within a tight range of α(J) sym , (b) the symmetry energy at supra-saturation densities, and (c) the radius of neutron stars.

Investigation of the energy correlations of spallation neutrons by the MCNPX code

International Nuclear Information System (INIS)

Szieberth, Mate; Radocz, Gabor

2011-01-01

Earlier works have suggested that the energy correlations in a spallation source may influence the neutron noise measurements in an ADS. For the calculation of this effect not only the generally known and used one-particle spectrum is needed but also the so-called two particle spectrum, which describes also the energy correlations. Since measured data are not available for the energy distribution of the neutrons from a single spallation event the physical models of the MCNPX code have been used to investigate the effect. The calculational model has been successfully validated with measurements of the number distribution of spallation neutrons. The simulated one- and two-particle energy distributions and spectra proved that the energy correlations exist and have an important effect in low multiplicity spallation events and in thin targets. On the other hand for thick targets this effect appears negligible and the factorization of the two-particle spectrum seems an acceptable approximation. Further investigations are in hand to quantify the actual effect of the energy correlations on the neutron noise measurements. (author)

Robust energy storage scheduling for imbalance reduction of strategically formed energy balancing groups

International Nuclear Information System (INIS)

Chakraborty, Shantanu; Okabe, Toshiya

2016-01-01

Imbalance (on-line energy gap between contracted supply and actual demand, and associated cost) reduction is going to be a crucial service for a Power Producer and Supplier (PPS) in the deregulated energy market. PPS requires forward market interactions to procure energy as precisely as possible in order to reduce imbalance energy. This paper presents, 1) (off-line) an effective demand aggregation based strategy for creating a number of balancing groups that leads to higher predictability of group-wise aggregated demand, 2) (on-line) a robust energy storage scheduling that minimizes the imbalance energy and cost of a particular balancing group considering the demand prediction uncertainty. The group formation is performed by a Probabilistic Programming approach using Bayesian Markov Chain Monte Carlo (MCMC) method after applied on the historical demand statistics. Apart from the group formation, the aggregation strategy (with the help of Bayesian Inference) also clears out the upper-limit of the required storage capacity for a formed group, fraction of which is to be utilized in on-line operation. For on-line operation, a robust energy storage scheduling method is proposed that minimizes expected imbalance energy and cost (a non-linear function of imbalance energy) while incorporating the demand uncertainty of a particular group. The proposed methods are applied on the real apartment buildings' demand data in Tokyo, Japan. Simulation results are presented to verify the effectiveness of the proposed methods. - Highlights: • Strategic method for intelligent energy balancing group formation using Bayesian MCMC. • Stochastic programming based robust and online energy storage (battery) scheduling. • Imbalance cost (regulation) and energy reduction of a balancing group. • Imbalance cost reduction of 80% attainable by considerably lower battery capacity.

ENERGY RESPONSE OF FLUORESCENT NUCLEAR TRACK DETECTORS OF VARIOUS COLORATIONS TO MONOENERGETIC NEUTRONS.

Science.gov (United States)

Fomenko, V; Moreno, B; Million, M; Harrison, J; Akselrod, M

2017-10-25

The neutron-energy dependence of the track-counting sensitivity of fluorescent nuclear track detectors (FNTDs) at two ranges of Mg doping, resulting in different crystal colorations, was investigated. The performance of FNTDs was studied with the following converters: Li-glass for thermal to intermediate-energy neutrons, polyethylene for fast neutrons, and polytetrafluoroethylene (Teflon™) for photon- and radon-background subtraction. The irradiations with monoenergetic neutrons were performed at the National Physics Laboratory (NPL), UK. The energy range was varied from 144 keV to 16.5 MeV in the personal dose equivalent range from 1 to 3 mSv. Monte Carlo simulations were performed to model the response of FNTDs to monoenergetic neutrons. A good agreement with the experimental data was observed suggesting the development of a basic model for future MC studies. Further work will focus on increasing FNTD sensitivity to low-energy neutrons and developing a faster imaging technique for scanning larger areas to improve counting statistics. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

The neutron dose equivalent around high energy medical electron linear accelerators

Directory of Open Access Journals (Sweden)

Poje Marina

2014-01-01

Full Text Available The measurement of neutron dose equivalent was made in four dual energy linear accelerator rooms. Two of the rooms were reconstructed after decommissioning of 60Co units, so the main limitation was the space. The measurements were performed by a nuclear track etched detectors LR-115 associated with the converter (radiator that consist of 10B and with the active neutron detector Thermo BIOREM FHT 742. The detectors were set at several locations to evaluate the neutron ambient dose equivalent and/or neutron dose rate to which medical personnel could be exposed. Also, the neutron dose dependence on collimator aperture was analyzed. The obtained neutron dose rates outside the accelerator rooms were several times smaller than the neutron dose rates inside the accelerator rooms. Nevertheless, the measured neutron dose equivalent was not negligible from the aspect of the personal dosimetry with almost 2 mSv a year per person in the areas occupied by staff (conservative estimation. In rooms with 15 MV accelerators, the neutron exposure to the personnel was significantly lower than in the rooms having 18 MV accelerators installed. It was even more pronounced in the room reconstructed after the 60Co decommissioning. This study confirms that shielding from the neutron radiation should be considered when building vaults for high energy linear accelerators, especially when the space constraints exist.

Unique furnace system for high-energy-neutron experiments

International Nuclear Information System (INIS)

Panayotou, N.F.; Green, D.R.; Price, L.S.

1982-03-01

The low flux of high energy neutron sources requires optimum utilization of the available neutron field. A furnace system has been developed in support of the US DOE fusion materials program which meets this challenge. Specimens positioned in two temperature zones just 1 mm away from the outside surface of a neutron window in the furnace enclosure can be irradiated simultaneously at two independent, isothermal (+- 1 0 C) temperatures. The temperature difference between these closely spaced isothermal zones is controllable from 0 to 320 0 C and the maximum temperature is 400 0 C. The design of the system also provides a controlled specimen environment, rapid heating and cooling and easy access to heaters and thermocouples. This furnace system is in use at the Rotating Target Neutron Source-II of Lawrence Livermore National Laboratory

Measurement of prompt neutron spectra from the "2"3"9Pu(n, f ) fission reaction for incident neutron energies from 1 to 200 MeV

International Nuclear Information System (INIS)

Chatillon, A.; Belier, G.; Granier, T.; Laurent, B.; Morillon, B.; Taieb, J.; Haight, R.C.; Devlin, M.; Nelson, R.O.; Noda, R.S.; O'Donnell, J.M.

2014-01-01

Prompt fission neutron spectra in the neutron-induced fission of "2"3"9Pu have been measured for incident neutron energies from 1 to 200 MeV at the Los Alamos Neutron Science Center. Mean energies deduced from the prompt fission neutron spectra (PFNS) lead to the observation of the opening of the second chance fission at 7 MeV and to indications for the openings of fission channels of third and fourth chances. Moreover, the general trend of the measured PFNS is well reproduced by the different models. The comparison between data and models presents, however, two discrepancies. First, the prompt neutron mean energy seems constant for neutron energy, at least up to 7 MeV, whereas in the theoretical calculations it is continuously increasing. Second, data disagree with models on the shape of the high energy part of the PFNS, where our data suggest a softer spectrum than the predictions. (authors)

Supersoft Symmetry Energy Encountering Non-Newtonian Gravity in Neutron Stars

International Nuclear Information System (INIS)

Wen Dehua; Li Baoan; Chen Liewen

2009-01-01

Considering the non-Newtonian gravity proposed in grand unification theories, we show that the stability and observed global properties of neutron stars cannot rule out the supersoft nuclear symmetry energies at suprasaturation densities. The degree of possible violation of the inverse-square law of gravity in neutron stars is estimated using an equation of state of neutron-rich nuclear matter consistent with the available terrestrial laboratory data.

Development of a TPC for energy and fluence references in low energies neutronic fields (from 8 keV to 5 MeV)

International Nuclear Information System (INIS)

Maire, Donovan

2015-01-01

In order to judge the measurement reliability, metrology requires to measure quantities with their uncertainties, in relation to a reference through a documented and unbroken chain of calibrations. In neutron radiation field, instrument response has to be known as a function of the neutron energy. Then detector calibrations are required using reference neutron fields. In France, primary reference neutron fields are held by the LNE-IRSN, at the Laboratory for Neutron Metrology and Dosimetry (LMDN). In order to improve reference neutron field characterization, the LNE-IRSN MIMAC μTPC has been developed. This detector is a Time Projection Chamber (TPC), using a gas at low pressure (30 mbar abs. to 1 bar abs.). Nuclear recoils are generated by neutron elastic scattering onto gas atoms. By measuring the nuclear recoil energy and scattering angle, the μTPC detector is able to measure the energy distribution of the neutron fluence between 8 keV and 5 MeV. The main challenge was to perform accurate spectrometry of neutron fields in the keV range, following a primary procedure. First of all, a metrological approach was followed in order to master every physical process taking part in the neutron detection. This approach led to develop the direct and inverse models, representing the detector response function and its inverse function respectively. Using this detailed characterization, the energy distribution of the neutron fluence has been measured for a continuous neutron field of 27 keV. The reconstructed energy is 28,2 ± 4,5 keV, the difference between μTPC integral fluence measurement and other measurement methods is less than 6%. The LNE-IRSN MIMAC μTPC system becomes the only one system able to measure simultaneously energy and fluence at energies lower than 100 keV, following a primary procedure. The project goal is then reached. These measurements at energies lower than 100 keV shows also a non-linearity between the ionization charge and the ion kinetic energy

Delayed neutrons as a probe of nuclear charge distribution in fission of heavy nuclei by neutrons

International Nuclear Information System (INIS)

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

2002-01-01

A method of the determination of cumulative yields of delayed neutron precursors is developed. This method is based on the iterative least-square procedure applied to delayed neutron decay curves measured after irradiation of 235 U sample by thermal neutrons. Obtained cumulative yields in turns were used for deriving the values of the most probable charge in low-energy fission of the above-mentioned nucleus. (author)

Spectral correction factors for conventional neutron dose meters used in high-energy neutron environments improved and extended results based on a complete survey of all neutron spectra in IAEA-TRS-403

International Nuclear Information System (INIS)

Oparaji, U.; Tsai, Y. H.; Liu, Y. C.; Lee, K. W.; Patelli, E.; Sheu, R. J.

2017-01-01

This paper presents improved and extended results of our previous study on corrections for conventional neutron dose meters used in environments with high-energy neutrons (E n > 10 MeV). Conventional moderated-type neutron dose meters tend to underestimate the dose contribution of high-energy neutrons because of the opposite trends of dose conversion coefficients and detection efficiencies as the neutron energy increases. A practical correction scheme was proposed based on analysis of hundreds of neutron spectra in the IAEA-TRS-403 report. By comparing 252 Cf-calibrated dose responses with reference values derived from fluence-to-dose conversion coefficients, this study provides recommendations for neutron field characterization and the corresponding dose correction factors. Further sensitivity studies confirm the appropriateness of the proposed scheme and indicate that (1) the spectral correction factors are nearly independent of the selection of three commonly used calibration sources: 252 Cf, 241 Am-Be and 239 Pu-Be; (2) the derived correction factors for Bonner spheres of various sizes (6''-9'') are similar in trend and (3) practical high-energy neutron indexes based on measurements can be established to facilitate the application of these correction factors in workplaces. (authors)

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

International Nuclear Information System (INIS)

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

1987-02-01

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

A solution for neutron personal dosimetry in the absence of workplace spectrometry

International Nuclear Information System (INIS)

Hajek, M.; Cruz Suarez, R.

2016-01-01

In view of the widely varying energy spectra encountered in practical situations, accuracy of neutron dose assessment requires detailed knowledge of detector responses and workplace conditions to achieve an adequate level of protection. If the neutron spectrum should be a priori unknown and no measurement of the workplace spectrum is available, the 'Compendium of Neutron Spectra and Detector Responses for Radiation Protection Purposes' published in the International Atomic Energy Agency Technical Report Series offers a broad range of reference spectra that may be appropriate for many applications. The proposed approach applies a correction factor based on the ratio of 'personal dose equivalent indices' for a particular workplace spectrum and a reference field used for calibration of the dosemeter response. Amendments in the definition of operational quantities as well as introduction of new modalities that, for example, may be expected to give increased importance to high-energy neutrons necessitate frequent revision of the Compendium. Results from the European Radiation Dosimetry Group Intercomparison 2012 for neutron personal dosemeters provide evidence that workplace fields are insufficiently reflected. This is proposed to be considered as an improvement opportunity. (authors)

Neutron energy spectra produced by α-bombardment of light elements in thick targets

International Nuclear Information System (INIS)

Jacobs, G.J.H.

1982-01-01

The aim of the work, presented in this thesis, is to determine energy spectra of neutrons produced by α-particle bombardment of thick targets containing light elements. These spectra are required for nuclear waste management. The set-up of the neutron spectrometer is described, and its calibration discussed. Absolute efficiencies were determined at various neutron energies, using monoenergetic neutrons produced with the Van de Graaff accelerator in pulsed mode. The additional calibration of the neutron spectrometer as proton-recoil spectrometer was carried out primarily for future applications in measurements where no pulsed neutron source is available or the neutron flux density is too low. The basis for an accurate uncertainty analysis is made by the determination of the covariance matrix for the uncertainties in the efficiencies. The determination of the neutron energy spectra from time-of-flight and from proton-recoil measurements is described. A comparison of the results obtained from the two different types of measurements is made. The experimentally determined spectra were compared with spectra calculated from stopping powers and theoretically determined cross sections. These cross sections were calculated from optical model parameters and level parameters using the Hauser-Feshbach formalism. Measurements were carried out on thick targets of silicon, aluminium, magnesium, carbon, boron nitride, calcium fluoride, aluminium oxide, silicon oxide and uranium oxide at four different α-particle energies. (Auth.)

New Beta-delayed Neutron Measurements in the Light-mass Fission Group

Energy Technology Data Exchange (ETDEWEB)

Agramunt, J. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); García, A.R. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Algora, A. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Äystö, J. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Caballero-Folch, R.; Calviño, F. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Cano-Ott, D. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Cortés, G. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Domingo-Pardo, C. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Eronen, T. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Gelletly, W. [Department of Physics, University of Surrey, Guildford GU2 7XH (United Kingdom); Gómez-Hornillos, M.B. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); and others

2014-06-15

A new accurate determination of beta-delayed neutron emission probabilities from nuclei in the low mass region of the light fission group has been performed. The measurements were carried out using the BELEN 4π neutron counter at the IGISOL-JYFL mass separator in combination with a Penning trap. The new results significantly improve the uncertainties of neutron emission probabilities for {sup 91}Br, {sup 86}As, {sup 85}As, and {sup 85}Ge nuclei.

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

The total kinetic energy release in the fast neutron-induced fission of {sup 232}Th

Energy Technology Data Exchange (ETDEWEB)

King, Jonathan; Yanez, Ricardo; Loveland, Walter; Barrett, J. Spencer; Oscar, Breland [Oregon State University, Dept. of Chemistry, Corvallis, OR (United States); Fotiades, Nikolaos; Tovesson, Fredrik; Young Lee, Hye [Los Alamos National Laboratory, Physics Division, Los Alamos, NM (United States)

2017-12-15

The post-emission total kinetic energy release (TKE) in the neutron-induced fission of {sup 232}Th was measured (using white spectrum neutrons from LANSCE) for neutron energies from E{sub n} = 3 to 91 MeV. In this energy range the average post-neutron total kinetic energy release decreases from 162.3 ± 0.3 at E{sub n} = 3 MeV to 154.9 ± 0.3 MeV at E{sub n} = 91 MeV. Analysis of the fission mass distributions indicates that the decrease in TKE with increasing neutron energy is a combination of increasing yields of symmetric fission (which has a lower associated TKE) and a decrease in the TKE release in asymmetric fission. (orig.)

Measurements of time dependent energy spectra of neutrons in a small graphite assembly

International Nuclear Information System (INIS)

Fujita, Yoshiaki; Sakamoto, Shigeyasu; Aizawa, Otohiko; Takahashi, Akito; Sumita, Kenji.

1975-01-01

The time-dependent energy spectra of neutrons have been measured in a small 30x30x30 cm 3 graphite assembly by means of the linac-chopper method, with a view to establishing experimental evidence that there is no asymptotic spectrum in such a small assembly, and in order to study the non-asymptotic behavior of neutrons. The arrangement of a polyethylene pre-moderator adjacent to the assembly made the measurements possible with the improvement obtained thereby of the neutron counting statistics. It was indicated from calculation that the presence of the pre-moderator had little effect - at least above the Bragg cut-off energy - on the evolution in time of the energy spectra of neutrons in the graphite assembly. The experimental results indicated very probable disappearance of asymptotic spectra, and revealed significant enhancement of trapping at Bragg energies with the lapse of time. This is consistent with the results of pulsed neutron experiments in small assemblies conducted by Takahashi et al., and falls in line with de Saussure's approximation. The spectra in the graphite assembly showed significant space dependence, the spectra becoming harder with increasing distance from the pre-moderator. This hardening may be attributed to the relatively faster propagation of higher energy neutrons. (auth.)

PANDORA, a large volume low-energy neutron detector with real-time neutron-gamma discrimination

Science.gov (United States)

Stuhl, L.; Sasano, M.; Yako, K.; Yasuda, J.; Baba, H.; Ota, S.; Uesaka, T.

2017-09-01

The PANDORA (Particle Analyzer Neutron Detector Of Real-time Acquisition) system, which was developed for use in inverse kinematics experiments with unstable isotope beams, is a neutron detector based on a plastic scintillator coupled to a digital readout. PANDORA can be used for any reaction study involving the emission of low energy neutrons (100 keV-10 MeV) where background suppression and an increased signal-to-noise ratio are crucial. The digital readout system provides an opportunity for pulse shape discrimination (PSD) of the detected particles as well as intelligent triggering based on PSD. The figure of merit results of PANDORA are compared to the data in literature. Using PANDORA, 91 ± 1% of all detected neutrons can be separated, while 91 ± 1% of the detected gamma rays can be excluded, reducing the gamma ray background by one order of magnitude.

Neutron dose and energy spectra measurements at Savannah River Plant

International Nuclear Information System (INIS)

Brackenbush, L.W.; Soldat, K.L.; Haggard, D.L.; Faust, L.G.; Tomeraasen, P.L.

1987-08-01

Because some workers have a high potential for significant neutron exposure, the Savannah River Plant (SRP) contracted with Pacific Northwest Laboratory (PNL) to verify the accuracy of neutron dosimetry at the plant. Energy spectrum and neutron dose measurements were made at the SRP calibrations laboratory and at several other locations. The energy spectra measurements were made using multisphere or Bonner sphere spectrometers, 3 He spectrometers, and NE-213 liquid scintillator spectrometers. Neutron dose equivalent determinations were made using these instruments and others specifically designed to determine dose equivalent, such as the tissue equivalent proportional counter (TEPC). Survey instruments, such as the Eberline PNR-4, and the thermoluminescent dosimeter (TLD)-albedo and track etch dosimeters (TEDs) were also used. The TEPC, subjectively judged to provide the most accurate estimation of true dose equivalent, was used as the reference for comparison with other devices. 29 refs., 43 figs., 13 tabs

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Systematic studies of binding energy dependence of neutron-proton momentum correlation function

International Nuclear Information System (INIS)

Wei, Y B; Ma, Y G; Shen, W Q; Ma, G L; Wang, K; Cai, X Z; Zhong, C; Guo, W; Chen, J G; Fang, D Q; Tian, W D; Zhou, X F

2004-01-01

Hanbury Brown-Twiss (HBT) results of the neutron-proton correlation function have been systematically investigated for a series of nuclear reactions with light projectiles with the help of the isospin-dependent quantum molecular dynamics model. The relationship between the binding energy per nucleon of the projectiles and the strength of the neutron-proton HBT at small relative momentum has been obtained. Results show that neutron-proton HBT results are sensitive to the binding energy per nucleon

Comparative analysis between P1 and B1 equations for neutron moderation; Analise comparativa entre os metodos de obtencao e das solucoes das equacoes P1 e B1 para moderacao de neutrons

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

In order to calculate the neutron flux in nuclear reactors, B1 or P1 equations are solved by numerical methods for several groups of energy. The neutron fluxes obtained from the solutions of the B1 and P1 equations are similar when they are applied to large nuclear power reactors. However, an important difference between the two fluxes is that the system of P1 equations uses one more approximation than the B1 system and then, its flux is less precise. The present work shows the relations between both equations and analyzes for what conditions the two equations systems are equivalent. Furthermore, this equations are numerically solved in 54 groups of energy for a quadrangular arrange. (author)

Inelastic neutron scattering study of methyl groups rotation in some methylxanthines

Science.gov (United States)

Prager, M.; Pawlukojc, A.; Wischnewski, A.; Wuttke, J.

2007-12-01

The three isomeric dimethylxanthines and trimethylxanthine are studied by neutron spectroscopy up to energy transfers of 100meV at energy resolutions ranging from 0.7μeV to some meV. The loss of elastic intensity with increasing temperature can be modeled by quasielastic methyl rotation. The number of inequivalent methyl groups is in agreement with those of the room temperature crystal structures. Activation energies are obtained. In the case of theophylline, a doublet tunneling band is observed at 15.1 and 17.5μeV. In theobromine, a single tunneling band at 0.3μeV is found. Orientational disorder in caffeine leads to a 2.7μeV broad distribution of tunneling bands around the elastic line. At the same time, broad low energy phonon spectra characterize an orientational glassy state with weak methyl rotational potentials. Librational energies of the dimethylxanthines are clearly seen in the phonon densities of states. Rotational potentials can be derived which explain consistently all observables. While their symmetry in general is threefold, theophylline shows a close to sixfold potential reflecting a mirror symmetry.

Neutron radiography with ultracold neutrons

International Nuclear Information System (INIS)

Bates, J.C.

1981-01-01

The neutron transmission factor of very thin films may be low if the neutron energy is comparable to the pseudo-potential of the film material. Surprisingly, perhaps, it is relatively easy to obtain neutrons with such low energies in sufficient numbers to produce neutron radiographs. (orig.)

Neutron-triton scattering lengths for interactions reproducing low-energy trinucleon data

International Nuclear Information System (INIS)

Levashev, V.P.

1981-01-01

By solving the integral equations for four nucleons the neutron-triton scattering lengths and total cross section are calculated using different S-wave rank-one separable potentials. A number of linear correlations between the neutron-triton scattering lengths and triton binding energy are found. The scattering lengths consistent with low-energy trinucleon data. The results obtained are compared with available experimental data [ru

Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

1986-01-01

For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

Neutron Dose Measurement Using a Cubic Moderator

International Nuclear Information System (INIS)

Sheinfeld, M.; Mazor, T.; Cohen, Y.; Kadmon, Y.; Orion, I.

2014-01-01

The Bonner Sphere Spectrometer (BSS), introduced In July 1960 by a research group from Rice University, Texas, is a major approach to neutron spectrum estimation. The BSS, also known as multi-sphere spectrometer, consists of a set of a different diameters polyethylene spheres, carrying a small LiI(Eu) scintillator in their center. What makes this spectrometry method such widely used, is its almost isotropic response, covering an extraordinary wide range of energies, from thermal up to even hundreds of MeVs. One of the most interesting and useful consequences of the above study is the 12'' sphere characteristics, as it turned out that the response curve of its energy dependence, have a similar shape compared with the neutron's dose equivalent as a function of energy. This inexplicable and happy circumstance makes it virtually the only monitoring device capable providing realistic neutron dose estimates over such a wide energy range. However, since the detection mechanism is not strictly related to radiation dose, one can expect substantial errors when applied to widely different source conditions. Although the original design of the BSS included a small 4mmx4mmO 6LiI(Eu) scintillator, other thermal neutron detectors has been used over the years: track detectors, activation foils, BF3 filled proportional counters, etc. In this study we chose a Boron loaded scintillator, EJ-254, as the thermal neutron detector. The neutron capture reaction on the boron has a Q value of 2.78 MeV of which 2.34 MeV is shared by the alpha and lithium particles. The high manufacturing costs, the encasement issue, the installation efficiency and the fabrication complexity, led us to the idea of replacing the sphere with a cubic moderator. This article describes the considerations, as well as the Monte-Carlo simulations done in order to examine the applicability of this idea

Calculating disadvantage factor for fuel taking into account the neutron energy distribution; Odredjivanje 'disadvantage' faktora goriva sa uracunavanjem energetske raspodele neutrona

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-07-01

Errors in calculating the disadvantage factor are caused by applying the diffusion approximation and one-group method. This paper describes the method for calculating the fuel disadvantage factor by applying a non-diffusion method taking into account neutron thermalization.

Neutron scattering investigation of magnetic excitations at high energy transfers

International Nuclear Information System (INIS)

Loong, C.K.

1984-01-01

With the advance of pulsed spallation neutron sources, neutron scattering investigation of elementary excitations in magnetic materials can now be extended to energies up to several hundreds of MeV. We have measured, using chopper spectrometers and time-of-flight techniques, the magnetic response functions of a series of d and f transition metals and compounds over a wide range of energy and momentum transfer. In PrO 2 , UO 2 , BaPrO 3 and CeB 6 we observed crystal-field transitions between the magnetic ground state and the excited levels in the energy range from 40 to 260 MeV. In materials exhibiting spin-fluctuation or mixed-valent character such as Ce 74 Th 26 , on the other hand, no sharp crystal-field lines but a broadened quasielastic magnetic peak was observed. The line width of the quasielastic component is thought to be connected to the spin-fluctuation energy of the 4f electrons. The significance of the neutron scattering results in relation to the ground state level structure of the magnetic ions and the spin-dynamics of the f electrons is discussed. Recently, in a study of the spin-wave excitations in itinerant magnetic systems, we have extended the spin-wave measurements in ferromagnetic iron up to about 160 MeV. Neutron scattering data at high energy transfers are of particular interest because they provide direct comparison with recent theories of itinerant magnetism. 26 references, 7 figures

Knowledge Management in the Neutronics Group of CAREM Project

International Nuclear Information System (INIS)

Torres, L.; Lopasso, E.

2016-01-01

Full text: An analysis of the Neutronics Group of CAREM25 project was performed in order to plan for the gradual implementation of knowledge management. The group structure, performed tasks and the way these tasks are linked together were studied. Staff functions within the group, profiles of each position and the training and education of human resources were also analyzed. (author

A parameter study to determine the optimal source neutron energy in boron neutron capture therapy of brain tumours

Energy Technology Data Exchange (ETDEWEB)

Nievaart, V A [Reactor Physics Department, Delft University of Technology, Mekelweg 15, 2629JB Delft (Netherlands); Moss, R L [Joint Research Centre of the European Commission, Postbus 2, 1755ZG Petten (Netherlands); Kloosterman, J L [Reactor Physics Department, Delft University of Technology, Mekelweg 15, 2629JB Delft (Netherlands); Hagen, T H J J van der [Reactor Physics Department, Delft University of Technology, Mekelweg 15, 2629JB Delft (Netherlands); Dam, H van [Reactor Physics Department, Delft University of Technology, Mekelweg 15, 2629JB Delft (Netherlands)

2004-09-21

The values of the parameters used in boron neutron capture therapy (BNCT) to calculate a given dose to human tissue vary with patients due to different physical, biological and/or medical circumstances. Parameters include the tissue dimensions, the {sup 10}B concentration and the relative biological effectiveness (RBE) factors for the different dose components associated with BNCT. Because there is still no worldwide agreement on RBE values, more often than not, average values for these parameters are used. It turns out that the RBE-problem can be circumvented by taking into account all imaginable parameter values. Approaching this quest from another angle: the outcome will also provide the parameters (and values) which influence the optimal source neutron energy. For brain tumours it turns out that the {sup 10}B concentration, the RBE factors for {sup 10}B as well as fast neutrons, together with the dose limit set for healthy tissue, affect the optimal BNCT source neutron energy. By using source neutrons of a few keV together with neutrons of a few eV, it ensures that, under all imaginable circumstances, a maximum of alpha (and lithium) particles can be delivered in the tumour.

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

Science.gov (United States)

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

2015-09-11

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

Design and Fabrication of Titanium Target for Portable Neutron Generator

International Nuclear Information System (INIS)

Lee, Cheol Ho; Oh, Byunghoon; Chang, Daesik; Jang, Dohyun; In Sang Yeol; Park, Jaewon; Hong, Kwangpyo

2014-01-01

For the neutron generator to produce a neutron flux of the above order, a target that produces fast neutrons in the generator plays an important role, and the target is used and applied to develop the generator due to its simplicity and inexpensive. Making suitable targets for neutron production, especially mono-energy neutrons, has always been of interest. These targets have been used for neutron production reaction studies, calibration of detectors, and neutron therapy. Different studies have been carried out on deuterium and tritium for making solid targets to produce mono-energy neutron from D-D and D-T reactions. A lot of investigations have been carried out on solid target properties such as lifetime, thermal stability, neutron yield, and energy. Vaporized zirconium and titanium layers on a high thermal conductivity substrate (Cu, Mo, Ag) have been used as deuterium and tritium absorbing metals. The density of titanium is smaller than zirconium and the range of charged particles in the titanium targets is more than that in zirconium targets. Thus, titanium targets have more neutron yield than zirconium targets in a low energy beam and titanium is usually used to make a target. The titanium target was designed and simulated to determine the suitable thickness of the target. As a result of the simulation, the target was fabricated to generate fast neutrons by the reaction. The thickness of the target was measured using a profiler. The thickness of the two targets is 2.108 and 2.190 μm. The target will be applied to produce neutrons in a neutron generator

Delayed neutrons as a probe of nuclear charge distribution in fission of heavy nuclei by neutrons

CERN Document Server

Isaev, S G; Piksaikin, V M; Roshchenko, V A

2001-01-01

A method of the determination of cumulative yields of delayed neutron precursors is developed. This method is based on the iterative least-square procedure applied to delayed neutron decay curves measured after irradiation of sup 2 sup 3 sup 5 U sample by thermal neutrons. Obtained cumulative yields in turns were used for deriving the values of the most probable charge in low-energy fission of the above-mentioned nucleus.

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)

Measurements of Relative Biological Effectiveness and Oxygen Enhancement Ratio of Fast Neutrons of Different Energies

Energy Technology Data Exchange (ETDEWEB)

Barendsen, G. W.; Broerse, J. J. [Radiobiological Institute of the Health Research Council TNO, Rijswijk (ZH) (Netherlands)

1968-03-15

Impairment of the reproductive capacity of cultured cells of human kidney origin (T-l{sub g} cells) has been measured by the Puck cloning technique. From the dose-survival curves obtained in these experiments by irradiation of cells in equilibrium with air and nitrogen, respectively, the relative biological effectiveness (RBE) and the oxygen enhancement ratios (OER) were determined for different beams of fast neutrons. Monoenergetic neutrons of 3 and 15 MeV energy, fission spectrum fast neutrons (mean energy about 1.5 MeV), neutrons produced by bombarding Be with cyclotron-accelerated 16 MeV deuterons (mean energy about 6 MeV) and neutrons produced by bombarding Be with cyclotron- accelerated 20 MeV {sup 3}He ions (mean energy about 10 MeV) have been compared with 250 kVp X-rays as a standard reference. The RBE for 50% cell survival varies from 4.7 for fission-spectrum fast neutrons to 2.7 for 15 MeV monoenergetic neutrons. The OER is not strongly dependent on the neutron energy for the various beams investigated. For the neutrons with the highest and lowest energies used OER values of 1.6 {+-} 0.2 and 1.5 {+-} 0.1 were measured. An interpretation of these data on the basis of the shapes of the LET spectra is proposed and an approximate verification of this hypothesis is provided from measurements in which secondary particle equilibrium was either provided for or deliberately eliminated. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Electronic imaging applied to neutron radiography

International Nuclear Information System (INIS)

Garrett, D.A.; Bracher, D.A.

1976-01-01

A commercially - available image intensifier was used with a scan conversion memory and a mobile 252 Cf based neutron radiography system to obtain neutron radiographs on a television monitor in 0.5 minutes to 10.0 minutes

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...

μ-TPC: a future standard instrument for low energy neutron field characterization

International Nuclear Information System (INIS)

Maire, D.; Lebreton, L.; Petit, M.; Billard, J.; Bourrion, O.; Bosson, G.; Guillaudin, O.; Lamblin, J.; Mayet, F.; Medard, J.; Muraz, J.F.; Richer, J.P.; Riffard, Q.; Santos, D.

2013-06-01

In order to measure energy of neutron fields, with energy ranging from 8 keV to 1 MeV, a new primary standard is being developed at the IRSN (Institute for Radioprotection and Nuclear Safety). This project, μ-TPC (Micro Time Projection Chamber), carried out in collaboration with the LPSC (Laboratoire de Physique Subatomique et de Cosmologie), is based on the nucleus recoil detector principle. The instrument will be presented with the associated method to measure the neutron energy. This article will emphasize the proton energy calibration procedure and energy measurements of a neutron field produced at 127 keV on the IRSN facility AMANDE. Finally the COMIMAC device, dedicated to the calibration, will be described. This original device, developed at the LPSC, is able to produce proton and electron beams with an accurate energy ranging from 1 keV to 50 keV. (authors)

Calibration methodology for proportional counters applied to yield measurements of a neutron burst

Energy Technology Data Exchange (ETDEWEB)

Tarifeño-Saldivia, Ariel, E-mail: atarifeno@cchen.cl, E-mail: atarisal@gmail.com; Pavez, Cristian; Soto, Leopoldo [Comisión Chilena de Energía Nuclear, Casilla 188-D, Santiago (Chile); Center for Research and Applications in Plasma Physics and Pulsed Power, P4, Santiago (Chile); Departamento de Ciencias Fisicas, Facultad de Ciencias Exactas, Universidad Andres Bello, Republica 220, Santiago (Chile); Mayer, Roberto E. [Instituto Balseiro and Centro Atómico Bariloche, Comisión Nacional de Energía Atómica and Universidad Nacional de Cuyo, San Carlos de Bariloche R8402AGP (Argentina)

2014-01-15

This paper introduces a methodology for the yield measurement of a neutron burst using neutron proportional counters. This methodology is to be applied when single neutron events cannot be resolved in time by nuclear standard electronics, or when a continuous current cannot be measured at the output of the counter. The methodology is based on the calibration of the counter in pulse mode, and the use of a statistical model to estimate the number of detected events from the accumulated charge resulting from the detection of the burst of neutrons. The model is developed and presented in full detail. For the measurement of fast neutron yields generated from plasma focus experiments using a moderated proportional counter, the implementation of the methodology is herein discussed. An experimental verification of the accuracy of the methodology is presented. An improvement of more than one order of magnitude in the accuracy of the detection system is obtained by using this methodology with respect to previous calibration methods.

Calibration methodology for proportional counters applied to yield measurements of a neutron burst

International Nuclear Information System (INIS)

Tarifeño-Saldivia, Ariel; Pavez, Cristian; Soto, Leopoldo; Mayer, Roberto E.

2014-01-01

This paper introduces a methodology for the yield measurement of a neutron burst using neutron proportional counters. This methodology is to be applied when single neutron events cannot be resolved in time by nuclear standard electronics, or when a continuous current cannot be measured at the output of the counter. The methodology is based on the calibration of the counter in pulse mode, and the use of a statistical model to estimate the number of detected events from the accumulated charge resulting from the detection of the burst of neutrons. The model is developed and presented in full detail. For the measurement of fast neutron yields generated from plasma focus experiments using a moderated proportional counter, the implementation of the methodology is herein discussed. An experimental verification of the accuracy of the methodology is presented. An improvement of more than one order of magnitude in the accuracy of the detection system is obtained by using this methodology with respect to previous calibration methods

A practical neutron shielding design based on data-base interpolation

International Nuclear Information System (INIS)

Jiang, S.H.; Sheu, R.J.

1993-01-01

Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry

Activation neutron detector

International Nuclear Information System (INIS)

Ambardanishvili, T.S.; Kolomiitsev, M.A.; Zakharina, T.Y.; Dundua, V.J.; Chikhladze, N.V.

1976-01-01

An activation neutron detector made as a moulded and cured composition of a material capable of being neutron-activated is described. The material is selected from a group consisting of at least two chemical elements, a compound of at least two chemical elements and their mixture, each of the chemical elements and their mixture, each of the chemical elements being capable of interacting with neutrons to form radioactive isotopes having different radiation energies when disintegrating. The material capable of being neutron-activated is distributed throughout the volume of a polycondensation resin inert with respect to neutrons and capable of curing. 17 Claims, No Drawings

Evaluation of the fluence to dose conversion coefficients for high energy neutrons using a voxel phantom coupled with the GEANT4 code

CERN Document Server

Paganini, S

2005-01-01

Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from Galactic cosmic radiation. Crews of future high-speed commercial flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the male adult voxels phantom MAX, developed in the Nuclear Energy Department of Pernambuco Federal University in Brazil, has been coupled with the Monte Carlo simulation code GEANT4. This toolkit, distributed and upgraded from the international scientific community of CERN/Switzerland, simulates thermal to ultrahigh energy neutrons transport and interactions in the matter. The high energy neutrons are pointed as the component that contribute about 70% of the neutron effective dose that represent the 35% to 60% total dose at aircraft altitude. In this research calculations of conversion coefficients from fluence to effective dose are performed for neutrons of energies from 100 MeV ...

Monte carlo calculation of energy-dependent response of high-sensitive neutron monitor, HISENS

International Nuclear Information System (INIS)

Imanaka, Tetsuji; Ebisawa, Tohru; Kobayashi, Keiji; Koide, Hiroaki; Seo, Takeshi; Kawano, Shinji

1988-01-01

A highly sensitive neutron monitor system, HISENS, has been developed to measure leakage neutrons from nuclear facilities. The counter system of HISENS contains a detector bank which consists of ten cylindrical proportional counters filled with 10 atm 3 He gas and a paraffin moderator mounted in an aluminum case. The size of the detector bank is 56 cm high, 66 cm wide and 10 cm thick. It is revealed by a calibration experiment using an 241 Am-Be neutron source that the sensitivity of HISENS is about 2000 times as large as that of a typical commercial rem-counter. Since HISENS is designed to have a high sensitivity in a wide range of neutron energy, the shape of its energy dependent response curve cannot be matched to that of the dose equivalent conversion factor. To estimate dose equivalent values from neutron counts by HISENS, it is necessary to know the energy and angular characteristics of both HISENS and the neutron field. The area of one side of the detector bank is 3700 cm 2 and the detection efficiency in the constant region of the response curve is about 30 %. Thus, the sensitivity of HISENS for this energy range is 740 cps/(n/cm 2 /sec). This value indicates the extremely high sensitivity of HISENS as compared with exsisting highly sensitive neutron monitors. (Nogami, K.)

Information about the new 8-group delayed neutron set preparation

International Nuclear Information System (INIS)

Svarny, J.

1998-01-01

Some comments to the present state concerning delayed neutron data preparation is given and preliminary analysis of the new 8-group delayed data (relative abundances) is presented. Comparisons of the 8-group to 6-group set is given for rod drop experiment (Unit 1, Cycle 14, NPP Dukovany).(Author)

Measurements for the energy calibration of the TANSY neutron detectors

International Nuclear Information System (INIS)

Drozdowicz, K.; Hoek, M.; Aronsson, D.

1990-05-01

The report describes measurements performed for the energy calibration of the TANSY neutron detectors (two arrays of 16 detectors each one). The calibration procedure determines four calibration parameters for each detector. Results of the calibration measurements are given and test measurements are presented. A relation of the neutron detector calibration parameters to producer's data for the photomulipliers is analysed. Also the tests necessary during normal operation of the TANSY neutron spectrometer are elaborated (passive and active tests). A method how to quickly get the calibration parameters for a spare detector in an array of the neutron detectors is included

Accurate measurements of neutron activation cross sections

International Nuclear Information System (INIS)

Semkova, V.

1999-01-01

The applications of some recent achievements of neutron activation method on high intensity neutron sources are considered from the view point of associated errors of cross sections data for neutron induced reaction. The important corrections in -y-spectrometry insuring precise determination of the induced radioactivity, methods for accurate determination of the energy and flux density of neutrons, produced by different sources, and investigations of deuterium beam composition are considered as factors determining the precision of the experimental data. The influence of the ion beam composition on the mean energy of neutrons has been investigated by measurement of the energy of neutrons induced by different magnetically analysed deuterium ion groups. Zr/Nb method for experimental determination of the neutron energy in the 13-15 MeV energy range allows to measure energy of neutrons from D-T reaction with uncertainty of 50 keV. Flux density spectra from D(d,n) E d = 9.53 MeV and Be(d,n) E d = 9.72 MeV are measured by PHRS and foil activation method. Future applications of the activation method on NG-12 are discussed. (author)

A medium energy neutron deep penetration experiment

International Nuclear Information System (INIS)

Amian, W.; Cloth, P.; Druecke, V.; Filges, D.; Paul, N.; Schaal, H.

1986-11-01

A deep penetration experiment conducted at the Los Alamos WNR facility's Spallation Neutron Target is compared with calculations using intra-nuclear-cascade and S N -transport codes installed at KFA-IRE. In the experiment medium energy reactions induced by neutrons between 15 MeV and about 150 MeV inside a quasi infinite slab of iron have been measured using copper foil monitors. Details of the experimental procedure and the theoretical methods are described. A comparison of absolute reaction rates for both experimentally and theoretically derived reactions is given. The present knowledge of the corresponding monitor reaction cross sections is discussed. (orig.)

Energy dependence of the neutron multiplicity P/sub nu/ in fast neutron induced fission of /sup 235,238/U and 239Pu

International Nuclear Information System (INIS)

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

1986-01-01

Certain applications require knowledge of the higher moments of the neutron multiplicity probability. It can be shown that the second factorial moment is proportional to the fission rate in the sample, and that the third factorial moment can be of use in disentangling spontaneous fission from induced fission. Using a source of unpublished work in which neutron multiplicities were derived for the fast neutron induced fission of U-235, U-238, and Pu-239, the multiplicity probability has been calculated as a function of neutron energy for the energy range 0 to 10 MeV

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

International Nuclear Information System (INIS)

Sekita, Junichiro

1988-01-01

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

Effect of gas filling pressure and operation energy on ion and neutron emission in a medium energy plasma focus device

Science.gov (United States)

Niranjan, Ram; Rout, R. K.; Srivastava, Rohit; Kaushik, T. C.

2018-03-01

The effects of gas filling pressure and operation energy on deuterium ions and neutrons have been studied in a medium energy plasma focus device, MEPF-12. The deuterium gas filling pressure was varied from 1 to 10 mbar at an operation energy of 9.7 kJ. Also, the operation energy was varied from 3.9 to 9.7 kJ at a deuterium gas filling pressure of 4 mbar. Time resolved emission of deuterium ions was measured using a Faraday cup. Simultaneously, time integrated and time resolved emissions of neutrons were measured using a silver activation detector and plastic scintillator detector, respectively. Various characteristics (fluence, peak density, and most probable energy) of deuterium ions were estimated using the Faraday cup signal. The fluence was found to be nearly independent of the gas filling pressure and operation energy, but the peak density and most probable energy of deuterium ions were found to be varying. The neutron yield was observed to be varying with the gas filling pressure and operation energy. The effect of ions on neutrons emission was observed at each operation condition.

High energy radiation from neutron stars

International Nuclear Information System (INIS)

Ruderman, M.

1985-04-01

Topics covered include young rapidly spinning pulsars; static gaps in outer magnetospheres; dynamic gaps in pulsar outer magnetospheres; pulse structure of energetic radiation sustained by outer gap pair production; outer gap radiation, Crab pulsar; outer gap radiation, the Vela pulsar; radioemission; and high energy radiation during the accretion spin-up of older neutron stars. 26 refs., 10 figs

Investigation of high-energy inelastic neutron scattering from liquid water confined in silica xerogel

International Nuclear Information System (INIS)

Perelli-Cippo, E.; Andreani, C.; Casalboni, M.; Dire, S.; Fernandez-Canoto, D.; Gorini, G.; Imberti, S.; Pietropaolo, A.; Prosposito, P.; Schutzmann, S.; Senesi, R.; Tardocchi, M.

2006-01-01

High-energy inelastic neutron scattering (HINS) employing epithermal neutrons is a new technique under development at the VESUVIO spectrometer at ISIS, aiming to access the high-energy and low wave-vector transfer region in neutron scattering experiments at eV energies. New neutron detectors have been developed for HINS based on the resonant detector (RD). These make use of the detection of prompt gammas after neutron absorption in an analyzer foil. The RD is used in the very low angle detector (VLAD) bank, which will extend the explored kinematical region to momentum transfer -1 , whilst still keeping energy transfer >300 meV. The final VLAD will cover the scattering range 1-5 o and will be installed by the end of 2005. The results obtained with prototype VLAD detectors on polycrystalline ice and liquid water in silica xerogels provide a demonstration of the feasibility of the measurements under realistic conditions

The determination of neutron energy spectra of radioisotope sources

International Nuclear Information System (INIS)

Lutkin, J.E.

1975-08-01

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

Report on the meeting of the working groups on neutron scattering applied to studying condensed matter, and neutron scattering and complementary methods applied in chemistry and biology

International Nuclear Information System (INIS)

1985-06-01

The present volume consists of 55 lectures. The subjects are: 1) Elastic neutron diffraction, 2) Lattice dynamics, 3) Diffusion, 4) Polymers, 5) Biology, 6) Methods and tools, 7) Magnetism. For distinct papers see hints under relevant topics. (BHO)

Leading neutron energy and pT distributions in deep inelastic scattering and photoproduction at HERA

International Nuclear Information System (INIS)

Chekanov, S.; Derrick, M.; Magill, S.

2007-02-01

The production of energetic neutrons in ep collisions has been studied with the ZEUS detector at HERA. The neutron energy and p T 2 distributions were measured with a forward neutron calorimeter and tracker in a 40 pb -1 sample of inclusive deep inelastic scattering (DIS) data and a 6 pb -1 sample of photoproduction data. The neutron yield in photoproduction is suppressed relative to DIS for the lower neutron energies and the neutrons have a steeper p T 2 distribution, consistent with the expectation from absorption models. The distributions are compared to HERA measurements of leading protons. The neutron energy and transverse-momentum distributions in DIS are compared to Monte Carlo simulations and to the predictions of particle exchange models. Models of pion exchange incorporating absorption and additional secondary meson exchanges give a good description of the data. (orig.)

Dose conversion coefficients for high-energy photons, electrons, neutrons and protons

International Nuclear Information System (INIS)

Sakamoto, Yukio

2005-01-01

Dose conversion coefficients for photons, electrons and neutrons based on new ICRP recommendations were cited in the ICRP Publication 74, but the energy ranges of these data were limited and there are no data for high energy radiations produced in accelerator facilities. For the purpose of designing the high intensity proton accelerator facilities at JAERI, the dose evaluation code system of high energy radiations based on the HERMES code was developed and the dose conversion coefficients of effective dose were evaluated for photons, neutrons and protons up to 10 GeV, and electrons up to 100 GeV. The dose conversion coefficients of effective dose equivalent were also evaluated using quality factors to consider the consistency between radiation weighting factors and Q-L relationship. The effective dose conversion coefficients obtained in this work were in good agreement with those recently evaluated by using FLUKA code for photons and electrons with all energies, and neutrons and protons below 500 MeV. There were some discrepancy between two data owing to the difference of cross sections in the nuclear reaction models. The dose conversion coefficients of effective dose equivalents for high energy radiations based on Q-L relation in ICRP Publication 60 were evaluated only in this work. The previous comparison between effective dose and effective dose equivalent made it clear that the radiation weighting factors for high energy neutrons and protons were overestimated and the modification was required. (author)

Low-energy neutron measurements in an iron calorimeter structure irradiated by 200 GeV/c hadrons

Energy Technology Data Exchange (ETDEWEB)

Russ, J S [Carnegie-Mellon University, Pittsburgh, PA (United States); Stevenson, G R; Fasso, A; Nielsen, M C [CERN, Geneva (Switzerland); Furetta, C; Rancoita, P G; Vismara, I [INFN, Milan (Italy)

1989-04-21

Of serious concern in the design of detectors for the new high-luminosity hadron-hadron colliders are the radiation damage effects on silicon and other detectors of low-energy neutrons produced by spallation evaporation or fission processes. Because of the lack of experimental information on the number of neutrons with energies between 0.1 and 10 MeV in the cascades originating from high-energy hadrons, an experiment was carried out using activation detector techniques to measure the neutron fluence in a cascade initiated by 200 GeV hadrons in acalorimeter-like iron structure. It was found that at the maximum of the cascade one produces approximately 3 neutrons per GeV of incident energy: some 70% of these are of energies between 0.1 and 5 MeV, the remainder are fairly uniformly distributed in energy between 5 and several hundred MeV. The number of albedo neutrons leaving the front face of the calorimeter structure was about 0.3 neutrons per GeV of incident energy with in energy distribution similar to those at cascade maximum These data confirm that neutron-induced damage will he of concern in the design of detectors for the new colliders and that further measurements and calculations are necessary for a correct assessment of this damage. (author)

Determination of the total neutron cross section using average energy shift method for filtered neutron beam