Using anisotropies in prompt fission neutron coincidences to assess the neutron multiplication of highly multiplying subcritical plutonium assemblies

Energy Technology Data Exchange (ETDEWEB)

Mueller, J.M., E-mail: jonathan_mueller@ncsu.edu; Mattingly, J.

2016-07-21

There is a significant and well-known anisotropy between the prompt neutrons emitted from a single fission event; these neutrons are most likely to be observed at angles near 0° or 180° relative to each other. However, the propagation of this anisotropy through different generations of a fission chain reaction has not been previously studied. We have measured this anisotropy in neutron–neutron coincidences from a subcritical highly-multiplying assembly of plutonium metal. The assembly was a 4.5 kg α-phase plutonium metal sphere composed of 94% {sup 239}Pu and 6% {sup 240}Pu by mass. Data were collected using two EJ-309 liquid scintillators and two EJ-299 plastic scintillators. The angular distribution of neutron–neutron coincidences was measured at 90° and 180° and found to be largely isotropic. Simulations were performed using MCNPX-PoliMi of similar plutonium metal spheres of varying sizes and a correlation between the neutron multiplication of the assembly and the anisotropy of neutron–neutron coincidences was observed. In principle, this correlation could be used to assess the neutron multiplication of an unknown assembly.

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.

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

International Nuclear Information System (INIS)

Rassow, J.

1984-01-01

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

Angular distribution of oriented nucleus fission neutrons

International Nuclear Information System (INIS)

Barabanov, A.L.; Grechukhin, D.P.

1982-01-01

Calculations of anisotropy of angular distribution of oriented 235 U nuclei thermal fission neutrons have been carried out. the neutrons were assumed to evaporate isotropically by completely accelerated fragements in the fragment system with only its small part, i. e. fission-producing neutrons, emitted at the moment of neck break. It has been found out that at low energies of neutrons Esub(n)=1-2 MeV the sensitivity of the angular distribution anisotropy to variations of spectrum of neutron evaporation from fragments and the magnitude of a share of fission-producing neutrons reaches approximately 100%, which at high energies, Esub(n) > 5 MeV it does not exceed approximately 20%. Therefore the angular distribution of fast neutrons to a greater degree of confidence may be used for restoring the angular distribution anisotropy of fragments while the angular distribution of low energy neutrons may be used for deriving information on the fission process, but only in case 6f the experiment accuracy is better than approximately 3%

Research on neutron source multiplication method in nuclear critical safety

International Nuclear Information System (INIS)

Zhu Qingfu; Shi Yongqian; Hu Dingsheng

2005-01-01

The paper concerns in the neutron source multiplication method research in nuclear critical safety. Based on the neutron diffusion equation with external neutron source the effective sub-critical multiplication factor k s is deduced, and k s is different to the effective neutron multiplication factor k eff in the case of sub-critical system with external neutron source. The verification experiment on the sub-critical system indicates that the parameter measured with neutron source multiplication method is k s , and k s is related to the external neutron source position in sub-critical system and external neutron source spectrum. The relation between k s and k eff and the effect of them on nuclear critical safety is discussed. (author)

Neutron recognition in the LAND detector for large neutron multiplicity

Energy Technology Data Exchange (ETDEWEB)

Pawlowski, P., E-mail: piotr.pawlowski@ifj.edu.pl [Institute of Nuclear Physics, PAN, Radzikowskiego 152, 31-342 Krakow (Poland); Brzychczyk, J. [Institute of Physics, Jagiellonian University, Reymonta 4, 30-059 Krakow (Poland); Leifels, Y.; Trautmann, W. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Adrich, P. [National Centre for Nuclear Research, PL-00681 Warsaw (Poland); GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Aumann, T. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Bacri, C.O. [Institut de Physique Nucleaire, IN2P3-CNRS et Universite, F-91406 Orsay (France); Barczyk, T. [Institute of Physics, Jagiellonian University, Reymonta 4, 30-059 Krakow (Poland); Bassini, R. [Istituto di Scienze Fisiche, Universita degli Studi and INFN, I-20133 Milano (Italy); Bianchin, S. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Boiano, C. [Istituto di Scienze Fisiche, Universita degli Studi and INFN, I-20133 Milano (Italy); Boretzky, K. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Boudard, A. [IRFU/SPhN, CEA/Saclay, F-91191 Gif-sur-Yvette (France); Chbihi, A. [GANIL, CEA et IN2P3-CNRS, F-14076 Caen (France); Cibor, J.; Czech, B. [Institute of Nuclear Physics, PAN, Radzikowskiego 152, 31-342 Krakow (Poland); De Napoli, M. [Dipartimento di Fisica e Astronomia-Universita and INFN-CT and LNS, I-95123 Catania (Italy); and others

2012-12-01

The performance of the LAND neutron detector is studied. Using an event-mixing technique based on one-neutron data obtained in the S107 experiment at the GSI laboratory, we test the efficiency of various analytic tools used to determine the multiplicity and kinematic properties of detected neutrons. A new algorithm developed recently for recognizing neutron showers from spectator decays in the ALADIN experiment S254 is described in detail. Its performance is assessed in comparison with other methods. The properties of the observed neutron events are used to estimate the detection efficiency of LAND in this experiment.

Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

International Nuclear Information System (INIS)

Kim, S.S.; Levine, S.H.

1985-01-01

An analytical/experimental method has been developed to monitor the subcritical reactivity and unfold the k/sub infinity/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. Therefore, it is called the Asymmetric Multiple Position Neutron Source (AMPNS) method. The AMPNS method employs nucleonic codes to analyze the neutron multiplication of a Cf-252 neutron source. An optimization program, GPM, is utilized to unfold the k/sub infinity/ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn State Breazeale TRIGA Reactor (PSBR). A significant result of this study is that it provides a method to monitor the criticality of a damaged core during the recovery period

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)

MCRTOF, Multiple Scattering of Resonance Region Neutron in Time of Flight Experiments

International Nuclear Information System (INIS)

Ohkubo, Mako

1984-01-01

1 - Description of program or function: Multiple scattering of neutrons in the resonance energy region impinging on a disk with an arbitrary angle. 2 - Method of solution: The Monte Carlo method is employed to simulate the path of an incident neutron in a medium for which macroscopic cross sections are determined by resonance parameters. By tracing a large number of neutrons, probabilities for capture, transmission, front-face scattering, rear-face scattering and side-face scattering are determined and printed out as function of incident neutron energy. Optionally, the distribution of capture locations in the disk can be printed. The incident neutron energy is swept to fit a situation as encountered in time-of-flight experiments. 3 - Restrictions on the complexity of the problem: The cross section file is constructed from input resonance parameters with a single- level Breit-Wigner formula. The following restrictions and simplifications apply: - The maximum number of resonances is five. - Reactions other than capture and scattering are neglected. - The angular scattering distribution in the center-of-mass system is assumed to be uniform. - Chemical binding effects are neglected

Inventory verification measurements using neutron multiplicity counting

International Nuclear Information System (INIS)

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

1998-01-01

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

FB-line neutron multiplicity counter operation manual

International Nuclear Information System (INIS)

Langner, D.G.; Sweet, M.R.; Salazar, S.D.; Kroncke, K.E.

1998-01-01

This manual describes the design features, performance, and operating characteristics for the FB-Line Neutron Multiplicity Counter (FBLNMC). The FBLNMC counts neutron multiplicities to quantitatively assay plutonium in many forms, including impure scrap and waste. Monte Carlo neutronic calculations were used to design the high-efficiency (57%) detector that has 113 3 H tubes in a high-density polyethylene body. The new derandomizer circuit is included in the design to reduce deadtime. The FBLNMC can be applied to plutonium masses in the range from a few tens of grams to 5 kg; both conventional coincidence counting and multiplicity counting can be used as appropriate. This manual gives the performance data and preliminary calibration parameters for the FBLNMC

Neutron magnetic multiple diffraction in a natural magnetite crystal

International Nuclear Information System (INIS)

Mazzocchi, V.L.; Parente, C.B.R.

1988-09-01

Neutron multiple diffraction has been employed in the study of the magnetism in magnetite (Fe 3 O 4 ). Magnetite has a crystallographic structure of an inverted spinel with tetrahedral A sites occupied solely by trivalent Fe 3+ ions and octahedral B sites occupied both by divalent Fe 2+ ions and the remaining Fe 3+ ions in random distribution. At room temperature magnetite is a Neel A-B ferrimagnet where the ions on the A, B sites are coupled antiferromagneticaly. This coupling disappears at T sup c approx. or approx.= 580 0 C. Employing a natural single crystal of magnetite experimental neutron multiple diffraction patterns were obtained for the primary reflection 111 at room temperature and 703 0 C. This reflection is almost entirely magnetic in origin resulting in 'Aufhellung' patterns below T c and mixed 'Aufhellung-Umweganregung' patterns above T c . Theoretical patterns were calculated employing the iterative method for the approximation of intensities by a Taylor series and compared to the experimental results. (author) [pt

Neutron multiplication in lead in the experiments with neutron generators

International Nuclear Information System (INIS)

Markovskij, D.V.

1989-01-01

A calculational analysis of neutron multiplication in lead, including the estimates of multiplication limits for the standard ENDF/BIV data set and the effects of various changes in the data themselves is performed. 10 refs, 5 figs

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

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

Yields of correlated fragment pairs and neutron multiplicity in spontaneous fission of {sup 242}Pu

Energy Technology Data Exchange (ETDEWEB)

Veselsky, M.; Kliman, J.; Morhaccaron, M. [Institute of Physics of Slovak Academy of Sciences, Dubravska 9, 84228 Bratislava (Slovakia); Ramayya, A.V.; Kormicki, J.; Daniel, A.V. [Physics Department, Vanderbilt University, Nashville (United States)] Rasmussen, J.O. [Lawrence Berkeley National Laboratory, Berkeley (United States)] Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore (United States); Daniel, A.V.; Popeko, G.S.; Oganessian, Yu. Ts. [Joint Institute for Nuclear Research, Dubna (Russia)] Greiner, W. [Institut fur Theoretische Physik, J. W. Goethe Universitaet, Frankfurt a. M. (Germany); Aryaeinejad, R. [Idaho National Engineering Laboratory, Idaho Falls (United States)

1998-10-01

Yields of correlated fragment pairs were obtained in spontaneous fission of {sup 242}Pu. Charge, mass and neutron multiplicity distributions of fragment pairs were determined and compared to available data. The yield of cold fission without neutron emission was determined to about 10{percent} for the set of observed correlated fragment pairs. {copyright} {ital 1998 American Institute of Physics.}

Neutron Flux Distribution on Neutron Radiography Facility After Fixing the Collimator

International Nuclear Information System (INIS)

Supandi; Parikin; Mohtar; Sunardi; Roestam, S

1996-01-01

The Radiography Neutron Facility consists of an inner collimator, outer collimator, main shutter, second shutter and the sample chamber with 300 mm in diameter. Neutron beam quality depends on the neutron flux intensities distribution, L/D ratio Cd ratio, neutron/gamma ratio. The results show that the neutron flux intensity was 2.83 x 107 n cm-2.s-1, with deviation of + 7.8 % and it was distributed homogeneously at the sample position of 200 mm diameter. The beam characteristics were L/D ratio 98 and Rod 8, and neutron gamma ratio 3.08 x 105n.cm-2.mR-1 and Reactor Power was 20 MW. This technique can be used to examine sample with diameter of < 200 mm

The relationship between neutron multiplication and keff

International Nuclear Information System (INIS)

Brewer, R.W.

1995-01-01

In recent years the International Criticality Safety Benchmark Evaluation Project under the sponsorship of the Department of Energy has undertaken the task of evaluating past critical experiments. Many of the experiments involving metals were subcritical with extrapolation to some critical characteristic dimension. The metal experiments were commonly limited to a maximum multiplication of 100 for obvious safety considerations. Also many critical experiments often used subcritical measurements to obtain the critical specifications, e.g. Jezebel used subcritical measurements to assess the magnitude of neutron reflection from the surrounding structures. Therefore, the task of evaluating the experimentally derived critical configuration often involves evaluating the subcritical measurements made by the experimentalist. The purpose of past experiments was to determine critical configurations. Many of the modem computer codes (KENO, MCNP, and ONEDANT) calculate values of k eff . However, the subcritical measurements made during the course of the experiment are usually measurements of the neutron multiplication. To evaluate the subcritical experiments, a link was established between the neutronic theory and the practical application of such when using subcritical measurements to establish the critical characteristic dimension. A more in depth derivation of the relationship between k eff and neutron multiplication will be shown along with comparisons between calculated and measured multiplications

Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

Science.gov (United States)

Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

2018-03-01

In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

Active neutron multiplicity analysis and Monte Carlo calculations

International Nuclear Information System (INIS)

Krick, M.S.; Ensslin, N.; Langner, D.G.; Miller, M.C.; Siebelist, R.; Stewart, J.E.; Ceo, R.N.; May, P.K.; Collins, L.L. Jr

1994-01-01

Active neutron multiplicity measurements of high-enrichment uranium metal and oxide samples have been made at Los Alamos and Y-12. The data from the measurements of standards at Los Alamos were analyzed to obtain values for neutron multiplication and source-sample coupling. These results are compared to equivalent results obtained from Monte Carlo calculations. An approximate relationship between coupling and multiplication is derived and used to correct doubles rates for multiplication and coupling. The utility of singles counting for uranium samples is also examined

Characterizations of double pulsing in neutron multiplicity and coincidence counting systems

Energy Technology Data Exchange (ETDEWEB)

Koehler, Katrina E., E-mail: kkoehler@lanl.gov [Los Alamos National Laboratory, P. O. Box 1663, Los Alamos, NM 87545 (United States); Henzl, Vladimir [Los Alamos National Laboratory, P. O. Box 1663, Los Alamos, NM 87545 (United States); Croft, Stephen S. [Oak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN 37831 (United States); Henzlova, Daniela; Santi, Peter A. [Los Alamos National Laboratory, P. O. Box 1663, Los Alamos, NM 87545 (United States)

2016-10-01

Passive neutron coincidence/multiplicity counters are subject to non-ideal behavior, such as double pulsing and dead time. It has been shown in the past that double-pulsing exhibits a distinct signature in a Rossi-alpha distribution, which is not readily noticed using traditional Multiplicity Shift Register analysis. However, it has been assumed that the use of a pre-delay in shift register analysis removes any effects of double pulsing. In this work, we use high-fidelity simulations accompanied by experimental measurements to study the effects of double pulsing on multiplicity rates. By exploiting the information from the double pulsing signature peak observable in the Rossi-alpha distribution, the double pulsing fraction can be determined. Algebraic correction factors for the multiplicity rates in terms of the double pulsing fraction have been developed. We discuss the role of these corrections across a range of scenarios.

The 4π neutron detector CARMEN

Energy Technology Data Exchange (ETDEWEB)

Ledoux, X., E-mail: Xavier.ledoux@ganil.fr [CEA/DAM/DIF, F-91297 Arpajon (France); GANIL, CEA/DRF-CNRS/IN2P3, Caen, F-14076 France (France); Laborie, J.-M.; Pras, P.; Lantuéjoul-Thfoin, I.; Varignon, C. [CEA/DAM/DIF, F-91297 Arpajon (France)

2017-02-01

CARMEN is a 4π neutron detector filled with a gadolinium-loaded liquid scintillator built to measure neutron multiplicity distributions. It is used to study fission and (n,xn) reactions. In addition to neutron multiplicity measurements, CARMEN can be used to measure neutron energy spectra with the time-of-flight technique, thanks to the time properties of the prompt signal. The detector, detection technique and efficiency determination are presented in detail. Two examples are also presented: the measurement of {sup 252}Cf spontaneous fission neutron multiplicity probability distribution and the measurement of the neutron energy spectrum emitted by an Am-Be radioactive source.

To the question of definition of fissile material mass and neutron multiplication in deep sub-critical systems

International Nuclear Information System (INIS)

Dulin, V.V.

2006-01-01

A method of determination neutrons multiplication in deep sub-critical multiplying media has been developed. It is based on a modified of Rossi - alpha method. It will consist in use of integral on time (a method of the areas) from correlated parts of distribution and integral in area, independent of time a part of distribution (area of a constant background). It allows to spend the calculated analysis, using the integrated equation on time for a neutrons flux and to not use representation of point kinetic model. A calculation spatially-correlation factor the adjoint (relative the detector count rate) inhomogeneous equation is used. Its calculation takes into account fission both in multiplying media and in a spontaneous neutron source. Measurements with plutonium-steel and uranium-steel blocks, and blocks from uranium and plutonium dioxide of different enrichment are have been carried out. The measured values of neutrons multiplication in a range 1.03-1.82 will be well coordinated to results of calculations. The question on an opportunity of definition of weight of the measured blocks of fissile material is considered [ru

Neutron flux distribution forecasting device of reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1991-01-01

A neutron flux distribution is forecast by using current data obtained from a reactor. That is, the device of the present invention comprises (1) a neutron flux monitor disposed in various positions in the reactor, (2) a forecasting means for calculating and forecasting a one-dimensional neutron flux distribution relative to imaginable events by using data obtained from the neutron flux monitor and physical models, and (3) a display means for displaying the results forecast in the forecasting means to a reactor operation console. Since the forecast values for the one-dimensional neutron flux distribution relative to the imaginable events are calculated in the device of the present invention by using data obtained from the neutron flux monitor and the physical models, the data as a base of the calculation are new and the period for calculating the forecast values can be shortened. Accordingly, although there is a worry of providing some errors in the forecast values, they can be utilized sufficiently as reference data. As a result, the reactor can be operated more appropriately. (I.N.)

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Correction for variable moderation and multiplication effects associated with thermal neutron coincidence counting

International Nuclear Information System (INIS)

Baron, N.

1978-01-01

A correction is described for multiplication and moderation when doing passive thermal neutron coincidence counting nondestructive assay measurements on powder samples of PuO 2 mixed arbitrarily with MgO, SiO 2 , and moderating material. The multiplication correction expression is shown to be approximately separable into the product of two independent terms; F/sub Pu/ which depends on the mass of 240 Pu, and F/sub αn/ which depends on properties of the matrix material. Necessary assumptions for separability are (1) isotopic abundances are constant, and (2) fission cross sections are independent of incident neutron energy: both of which are reasonable for the 8% 240 Pu powder samples considered here. Furthermore since all prompt fission neutrons are expected to have nearly the same energy distributions, variations among different samples can be due only to the moderating properties of the samples. Relative energy distributions are provided by a thermal neutron well counter having two concentric rings of 3 He proportional counters placed symmetrically about the well. Measured outer-to-inner ring ratios raised to an empirically determined power for coincidences, (N/sup I//N/sup O/)/sup Z/, and singles, (T/sup O//T/sup I/)/sup delta/, provide corrections for moderation and F/sub αn/ respectively, and F/sub Pu/ is approximated by M 240 /sup X//M 240 . The exponents are calibration constants determined by a least squares fitting procedure using standards' data. System calibration is greatly simplified using the separability principle. Once appropriate models are established for F/sub Pu/ and F/sub αn/, only a few standards are necessary to determine the calibration constants associated with these terms. Since F/sub Pu/ is expressed as a function of M 240 , correction for multiplication in a subsequent assay demands only a measurement of F/sub αn/

Multiplicity of neutrons from violent heavy-ion collisions: 40Ar + Th and U from 10 to 77 MeV/u

International Nuclear Information System (INIS)

Jahnke, U.; Cramer, B.; Ingold, G.; Schwinn, E.; Charvet, J.L.; Frehaut, J.; Lott, B.; Morjean, M.; Patin, Y.; Pranal, Y.; Uzureau, J.L.; Doubre, H.; Galin, J.; Guerreau, D.; Jiang, D.X.; Pouthas, J.; Sokolov, A.; Gatty, B.; Jacquet, D.; Magnago, C.

1988-12-01

With large 4π scintillator tanks the multiplicity of neutrons released from the most dissipative reactions in 10 to 77 MeV/u 40 Ar induced collisions with Th and U nuclei has been investigated. The central issue is the relation between the fission-fragment folding angle or linear momentum transfer and the multiplicity of evaporated neutrons and charged particles or dissipated energy. Their multiplicity points to a 'soft saturation' of the maximum energy deposit with increasing bombarding energy near 700 MeV of excitation. Unlike the folding-angle distributions, the inclusive neutron multiplicity spectra, which are unbiased by specific decay properties of the intermediate nuclear system, do not show a decline of the most dissipative processes within this range of incident energy. (orig.)

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

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

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

NEUTRON SPECTRUM MEASUREMENTS USING MULTIPLE THRESHOLD DETECTORS

Energy Technology Data Exchange (ETDEWEB)

Gerken, William W.; Duffey, Dick

1963-11-15

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

Effect of fission dynamics on the spectra and multiplicities of prompt fission neutrons

International Nuclear Information System (INIS)

Nix, J.R.; Madland, D.G.; Sierk, A.J.

1985-01-01

With the goal of examining their effect on the spectra and multiplicities of the prompt neutrons emitted in fission, we discuss recent advances in a unified macroscopic-microscopic description of large-amplitude collective nuclear dynamics. The conversion of collective energy into single-particle excitation energy is calculated for a new surface-plus-window dissipation mechanism. By solving the Hamilton equations of motion for initial conditions appropriate to fission, we obtain the average fission-fragment translational kinetic energy and excitation energy. The spectra and multiplicities of the emitted neutrons, which depend critically upon the average excitation energy, are then calculated on the basis of standard nuclear evaporation theory, taking into account the average motion of the fission fragments, the distribution of fission-fragment residual nuclear temperature, the energy dependence of the cross section for the inverse process of compound-nucleus formation, and the possibility of multiple-chance fission. Some illustrative comparisons of our calculations with experimental data are shown

Measurement of accelerator-based neutron distributions using nuclear track detectors

International Nuclear Information System (INIS)

Al-Jarallah, M.I.; Abu-Jarad, F.; Rehman, Fazal-ur-; Khiari, F.Z.; Aksoy, A.; Nassar, R.

2000-01-01

Nuclear track detectors were used to measure the longitudinal and transverse distributions of slow neutrons in a moderated neutron field as well as the longitudinal and transverse distributions of fast neutrons produced on the 0 deg. beam line of the KFUPM 350 keV ion accelerator. The neutrons were first produced from the T(d,n) 4 He reaction with a neutron energy of approximately 14 MeV and were then moderated in a cylindrical polyethylene moderator placed at the end of the 0 deg. beam line. The optimal transverse slow neutron distribution was found to be uniform within ±4.5% at a 3 cm depth inside the moderator. The fast neutron distribution component along the moderator central axis exhibited an exponential-like drop in intensity with depth. Linearity checks of alpha and proton recoil track density with irradiation time for the nuclear track detectors were verified for both slow and fast neutrons

Measurement of accelerator-based neutron distributions using nuclear track detectors

Energy Technology Data Exchange (ETDEWEB)

Al-Jarallah, M.I. E-mail: mibrahim@kfupm.edu.sa; Abu-Jarad, F.; Rehman, Fazal-ur-; Khiari, F.Z.; Aksoy, A.; Nassar, R

2000-12-01

Nuclear track detectors were used to measure the longitudinal and transverse distributions of slow neutrons in a moderated neutron field as well as the longitudinal and transverse distributions of fast neutrons produced on the 0 deg. beam line of the KFUPM 350 keV ion accelerator. The neutrons were first produced from the T(d,n){sup 4}He reaction with a neutron energy of approximately 14 MeV and were then moderated in a cylindrical polyethylene moderator placed at the end of the 0 deg. beam line. The optimal transverse slow neutron distribution was found to be uniform within {+-}4.5% at a 3 cm depth inside the moderator. The fast neutron distribution component along the moderator central axis exhibited an exponential-like drop in intensity with depth. Linearity checks of alpha and proton recoil track density with irradiation time for the nuclear track detectors were verified for both slow and fast neutrons.

Research on determine the absolute neutron output of distributed pulse generators

International Nuclear Information System (INIS)

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

2009-01-01

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

Neutron multiplicity of fission fragments

Energy Technology Data Exchange (ETDEWEB)

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

1995-10-01

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

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

Science.gov (United States)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-07-01

An extension of the point kinetics model is developed to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. The spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

Experimental determination of the slow-neutron wavelength distribution

DEFF Research Database (Denmark)

Lebech, Bente; Mikke, K.; Sledziewska-Blocka, D.

1970-01-01

Different experiments for determining the slow-neutron wavelength distribution in the region 227-3 meV have been carried out, and the results compared. It is concluded that the slow-neutron wave-length distribution can be determined accurately by elastic scattering on a pure incoherent or a pure...

Effects of (α,n) contaminants and sample multiplication on statistical neutron correlation measurements

International Nuclear Information System (INIS)

Dowdy, E.J.; Hansen, G.E.; Robba, A.A.; Pratt, J.C.

1980-01-01

The complete formalism for the use of statistical neutron fluctuation measurements for the nondestructive assay of fissionable materials has been developed. This formalism includes the effect of detector deadtime, neutron multiplicity, random neutron pulse contributions from (α,n) contaminants in the sample, and the sample multiplication of both fission-related and background neutrons

Experimental determination of neutron temperature distribution in reactor cell

International Nuclear Information System (INIS)

Bosevski, T.

1965-12-01

This paper describes theoretical preparation of the experiment for measuring neutron temperature distribution at the RB reactor by activation foils. Due to rather low neutron flux Cu and Lu foil were irradiated for 4 days. Special natural uranium fuel element was prepared to enable easy removal of foils after irradiation. Experimental device was placed in the reactor core at half height in order to measure directly the mean neutron density. Experimental data of neutron temperature distribution for square lattice pitch 16 cm are presented with mean values of neutron temperature in the moderator, in the fuel and on the fuel element surface

Measurement of the neutron flux distributions, epithermal index, Westcott thermal neutron flux in the irradiation capsules of hydraulic conveyer (Hyd) and pneumatic tubes (Pn) facilities of the KUR

International Nuclear Information System (INIS)

Chatani, Hiroshi

2001-05-01

The reactions of Au(n, γ) 198 Au and Ti(n, p) 47 or 48 Sc were used for the measurements of the thermal and epithermal (thermal + epithermal) and the fast neutron flux distributions, respectively. In the case of Hyd (Hydraulic conveyer), the thermal + epithermal and fast neutron flux distributions in the horizontal direction in the capsule are especially flat; the distortion of the fluxes are 0.6% and 5.4%, respectively. However, these neutron fluxes in the vertical direction are low at the top and high at the bottom of the capsule. These differences between the top and bottom are 14% for both distributions. On the other hand, in polyethylene capsules of Pn-1, 2, 3 (Pneumatic tubes Nos. 1, 2, 3), in contrast with Hyd, these neutron flux distributions in the horizontal direction have gradients of 8 - 18% per 2.5 cm diameter, and those on the vertical axis have a distortion of approximately 5%. The strength of the epithermal dE/E component relative to the neutron density including both thermal and epithermal neutrons, i.e., the epithermal index, for the hydraulic conveyer (Hyd) and pneumatic tube No.2 (Pn-2), in which the irradiation experiments can be achieved, are determined by the multiple foil activation method using the reactions of Au(n, γ) 198 Au and Co(n, γ) 60(m+g) Co. The epithermal index observed in an aluminum capsule of Hyd is 0.034-0.04, and the Westcott thermal neutron flux is 1.2x10 14 cm -2 sec -1 at approximately 1 cm above the bottom. The epithermal index in a Pn-2 polyethylene capsule was measured by not only the multiple foil activation method but also the Cd-ratio method in which the Au(n, γ) 198 Au reaction in a cadmium cover is also used. The epithermal index is 0.045 - 0.055, and the thermal neutron flux is 1.8x10 13 cm -2 sec -1 . (J.P.N.)

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.

Lubricant distribution determination by neutron radiography

International Nuclear Information System (INIS)

Stewart, P.A.E.

1979-01-01

Cold neutron radiation of energy less than 0.025 eV having a flux greater than 1 x 10 3 neutrons/square centimeter/second is used to diagnose temporal information about the spatial distribution of hydrocarbon fuel and lubrication oils in internal combustion engines, gas turbine engines and fuel systems. Images of the movement of fuel or oil are recorded by directing a beam of neutrons through an engine and using an image intensifier responsive to low light levels to intensify an image formed by neutrons which have been directed through the engine onto a fluorescent screen. The output image from the intensifier is recorded by a video or cine camera

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

Energy Technology Data Exchange (ETDEWEB)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; Folsom, Micah; Kouzes, Richard; Kukharev, Vladislav; Lintereur, Azaree; Robinson, Sean; Siciliano, Edward; Stave, Sean; Valdez, Patrick

2018-04-01

Past 3He shortages led to investigations into replacement options for neutron detectors in systems that previously used 3He-based technologies. The goal of this research was to investigate the feasibility of a full-scale lithium fluoride with silver activated zinc sulfide (LiF/ZnS) based neutron multiplicity counter. The LiF/ZnS based neutron multiplicity counter (LiNMC) was developed based on an iterative process between modeling and experimental measurements. Each active region of the LiNMC contains five sheets of LiF/ZnS sandwiched between six sheets of wavelength shifting plastic to form neutron detection stacks. The wavelength shifted scintillation light was collected by photomultiplier tubes located on each end of the stacks. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high density polyethylene blocks in the corners to reflect high energy neutrons and capture low energy neutrons. Preliminary calibration with a 252Cf neutron source showed that the LiNMC was able to achieve 36% neutron detection efficiency (ε) and an 11.7 μs neutron die-away time (τ) for a doubles Figure-of-merit (ε2/ τ) of 109. This is the highest doubles Figure-of-merit performance measured to-date for a 3He-free neutron multiplicity counter system. By the end of this project, the LiNMC’s basic components were integrated into a single laboratory scale system capable of proof-of-concept measurements.

Statistical γ-ray multiplicity distributions in Dy and Yb nuclei

International Nuclear Information System (INIS)

Tveter, T.S.; Bergholt, L.; Guttormsen, M.; Rekstad, J.

1994-03-01

The statistical γ-ray multiplicity distributions following the reactions 163 Dy( 3 He,αxn) 162-x Dy and 173 Yb( 3 He,αxn) 172-x Yb have been studied. The mean value and standard deviation have been extracted as functions of excitation energy. The method is based on the probability distribution of k-fold events, where an α-particle is observed in coincidence with signals in k γ-ray detectors. Techniques for isolating statistical γ-rays and subtracting random background, cross-talk and neutron contributions are discussed. 22 refs., 10 figs., 3 tabs

Neutron importance and the generalized Green function for the conventionally critical reactor with normalized neutron distribution

International Nuclear Information System (INIS)

Khromov, V.V.

1978-01-01

The notion of neutron importance when applied to nuclear reactor statics problems described by time-independent homogeneous equations of neutron transport with provision for normalization of neutron distribution is considered. An equation has been obtained for the function of neutron importance in a conditionally critical reactor with respect to an arbitrary nons linear functional determined for the normalized neutron distribution. Relation between this function and the generalized Green function of the selfconjugated operator of the reactor equation is determined and the formula of small perturbations for the functionals of a conditionally critical reactor is deduced

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

Science.gov (United States)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; Folsom, Micah; Kouzes, Richard; Kukharev, Vladislav; Lintereur, Azaree; Robinson, Sean; Siciliano, Edward; Stave, Sean; Valdez, Patrick

2018-04-01

The feasibility of a full-scale lithium fluoride zinc sulfide (LiF/ZnS) based neutron multiplicity counter has been demonstrated. The counter was constructed of modular neutron detecting stacks that each contain five sheets of LiF/ZnS interleaved between six sheets of wavelength shifting plastic with a photomultiplier tube on each end. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high-density polyethylene blocks in the corners to reflect high-energy neutrons and capture low-energy neutrons. The final system design was optimized via modeling and small-scale test. Measuring neutrons from a 252Cf source, the counter achieved a 36% neutron detection efficiency (ɛ) and an 11 . 7 μs neutron die-away time (τ) for a doubles figure-of-merit (ɛ2 / τ) of 109. This is the highest doubles figure-of-merit measured to-date for a 3He-free neutron multiplicity counter.

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

International Nuclear Information System (INIS)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-01-01

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

Fission meter and neutron detection using poisson distribution comparison

Science.gov (United States)

Rowland, Mark S; Snyderman, Neal J

2014-11-18

A neutron detector system and method for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. Comparison of the observed neutron count distribution with a Poisson distribution is performed to distinguish fissile material from non-fissile material.

Neutron spatial distribution measurement with 6Li-contained thermoluminescent sheets

International Nuclear Information System (INIS)

Konnai, A.; Odano, N.; Sawamura, H.; Ozasa, N.; Ishikawa, Y.

2006-01-01

We have been developing a thermoluminescent (TL) sheet for photon dosimetry (TL sheet) with thermoluminescent material of LiF:Mg, Cu, P and a co-polymer of ethylene and tetrafluoroethylene. For the purpose of a development of simple method for neutron spatial distribution measurement, TL sheet for neutron detection (NTL sheet) is made by adding 94.7% enriched 6 LiF to TL sheet. TL material in TL sheet is directly excited by ionizing radiation whereas, in the case of neutron detection, TL material in NTL sheet is indirectly excited by neutron capture reaction. That is neutron distribution can be obtained with TL caused by α particle from 6 Li(n, α) 3 H reaction. Responses of NTL sheets to neutrons were examined at the neutron beam irradiation facility for Boron Neutron Capture Therapy (BNCT) in JRR-4 research reactor in Japan Atomic Energy Agency. TL and NTL sheets were exposed to striped and roundly distributed neutron fields. Attenuations of neutron flux in air and water were also observed using NTL sheets. TL sheets were also exposed on the same conditions and compared with NTL sheets. TL intensity ratios of NTL sheet to TL sheet were consistent with the calculated value from 6 Li content. Thermal neutron attenuation observed by NTL sheet also corresponded with the result measured by Au wire radioactivation and TLD chips, which were currently used in BNCT at JRR-4. These results were analyzed with by Monte Carlo simulation. The present results indicated that NTL sheet is applicable to measurement of neutron spatial distribution. (author)

neutron multiplicity measurements on 220 l waste drums containing Pu in the range 0.1-1 g 240Pueff with the time interval analysis method

International Nuclear Information System (INIS)

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

1998-01-01

Measurement results are presented for the assay of plutonium in 220 l waste drums containing Pu-masses in the range 0.1-1 g 240 Pu eff obtained with the time interval analysis (TIA) method. TIA is a neutron multiplicity method based on the concept of one- and two-dimensional Rossi-alpha distributions. The main source of measurement bias in neutron multiplicity measurements at low count-rates is the impredictable variation of the high-multiplicity neutron background of spallation neutrons induced by cosmic rays. The TIA-method was therefore equipped with a special background filter, which is designed and optimized to reduce the influence of these spallation neutrons by rejecting the high-multiplicity events. The measurement results, obtained with the background correction filter outlined in this paper, prove the repeatability and validity of the TIA-method and show that multiplicity counting with the TIA-technique is applicable for masses as low as 0.1 g 240 Pu eff even at a detection efficiency of 12%. (orig.)

ON THE MASS DISTRIBUTION AND BIRTH MASSES OF NEUTRON STARS

International Nuclear Information System (INIS)

Özel, Feryal; Psaltis, Dimitrios; Santos Villarreal, Antonio; Narayan, Ramesh

2012-01-01

We investigate the distribution of neutron star masses in different populations of binaries, employing Bayesian statistical techniques. In particular, we explore the differences in neutron star masses between sources that have experienced distinct evolutionary paths and accretion episodes. We find that the distribution of neutron star masses in non-recycled eclipsing high-mass binaries as well as of slow pulsars, which are all believed to be near their birth masses, has a mean of 1.28 M ☉ and a dispersion of 0.24 M ☉ . These values are consistent with expectations for neutron star formation in core-collapse supernovae. On the other hand, double neutron stars, which are also believed to be near their birth masses, have a much narrower mass distribution, peaking at 1.33 M ☉ , but with a dispersion of only 0.05 M ☉ . Such a small dispersion cannot easily be understood and perhaps points to a particular and rare formation channel. The mass distribution of neutron stars that have been recycled has a mean of 1.48 M ☉ and a dispersion of 0.2 M ☉ , consistent with the expectation that they have experienced extended mass accretion episodes. The fact that only a very small fraction of recycled neutron stars in the inferred distribution have masses that exceed ∼2 M ☉ suggests that only a few of these neutron stars cross the mass threshold to form low-mass black holes.

The effect of albedo neutrons on the neutron multiplication of small plutonium oxide samples in a PNCC chamber

CERN Document Server

Bourva, L C A; Weaver, D R

2002-01-01

This paper describes how to evaluate the effect of neutrons reflected from parts of a passive neutron coincidence chamber on the neutron leakage self-multiplication, M sub L , of a fissile sample. It is shown that albedo neutrons contribute, in the case of small plutonium bearing samples, to a significant part of M sub L , and that their effect has to be taken into account in the relationship between the measured coincidence count rates and the sup 2 sup 4 sup 0 Pu effective mass of the sample. A simple one-interaction model has been used to write the balance of neutron gains and losses in the material when exposed to the re-entrant neutron flux. The energy and intensity profiles of the re-entrant flux have been parameterised using Monte Carlo MCNP sup T sup M calculations. This technique has been implemented for the On Site Laboratory neutron/gamma counter within the existing MEPL 1.0 code for the determination of the neutron leakage self-multiplication. Benchmark tests of the resulting MEPL 2.0 code with MC...

Non-destructive isotopic uranium assay by multiple delayed neutron measurements

International Nuclear Information System (INIS)

Papadopoulos, N.N.; Tsagas, N.F.

1991-01-01

The high accuracy and precision required in nuclear safeguards measurements can be achieved by an improved neutron activation technique based on multiple delayed fission neutron counting under various experimental conditions. For the necessary ultrahigh counting statistics required, cyclic activation of multiple subsamples has been applied. The home-made automated flexible analytical system with neutron flux and spectrum differentiation by irradiation position adjustment and cadmium screening, permits the non-destructive determination of the U235 abundance and the total U element concentration needed in nuclear safeguards sample analysis, with a high throughout and a low operational cost. Careful experimental optimization led to considerable improvement of the results

Measurement of neutron multiplication in Pb by Mn foils

International Nuclear Information System (INIS)

Chen Yuan; Liu Rong; Guo Haiping; Jiang Wenmian; Shen Jian

1994-01-01

The Leakage neutron multiplication in bulk lead has been measured using the total absorption detector and relative method. The polyethylene sphere of 138 cm in diameter is used as the moderator and total absorption detector. The measured results from 55 Mn foils and 6 Li glass are compared. The neutron multiplication is 1.74 with the lead shell of 23.1 cm thick. The measured result is consistent with the calculated one with ANISN code and ENDF/B-6 evaluated data within the experimental error. (4 figs., 3 tabs.)

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

International Nuclear Information System (INIS)

Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

1994-01-01

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

Expected precision of neutron multiplicity measurements of waste drums

International Nuclear Information System (INIS)

Ensslin, N.; Krick, M.S.; Menlove, H.O.

1995-01-01

DOE facilities are beginning to apply passive neutron multiplicity counting techniques to the assay of plutonium scrap and residues. There is also considerable interest in applying this new measurement technique to 208-liter waste drums. The additional information available from multiplicity counting could flag the presence of shielding materials or improve assay accuracy by correcting for matrix effects such as (α,n) induced fission or detector efficiency variations. The potential for multiplicity analysis of waste drums, and the importance of better detector design, can be estimated by calculating the expected assay precision using a Figure of Merit code for assay variance. This paper reports results obtained as a function of waste drum content and detector characteristics. We find that multiplicity analysis of waste drums is feasible if a high-efficiency neutron counter is used. However, results are significantly poorer if the multiplicity analysis must be used to solve for detection efficiency

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

Energy Technology Data Exchange (ETDEWEB)

Costa da Silva, Adilson; Carvalho da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil); Senra Martinez, Aquilino, E-mail: aquilino@lmp.ufrj.br [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil)

2011-07-15

Highlights: > We proposed a new neutron diffusion hybrid equation with external neutron source. > A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. > 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

International Nuclear Information System (INIS)

Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

2011-01-01

Highlights: → We proposed a new neutron diffusion hybrid equation with external neutron source. → A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. → 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

Number distribution of leakage neutrons for single neutron emission event and one source emission event in multiplying medium for two variables - a GEANT4 study

International Nuclear Information System (INIS)

Roy, Arup Singha; Raman, Anand; Chaudhury, Probal; Thomas, Renju G.

2018-01-01

A quantitative knowledge about the neutron multiplying character of a neutron multiplying medium such as High enriched Uranium (HEU), Weapon Graded plutonium (WGPu) and similar special nuclear materials is essential for improving the probability of detection of these materials to check against illicit trafficking. The objective of this study is to gain a deeper insight in to the neutron and gamma multiplication behaviour of these materials. The leakage number distribution of neutron and gamma initiated by a source emission event (Spontaneous Fission) as well as single neutron emission event has been obtained in the course of this study. The computations for this study were carried out through GEANT4 simulation and also with the help of FREYA incorporated into it. This helped to carry out a detailed analysis of each history more realistically and obtain more reliable results

Neutron coincidence counting based on time interval analysis with dead time corrected one and two dimensional Rossi-alpha distributions: an application for passive neutron waste assay

International Nuclear Information System (INIS)

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

1996-03-01

The report describes a new neutron multiplicity counting method based on Rossi-alpha distributions. The report also gives the necessary dead time correction formulas for the multiplicity counting method. The method was tested numerically using a Monte Carlo simulation of pulse trains. The use of this multiplicity method in the field of waste assay is explained: it can be used to determine the amount of fissile material in a waste drum without prior knowledge of the actual detection efficiency

Neutron momentum distributions from ''core break-up'' reactions of halo nuclei

International Nuclear Information System (INIS)

Nilsson, T.; Blaich, T.; Borege, M.J.G.

1995-01-01

Neutron angular distributions from violent break-up reactions of 11 Li and 11 Be have been measured at 28 MeV/u and 280 MeV/u and at 41 MeV/u and 460 MeV/u, respectively. The derived neutron momentum distributions show a narrow component in transverse momentum that is within uncertainties independent of beam energy and target charge. This component is suggested to be simply related to the momentum distribution of the loosely bound halo neutron(s) in the projectiles. (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

Determination of shell correction energies at saddle point using pre-scission neutron multiplicities

International Nuclear Information System (INIS)

Golda, K.S.; Saxena, A.; Mittal, V.K.; Mahata, K.; Sugathan, P.; Jhingan, A.; Singh, V.; Sandal, R.; Goyal, S.; Gehlot, J.; Dhal, A.; Behera, B.R.; Bhowmik, R.K.; Kailas, S.

2013-01-01

Pre-scission neutron multiplicities have been measured for 12 C + 194, 198 Pt systems at matching excitation energies at near Coulomb barrier region. Statistical model analysis with a modified fission barrier and level density prescription have been carried out to fit the measured pre-scission neutron multiplicities and the available evaporation residue and fission cross sections simultaneously to constrain statistical model parameters. Simultaneous fitting of the pre-scission neutron multiplicities and cross section data requires shell correction at the saddle point

Device for measuring neutron-flux distribution density

International Nuclear Information System (INIS)

Rozenbljum, N.D.; Mitelman, M.G.; Kononovich, A.A.; Kirsanov, V.S.; Zagadkin, V.A.

1977-01-01

An arrangement is described for measuring the distribution of neutron flux density over the height of a nuclear reactor core and which may be used for monitoring energy release or for detecting deviations of neutron flux from an optimal level so that subsequent balance can be achieved. It avoids mutual interference of detectors. Full constructional details are given. (UK)

Multicounter neutron detector for examination of content and spatial distribution of fissile materials in bulk samples

International Nuclear Information System (INIS)

Swiderska-Kowalczyk, M.; Starosta, W.; Zoltowski, T.

1999-01-01

A new neutron coincidence well-counter is presented. This experimental device can be applied for passive assay of fissile and, in particular, for plutonium bearing materials. It contains of a set of the 3 He tubes placed inside a polyethylene moderator. Outputs from the tubes, first processed by preamplifier/amplifier/discriminator circuits, are then analysed using a correlator connected with PC, and correlation techniques implemented in software. Such a neutron counter enables determination of the 240 Pu effective mass in samples of a small Pu content (i.e., where the multiplication effects can be neglected) having a fairly big volume (up to 0.17 m 3 ), if only the isotopic composition is known. For determination of neutron sources distribution inside a sample, a heuristic method based on hierarchical cluster analysis was applied. As input parameters, amplitudes and phases of two-dimensional Fourier transformation of the count profiles matrices for known point sources distributions and for the examined samples were taken. Such matrices of profiles counts are collected using the sample scanning with detection head. In the clustering processes, process, counts profiles of unknown samples are fitted into dendrograms employing the 'proximity' criterion of the examined sample profile to standard samples profiles. Distribution of neutron sources in the examined sample is then evaluated on the basis of a comparison with standard sources distributions. (author)

A linear multiple balance method for discrete ordinates neutron transport equations

International Nuclear Information System (INIS)

Park, Chang Je; Cho, Nam Zin

2000-01-01

A linear multiple balance method (LMB) is developed to provide more accurate and positive solutions for the discrete ordinates neutron transport equations. In this multiple balance approach, one mesh cell is divided into two subcells with quadratic approximation of angular flux distribution. Four multiple balance equations are used to relate center angular flux with average angular flux by Simpson's rule. From the analysis of spatial truncation error, the accuracy of the linear multiple balance scheme is ο(Δ 4 ) whereas that of diamond differencing is ο(Δ 2 ). To accelerate the linear multiple balance method, we also describe a simplified additive angular dependent rebalance factor scheme which combines a modified boundary projection acceleration scheme and the angular dependent rebalance factor acceleration schme. It is demonstrated, via fourier analysis of a simple model problem as well as numerical calculations, that the additive angular dependent rebalance factor acceleration scheme is unconditionally stable with spectral radius < 0.2069c (c being the scattering ration). The numerical results tested so far on slab-geometry discrete ordinates transport problems show that the solution method of linear multiple balance is effective and sufficiently efficient

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

International Nuclear Information System (INIS)

Matsumoto, Tetsuo; Aizawa, Otohiko; Nozaki, Tetsuya

1981-01-01

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

Spatial distribution of neutron flux for the A-711 neutron generator

International Nuclear Information System (INIS)

Essiet, A. E.; Owolabi, S. A.; Adesanmi, C. A.; Balogun, F. A.

1996-01-01

The spatial distribution of neutron flux for the Kaman sciences A-711 neutron generator recently installed at the Centre for Energy Research and Development (CERD), Ile-Ife Nigeria has been determined. At an operational tube current of 2.0 mA and high voltage power supply (HVPS) of 158 kV, the neutron flux increases from 1.608 ± 0.021*10 8 n/cm 2 s at the top of the irradiated plastic vial to 2.640 ± 0.022*10 8 n/cm 2 s at the centre, and then decreases to 1.943 ± 0.02* 8 n/cm 2 s at the bottom. The flux density is strongly dependent on the diameter of deuteron at the tritium target, and within this range a source strength of 10 8 n/s has been measured for the A-711 neutron generator

Reducing neutron multiplicity counting bias for plutonium warhead authentication

Energy Technology Data Exchange (ETDEWEB)

Goettsche, Malte

2015-06-05

Confidence in future nuclear arms control agreements could be enhanced by direct verification of warheads. It would include warhead authentication. This is the assessment based on measurements whether a declaration that a specific item is a nuclear warhead is true. An information barrier can be used to protect sensitive information during measurements. It could for example show whether attributes such as a fissile mass exceeding a threshold are met without indicating detailed measurement results. Neutron multiplicity measurements would be able to assess a plutonium fissile mass attribute if it were possible to show that their bias is low. Plutonium measurements have been conducted with the He-3 based Passive Scrap Multiplicity Counter. The measurement data has been used as a reference to test the capacity of the Monte Carlo code MCNPX-PoliMi to simulate neutron multiplicity measurements. The simulation results with their uncertainties are in agreement with the experimental results. It is essential to use cross-sections which include neutron scattering with the detector's polyethylene molecular structure. Further MCNPX-PoliMi simulations have been conducted in order to study bias that occurs when measuring samples with large plutonium masses such as warheads. Simulation results of solid and hollow metal spheres up to 6000 g show that the masses are underpredicted by as much as 20%. The main source of this bias has been identified in the false assumption that the neutron multiplication does not depend on the position where a spontaneous fission event occurred. The multiplication refers to the total number of neutrons leaking a sample after a primary spontaneous fission event, taking induced fission into consideration. The correction of the analysis has been derived and implemented in a MATLAB code. It depends on four geometry-dependent correction coefficients. When the sample configuration is fully known, these can be exactly determined and remove this type of

Gamma spectroscopic studies of the neutron-deficient g-g nucleus 74Kr by means of a neutron multiplicity measurement technique

International Nuclear Information System (INIS)

Roth, J.

1981-01-01

The g-g nucleus 74 Kr was studied by means of the reaction 58 Ni ( 19 F, p2n#betta#) 74 Kr. In order to make gamma spectroscopic studies at neutron deficient nuclei like 74 Kr a neutron multiplicity measurement technique was developed. Beside #betta# single spectra, #betta# excitation functions, #betta#-#betta# coincidences, #betta# angular distributions, and lifetime measurements by means of this technique all measurements in coincidence with up to two neutrons were taken up. From these measurement data an extended term scheme with 17 newly found excited states could be extracted. To all levels spins and parities could be assigned. From the four energetically lowest levels of the yrast cascade the mean lifetimes could be determined. A double backbending in the sequence of the yrast cascade was interpreted as crossing of the g 9/2 bands. The irregularities in the lower part of the yrast band correspond to the shape consistence picture. The results were considered in connection with the systematics of the even krypton isotopes and compared with a two-quasiparticle-plas-rotor model calculation. (HSI)

Preliminary performance analysis of exponential experimental system for the determination of neutron effective multiplication factor of PWR spent fuel

International Nuclear Information System (INIS)

Shin, Heesung; Lee, Sang-Yun; Ro, Seung-Gy; Seo, Gi-Seok; Kim, Ho-Dong

2002-01-01

An exponential experiment system which is composed of neutron detector, signal analysis system and neutron source, 10 mCi Cf-252 has been installed in the storage pool of PIEF at KAERI in order to experimentally determining neutron effective multiplication factors of PWR spent fuel assemblies. Preliminary functional characteristic tests of the experimental system are performed for C15, J14 and J44 assemblies loaded in the pool. As a result of preliminary tests, the average neutron counts obtained for 3 minutes in the plateau of the C15, J14 and J44 assemblies are about 1900, 3800 and 3200, respectively. A dip of the neutron flux density distribution is noticed in the spacer grid position. Neutron counts at those positions appear to be reduced to about 70 % in comparison to the fuel position. The measured axial neutron distribution shapes are compared with the result for the P14 assembly and Cs-137 gamma scanning data performed in KAERI. It is revealed that the spacer grid position measured is consistent with the design specifications within a 2.3 % error. The exponential decay constants for the C15 assembly were determined to be 0.152 and 0.165 for detector and source scanning, respectively. (author)

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

Measurements of 14 MeV neutron multiplication in spherical beryllium shells

International Nuclear Information System (INIS)

Moellendorff, U. von; Alevra, A.V.; Giese, H.; Kappler, F.; Klein, H.; Klein, H.; Tayama, R.

1995-01-01

New results of spherical-shell transmission measurements with 14MeV neutrons on pure beryllium shells up to 17cm thick are reported. The total leakage neutron multiplications were measured using a Bonner sphere system. Independently, the leakage neutron spectra were measured over the entire energy range, 15MeV to thermal energies, by proton-recoil and time-of-flight methods. The total leakage multiplications are in excellent agreement with three-dimensional Monte Carlo calculations using beryllium nuclear data based on the Young and Stewart evaluation. The leakage in the evaporation energy window confirms the Be(n,2n) cross-section of the Young and Stewart evaluation rather than that used in the ENDF/B-VI library. At energies below 1keV, a surplus of leakage neutrons over the calculation is found for smaller beryllium thicknesses. (orig.)

Non-destructive assay of fissile materials by detection and multiplicity analysis of spontaneous neutrons

International Nuclear Information System (INIS)

Prosdocimi, A.

1979-01-01

A method for determining the absolute reaction rate of nuclear events giving rise to neutron emission, according to their neutron multiplicity, is proposed. A typical application is the measurement of the (α, n) and spontaneous fission rates in a fissile material sample, particularly of Pu oxide composition. An analysis of random and correlated neutron pulses is carried out on the basis of sequential order without requiring any time interval analysis, then the primary nuclear events are sorted versus their neutron multiplicity. Suitable theoretical relationships enable to derive the absolute (α, n) and SF reaction rates when the physical parameters of the neutron detector and the multiplicity spectrumm of pulses are known. A typical device is described and the results of experiments leading to Pu-239 and Pu-240 assay are given

Variable dead time counters. 1 - theoretical responses and the effects of neutron multiplication

International Nuclear Information System (INIS)

Lees, E.W.; Hooton, B.W.

1978-10-01

A theoretical expression is derived for calculating the response of any variable dead time counter (VDC) used in the passive assay of plutonium by neutron counting of the natural spontaneous fission activity. The effects of neutron multiplication in the sample arising from interactions of the original spontaneous fission neutrons is shown to modify the linear relationship between VDC signal and Pu mass. Numerical examples are shown for the Euratom VDC and a systematic investigation of the various factors affecting neutron multiplication is reported. Limited comparisons between the calculations and experimental data indicate provisional validity of the calculations. (author)

Coulomb displacement energies and neutron density distributions

International Nuclear Information System (INIS)

Shlomo, S.

1979-01-01

We present a short review of the present status of the theory of Coulomb displacement energies, ΔEsub(c), discussing the Okamoto-Nolem-Schiffer anomaly and its solution. We emphasize, in particular, that contrary to previous hopes, ΔEsub(c) does not determine rsub(ex), the root-mean square (rms) radius of the excess (valence) neutron density distribution. Instead, ΔEsub(c) is very sensitive to the value of Δr = rsub(n) - rsub(p), the difference between the rms radii of the density distributions of all neutrons and all protons. For neutron rich nuclei, such as 48 Ca and 208 Pb, a value of Δr = 0.1 fm is found to be consistent with ΔEsub(c). This value of Δr, which is considerably smaller than that (of 0.2 - 0.3 fm) predicted by some common Hartree-Fock calculations, seems to be confirmed by very recent experimental results. (orig.)

Orion, a high efficiency 4π neutron detector

International Nuclear Information System (INIS)

Crema, E.; Piasecki, E.; Wang, X.M.; Doubre, H.; Galin, J.; Guerreau, D.; Pouthas, J.; Saint-Laurent, F.

1990-01-01

In intermediate energy heavy ion collisions the multiplicity of emitted neutrons is strongly connected to energy dissipation and to impact parameter. We present the 4π detector ORION, a high efficiency liquid scintillator detector which permits to get information on the multiplicity of neutrons measured event-wise and on the spatial distribution of these neutrons [fr

Study of gamma ray multiplicity spectra for radiative capture of neutrons in 113,115In

International Nuclear Information System (INIS)

Georgiev, G.P.; Fajkov-Stanchik, Kh.; Grigor'ev, Yu.V.; Muradyan, G.V.; Yaneva, N.B.

1997-08-01

Neutron radiative capture measurements were performed for the enriched isotopes 113 In and 115 In on the neutron spectrometer at the Neutron Physics Laboratory of the Joint Institute for Nuclear Research employing the gamma ray multiplicity technique and using a ''Romashka'' multi-sectional 4p detector on the 500 m time base of the IBR-30 booster. The gamma multiplicity spectra of resolved resonances were obtained for the 20-500 eV energy range. The mean gamma ray multiplicity was determined for each resonance. The dependence of the ratio S of the low-energy coincidence multiplicity spectrum to the high-energy coincidence multiplicity spectrum on resonance energy exhibits a non-statistical structure. This structure was found to correlate with the local neutron strength function. (author). 10 refs, 6 figs, 2 tabs

Measurable distributions of unpolarized neutron decay

International Nuclear Information System (INIS)

Glueck, F.

1992-05-01

Several two- and one-dimensional distributions of unpolarized free neutron decay are calculated. The results of the order-α model independent radiative correction calculations are tabulated numerically. With these corrections the theoretical distributions become precise enough to make possible the determination of the ratio of the axial-vector to the vector weak coupling constants to a precision of ∼0.001. (author) 39 refs.; 7 tabs

Time dependent and asymptotic neutron number probability distribution calculation using discrete Fourier transform

International Nuclear Information System (INIS)

Humbert, Ph.

2005-01-01

In this paper we consider the probability distribution of neutrons in a multiplying assembly. The problem is studied using a space independent one group neutron point reactor model without delayed neutrons. We recall the generating function methodology and analytical results obtained by G.I. Bell when the c 2 approximation is used and we present numerical solutions in the general case, without this approximation. The neutron source induced distribution is calculated using the single initial neutron distribution which satisfies a master (Kolmogorov backward) equation. This equation is solved using the generating function method. The generating function satisfies a differential equation and the probability distribution is derived by inversion of the generating function. Numerical results are obtained using the same methodology where the generating function is the Fourier transform of the probability distribution. Discrete Fourier transforms are used to calculate the discrete time dependent distributions and continuous Fourier transforms are used to calculate the asymptotic continuous probability distributions. Numerical applications are presented to illustrate the method. (author)

Characterisation of neutron fields: challenges in assessing the directional distribution

International Nuclear Information System (INIS)

Cauwels, Vanessa; Vanhavere, Filip; Reginatto, Marcel

2014-01-01

The SCK.CEN has carried out neutron field characterisation campaigns at several nuclear reactors. The main goal of these measurement campaigns was to evaluate the performance of different neutron personal dosemeters. To be able to evaluate the performance of neutron personal dosemeters in terms of H p (10), knowledge of the directional distribution is indispensable. This distribution was estimated by placing several personal dosemeters on all six sides of a slab phantom. The interpretation and conversion of this information into a reliable value for H p (10) requires great care. The data were analysed using three methods. In the first approach, a linear interpolation was performed on three perpendicular axes. In the other two approaches, an icosahedron was used to model the angle of incidence of the neutrons and a linear interpolation or a Bayesian analysis was performed. This study describes the limitations and advantages of each of these methods and provides recommendations for their use to estimate the personal dose equivalent H p (10) for neutron dosimetry. Neutron personal dosimetry is complicated by the fact that the neutron dose quantity H p (10) is strongly energy and angular dependent. Instead of simply assuming that the fluence is unidirectional or that the fluence is isotropic, an attempt was made to estimate the directional distribution of the neutron field using a relatively simple measurement procedure. A number of active and passive personal dosemeters were placed on the six faces of a slab phantom and the results were analysed via different algorithms to obtain partial fluences in several directions of incidence. The results from all calculations in this study show the importance of introducing information about the directional distribution of the neutron fluence for the estimation of the personal dose equivalent H p (10). The difference between H p (10) dose estimates carried out using a unidirectional or an isotropic distribution can be of up

The MARVEL assembly for neutron multiplication

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester; Mathew T. Kinlaw

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (keff=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication.

The MARVEL assembly for neutron multiplication.

Science.gov (United States)

Chichester, David L; Kinlaw, Mathew T

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (k(eff)=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication. Copyright © 2013 Elsevier Ltd. All rights reserved.

$\\gamma$-ray energy spectra and multiplicities from the neutron-induced fission of $^{235}$U using STEFF

CERN Document Server

An experiment is proposed to use the STEFF spectrometer at n_TOF to study fragment $\\gamma$-correlations following the neutron-induced fission of $^{235}$U. The STEFF array of 12 NaI detectors will allow measurements of the single $\\gamma$-energy, the $\\gamma$ multiplicity, and the summed $\\gamma$energy distributions as a function of the mass and charge split, and deduced excitation energy in the fission event. These data will be used to study the origin of fission-fragment angular momenta, examining angular distribution eects as a function of incident neutron energy. The principal application of this work is in meeting the NEA high-priority request for improved $\\gamma$ray data from $^{235}$U(n; F). To improve the detection rate and expand the range of detection angles, STEFF will be modied to include two new ssion-fragment detectors each at 45 to the beam direction.

The effect of temperature and the control rod position on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

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

2007-01-01

The effect of water and fuel temperature increase and changes in the control rod positions on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor (MNSR) is discussed. The cross sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the special neutron flux distribution using four energy groups. This work shows that water and fuel temperature increase in the reactor during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. Changing the control rod position does not affect as well the spatial neutron flux distribution except in the region around the control rod position. This stability in the spatial neutron flux distribution, especially in the inner and outer irradiation sites, makes MNSR as a good tool for the neutron activation analysis (NAA) technique and production of radioisotopes with medium or short half lives during the reactor daily operating time. (author)

Geometry-based multiplication correction for passive neutron coincidence assay of materials with variable and unknown (α,n) neutron rates

International Nuclear Information System (INIS)

Langner, D.G.; Russo, P.A.

1993-02-01

We have studied the problem of assaying impure plutonium-bearing materials using passive neutron coincidence counting. We have developed a technique to analyze neutron coincidence data from impure plutonium samples that uses the bulk geometry of the sample to correct for multiplication in samples for which the (α,n) neutron production rate is unknown. This technique can be applied to any impure plutonium-bearing material whose matrix constituents are approximately constant, whose self-multiplication is low to moderate, whose plutonium isotopic composition is known and not substantially varying, and whose bulk geometry is measurable or can be derived. This technique requires a set of reference materials that have well-characterized plutonium contents. These reference materials are measured once to derive a calibration that is specific to the neutron detector and the material. The technique has been applied to molten salt extraction residues, PuF 4 samples that have a variable salt matrix, and impure plutonium oxide samples. It is also applied to pure plutonium oxide samples for comparison. Assays accurate to 4% (1 σ) were obtained for impure samples measured in a High-Level Neutron Coincidence Counter II. The effects on the technique of variations in neutron detector efficiency with energy and the effects of neutron capture in the sample are discussed

Some problems of neutron source multiplication method for site measurement technology in nuclear critical safety

International Nuclear Information System (INIS)

Shi Yongqian; Zhu Qingfu; Hu Dingsheng; He Tao; Yao Shigui; Lin Shenghuo

2004-01-01

The paper gives experiment theory and experiment method of neutron source multiplication method for site measurement technology in the nuclear critical safety. The measured parameter by source multiplication method actually is a sub-critical with source neutron effective multiplication factor k s , but not the neutron effective multiplication factor k eff . The experiment research has been done on the uranium solution nuclear critical safety experiment assembly. The k s of different sub-criticality is measured by neutron source multiplication experiment method, and k eff of different sub-criticality, the reactivity coefficient of unit solution level, is first measured by period method, and then multiplied by difference of critical solution level and sub-critical solution level and obtained the reactivity of sub-critical solution level. The k eff finally can be extracted from reactivity formula. The effect on the nuclear critical safety and different between k eff and k s are discussed

Calculation of contribution of multiple interactions and efficiency of neutron detectors

International Nuclear Information System (INIS)

Androsenko, A.A.; Androsenko, P.A.; Kazakov, L.E.; Kononov, V.N.; Poletaev, E.D.

1986-01-01

Results of calculation of multiple neutron interactions contribution to efficiency of detectors with 6 Li glass and 10 B plate in the energy range of 0.01-1 MeV are given. The calculation was performed by the Monte-Carlo method using BRAND program complex. It is shown that a correction value for multiple neutron interaction in 6 Li glass of 1 mm thickness constitutes 4.5 % at energy of up to 100 keV and at higher energies has a complex energy dependence reaching 25 % at 440 keV

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

Science.gov (United States)

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

2018-01-01

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

Thermal-neutron multiple scattering: critical double scattering

International Nuclear Information System (INIS)

Holm, W.A.

1976-01-01

A quantum mechanical formulation for multiple scattering of thermal-neutrons from macroscopic targets is presented and applied to single and double scattering. Critical nuclear scattering from liquids and critical magnetic scattering from ferromagnets are treated in detail in the quasielastic approximation for target systems slightly above their critical points. Numerical estimates are made of the double scattering contribution to the critical magnetic cross section using relevant parameters from actual experiments performed on various ferromagnets. The effect is to alter the usual Lorentzian line shape dependence on neutron wave vector transfer. Comparison with corresponding deviations in line shape resulting from the use of Fisher's modified form of the Ornstein-Zernike spin correlations within the framework of single scattering theory leads to values for the critical exponent eta of the modified correlations which reproduce the effect of double scattering. In addition, it is shown that by restricting the range of applicability of the multiple scattering theory from the outset to critical scattering, Glauber's high energy approximation can be used to provide a much simpler and more powerful description of multiple scattering effects. When sufficiently close to the critical point, it provides a closed form expression for the differential cross section which includes all orders of scattering and has the same form as the single scattering cross section with a modified exponent for the wave vector transfer

MPACT Fast Neutron Multiplicity System Prototype Development

Energy Technology Data Exchange (ETDEWEB)

D.L. Chichester; S.A. Pozzi; J.L. Dolan; M.T. Kinlaw; S.J. Thompson; A.C. Kaplan; M. Flaska; A. Enqvist; J.T. Johnson; S.M. Watson

2013-09-01

This document serves as both an FY2103 End-of-Year and End-of-Project report on efforts that resulted in the design of a prototype fast neutron multiplicity counter leveraged upon the findings of previous project efforts. The prototype design includes 32 liquid scintillator detectors with cubic volumes 7.62 cm in dimension configured into 4 stacked rings of 8 detectors. Detector signal collection for the system is handled with a pair of Struck Innovative Systeme 16-channel digitizers controlled by in-house developed software with built-in multiplicity analysis algorithms. Initial testing and familiarization of the currently obtained prototype components is underway, however full prototype construction is required for further optimization. Monte Carlo models of the prototype system were performed to estimate die-away and efficiency values. Analysis of these models resulted in the development of a software package capable of determining the effects of nearest-neighbor rejection methods for elimination of detector cross talk. A parameter study was performed using previously developed analytical methods for the estimation of assay mass variance for use as a figure-of-merit for system performance. A software package was developed to automate these calculations and ensure accuracy. The results of the parameter study show that the prototype fast neutron multiplicity counter design is very nearly optimized under the restraints of the parameter space.

Measurements of fusion neutron multiplication in spherical beryllium shells

International Nuclear Information System (INIS)

Giese, H.; Kappler, F.; Tayama, R.; Moellendorff, U. von; Alevra, A.; Klein, H.

1996-01-01

New results of spherical-shell transmission measurements with 14-MeV neutrons on pure beryllium shells up to 17 cm thick are reported. The spectral flux above 3 MeV was measured using a liquid scintillation detector. At 17 cm thickness, also the total neutron multiplication was measured using a Bonner sphere system. The results agree well with calculations using beryllium nuclear data from the EFF-1 or the ENDF/B-Vi library. (author). 23 refs, 4 figs, 1 tab

Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

International Nuclear Information System (INIS)

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

2009-01-01

A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

A neutron multiplicity analysis method for uranium samples with liquid scintillators

Energy Technology Data Exchange (ETDEWEB)

Zhou, Hao, E-mail: zhouhao_ciae@126.com [China Institute of Atomic Energy, P.O.BOX 275-8, Beijing 102413 (China); Lin, Hongtao [Xi' an Reasearch Institute of High-tech, Xi' an, Shaanxi 710025 (China); Liu, Guorong; Li, Jinghuai; Liang, Qinglei; Zhao, Yonggang [China Institute of Atomic Energy, P.O.BOX 275-8, Beijing 102413 (China)

2015-10-11

A new neutron multiplicity analysis method for uranium samples with liquid scintillators is introduced. An active well-type fast neutron multiplicity counter has been built, which consists of four BC501A liquid scintillators, a n/γdiscrimination module MPD-4, a multi-stop time to digital convertor MCS6A, and two Am–Li sources. A mathematical model is built to symbolize the detection processes of fission neutrons. Based on this model, equations in the form of R=F*P*Q*T could be achieved, where F indicates the induced fission rate by interrogation sources, P indicates the transfer matrix determined by multiplication process, Q indicates the transfer matrix determined by detection efficiency, T indicates the transfer matrix determined by signal recording process and crosstalk in the counter. Unknown parameters about the item are determined by the solutions of the equations. A {sup 252}Cf source and some low enriched uranium items have been measured. The feasibility of the method is proven by its application to the data analysis of the experiments.

On radii of neutron distributions in nuclei

International Nuclear Information System (INIS)

Varma, G.K.; Zamick, L.

1978-01-01

The analyses of the differences between rms radii (Δ=rsub(n) - rsub(p) of neutron and proton distributions are considered in a wide variety of nuclei. It is noted that apart from its own intrinsic interest, the quantity Δ is of importance for isotope shifts, core polarization contributions to the Coulomb energy difference of mirror pairs (Nolen-Schiffer anomaly) and the renormalization of the effective interaction. For example, if Δ were very small in 48 Ca then the Nolen-Schiffer anomaly could be explained by a core polarization mechanism. The various methods of determining Δ are considered critically and it is concluded that at present probably the most reliable method is high energy (approximately 1 GeV) proton-nucleon scattering. The different theoretical analyses based upon, e.g. the multiple diffraction theory (where Glauber amplitude is the leading term) or the optical potential (KMT) formalisms appear to be converging to essentially the same answer when analyzing the same data. High energy α-particles and medium energy pions can also become useful sources of information if higher order optical potentials are treated with care. It is found that Δ is rather large in 48 Ca, i.e. there is a neutron skin, so that the Nolen-Schiffer anomaly cannot be explained by a core polarization mechanism. The results of high energy proton-nucleus scattering are in excellent agreement with current density dependent Hartree-Fock calculation. (Auth.)

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

International Nuclear Information System (INIS)

Tian Dongfeng; Ho Yukun; Yang Fujia

2001-01-01

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

Neutron emission effects on fragment mass and kinetic energy distribution from fission of 239Pu induced by thermal neutrons

International Nuclear Information System (INIS)

Montoya, M.; Rojas, J.; Lobato, I.

2010-01-01

The average of fragment kinetic energy (E-bar sign*) and the multiplicity of prompt neutrons (ν(bar sign)) as a function of fragment mass (m*), as well as the fragment mass yield (Y(m*)) from thermal neutron-induced fission of 239 Pu have been measured by Tsuchiya et al.. In that work the mass and kinetic energy are calculated from the measured kinetic energy of one fragment and the difference of time of flight of the two complementary fragments. However they do not present their results about the standard deviation σ E *(m*). In this work we have made a numerical simulation of that experiment which reproduces its results, assuming an initial distribution of the primary fragment kinetic energy (E(A)) with a constant value of the standard deviation as function of fragment mass (σ E (A)). As a result of the simulation we obtain the dependence σ E *(m*) which presents an enhancement between m* = 92 and m* = 110, and a peak at m* = 121.

Phantom experiment of depth-dose distributions for gadolinium neutron capture therapy

International Nuclear Information System (INIS)

Matsumoto, T.; Kato, K.; Sakuma, Y.; Tsuruno, A.; Matsubayashi, M.

1993-01-01

Depth-dose distributions in a tumor simulated phantom were measured for thermal neutron flux, capture gamma-ray and internal conversion electron dose rates for gadolinium neutron capture therapy. The results show that (i) a significant dose enhancement can be achieved in the tumor by capture gamma-rays and internal conversion electrons but the dose is mainly due to capture gamma-rays from the Gd(n, γ) reactions, therefore, is not selective at the cellular level, (ii) the dose distribution was a function of strongly interrelated parameters such as gadolinium concentrations, tumor site and neutron beam size (collimator aperture size), and (iii) the Gd-NCT by thermal neutrons appears to be a potential for treatment of superficial tumor. (author)

Charge collection and space charge distribution in neutron-irradiated epitaxial silicon detectors

International Nuclear Information System (INIS)

Poehlsen, Thomas

2010-04-01

In this work epitaxial n-type silicon diodes with a thickness of 100 μm and 150 μm are investigated. After neutron irradiation with fluences between 10 14 cm -2 and 4 x 10 15 cm -2 annealing studies were performed. CV-IV curves were taken and the depletion voltage was determined for different annealing times. All investigated diodes with neutron fluences greater than 2 x 10 14 cm -2 showed type inversion due to irradiation. Measurements with the transient current technique (TCT) using a pulsed laser were performed to investigate charge collection effects for temperatures of -40 C, -10 C and 20 C. The charge correction method was used to determine the effective trapping time τ eff . Inconsistencies of the results could be explained by assuming field dependent trapping times. A simulation of charge collection could be used to determine the field dependent trapping time τ eff (E) and the space charge distribution in the detector bulk. Assuming a linear field dependence of the trapping times and a linear space charge distribution the data could be described. Indications of charge multiplication were seen in the irradiated 100 μm thick diodes for all investigated fluences at voltages above 800 V. The space charge distribution extracted from TCT measurements was compared to the results of the CV measurements and showed good agreement. (orig.)

Fast neutron fluxes distribution in Egyptian ilmenite concrete

International Nuclear Information System (INIS)

Megahed, R.M.; Abou El-Nasr, T.Z.; Bashter, I.I.

1978-01-01

This work is concerned with the study of the distribution of fast neutron fluxes in a new type of heavy concrete made from Egyptian ilmenite ores. The neutron source used was a collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor. Measurements were carried-out using phosphorous activation detectors. Iso-flux curves were represented which give directly the shape and thickness required to attenuate the emitted fast neutron flux to a certain value. The relaxation lengths were also evaluated from the measured data for both disc monodirectional source and infinite plane monodirectional source. The obtained values were compared with that calculated using the derived values of relative number densities and microscopic removal cross-sections of the different constituents. The obtained data show that ilmenite concrete attenuates fast neutron flux more strongly than ordinary concrete. A semiemperical formula was derived to calculate the fast neutron flux at different thicknesses along the beam axis. Another semiemperical formula was also derived to calculate the fast neutron flux in ordinary concrete along the beam axis using the corresponding value in ilmenite concrete

Fast Neutron Dose Distribution in a Linac Radiotherapy Facility

International Nuclear Information System (INIS)

Al-Othmany, D.Sh.; Abdul-Majid, S.; Kadi, M.W.

2011-01-01

CR-39 plastic detectors were used for fast neutron dose mapping in the radiotherapy facility at King AbdulAziz University Hospital (KAUH). Detectors were calibrated using a 252 Cf neutron source and a neutron dosimeter. After exposure chemical etching was performed using 6N NaOH solution at 70 degree C. Tracks were counted using an optical microscope and the number of tracks/cm 2 was converted to a neutron dose. 15 track detectors were distributed inside and outside the therapy room and were left for 32 days. The average neutron doses were 142.3 mSv on the accelerator head, 28.5 mSv on inside walls, 1.4 mSv beyond the beam shield, and 1 mSv in the control room

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

International Nuclear Information System (INIS)

Sohn, Hee Dong

2012-02-01

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

Energy Technology Data Exchange (ETDEWEB)

Sohn, Hee Dong

2012-02-15

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Non-equilibrium thermodynamics and energy distribution function of neutron gas in constant power reactor under coupling of neutrons and medium

International Nuclear Information System (INIS)

Hayasaka, Hideo

1983-01-01

The thermodynamics and the energy distribution function of the neutron gas in a constant power reactor are considered, taking into account the burn-up of fuel. To separate the secular motion of neutrons owing to fuel burn-up and the microscopic fluctuations of neutrons around this motion, a long time of the order of several months is divided into m equal intervals, and the respective states corresponding to m small time intervals are treated as quasi-stationary states. The local energy distribution function of the neutron gas in the quasi-stationary state is given by a generalized Boltzmann distribution specified by the respective generalized activity coefficient for each subsystem. The effects of fuel burn-up on the respective distribution functions for successive small time intervals are taken into account through various quantities relating to reactor physics, depending upon the fuel burn-up, by successive approximation. (author)

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.)

The space distribution of neutrons generated in massive lead target by relativistic nuclear beam

International Nuclear Information System (INIS)

Chultem, D.; Damdinsuren, Ts.; Enkh-Gin, L.; Lomova, L.; Perelygin, V.; Tolstov, K.

1993-01-01

The present paper is devoted to implementation of solid state nuclear track detectors in the research of the neutron generation in extended lead spallation target. Measured neutrons space distribution inside the lead target and neutron distribution in the thick water moderator are assessed. (Author)

On angular distribution of nucleus fission fragments by fast neutrons

International Nuclear Information System (INIS)

Barabanov, A.L.; Grechukhin, D.P.

1987-01-01

Evaluation of amplitudes of quadrupole and hexadecapole components of angular distribution of nucleus fission fragments by neutrons with the energies E n < or approx. 6 MeV is conducted. Stability of this amplitude to permeability optical coefficient variations for neutrons is revealed. It is shown, that the ratio of these amplitudes as well as the character of their dependence on the target nucleus orientation degree are sensitive to the type of fission probability distribution along K projection if fissile nucleus J spin to the fragment scattering axis. This sensitivity may be used for fragment angular distribution anisotropy formation statistical model verification

Calculating the mass distribution of heavy nucleus fission product by neutrons

International Nuclear Information System (INIS)

Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.

1981-01-01

The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru

Comparison of MCNP6 and experimental results for neutron counts, Rossi-α, and Feynman-α distributions

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Sadovich, S.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.

2013-01-01

MCNP6, the general-purpose Monte Carlo N-Particle code, has the capability to perform time-dependent calculations by tracking the time interval between successive events of the neutron random walk. In fixed-source calculations for a subcritical assembly, the zero time value is assigned at the moment the neutron is emitted by the external neutron source. The PTRAC and F8 cards of MCNP allow to tally the time when a neutron is captured by 3 He(n, p) reactions in the neutron detector. From this information, it is possible to build three different time distributions: neutron counts, Rossi-α, and Feynman-α. The neutron counts time distribution represents the number of neutrons captured as a function of time. The Rossi-a distribution represents the number of neutron pairs captured as a function of the time interval between two capture events. The Feynman-a distribution represents the variance-to-mean ratio, minus one, of the neutron counts array as a function of a fixed time interval. The MCNP6 results for these three time distributions have been compared with the experimental data of the YALINA Thermal facility and have been found to be in quite good agreement. (authors)

From neutron Compton profiles to momentum distribution: Assessment of direct numerical determination

International Nuclear Information System (INIS)

Senesi, R.; Flammini, D.; Romanelli, G.; Andreani, C.

2013-01-01

Inelastic neutron scattering at high momentum transfers, in the neutron Compton scattering regime, provides an access to the neutron Compton profiles, the analogous of Compton profiles in X-ray scattering. The line shape analysis of the neutron Compton profiles is usually carried out making use of multiparametric nonlinear fitting, garnering detailed information about the momentum distribution of the target atoms. This paper presents the proposal to directly determine numerically the momentum distribution from the profiles, thus eliminating the possible instabilities present in multiparametric fitting. A comparison with Monte Carlo simulations and with previous measurements on polycrystalline ice provides quantitative assessments of the proposed method

Observation of multiple Bragg reflections of neutrons in bent perfect crystals

Czech Academy of Sciences Publication Activity Database

Mikula, Pavol; Vrána, Miroslav; Šaroun, Jan; Seong, B. S.; Moon, MK.

2011-01-01

Roč. 634, č. 1 (2011), S108-S111 ISSN 0168-9002. [International Workshop on Neutron Optics. Grenoble, 17.03.2010-19.03.2010] R&D Projects: GA ČR GAP204/10/0654 Institutional research plan: CEZ:AV0Z10480505 Keywords : Neutron diffraction * Bent perfect crystal * Multiple reflections Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.207, year: 2011

Improving neutron multiplicity counting for the spatial dependence of multiplication: Results for spherical plutonium samples

Energy Technology Data Exchange (ETDEWEB)

Göttsche, Malte, E-mail: malte.goettsche@physik.uni-hamburg.de; Kirchner, Gerald

2015-10-21

The fissile mass deduced from a neutron multiplicity counting measurement of high mass dense items is underestimated if the spatial dependence of the multiplication is not taken into account. It is shown that an appropriate physics-based correction successfully removes the bias. It depends on four correction coefficients which can only be exactly determined if the sample geometry and composition are known. In some cases, for example in warhead authentication, available information on the sample will be very limited. MCNPX-PoliMi simulations have been performed to obtain the correction coefficients for a range of spherical plutonium metal geometries, with and without polyethylene reflection placed around the spheres. For hollow spheres, the analysis shows that the correction coefficients can be approximated with high accuracy as a function of the sphere's thickness depending only slightly on the radius. If the thickness remains unknown, less accurate estimates of the correction coefficients can be obtained from the neutron multiplication. The influence of isotopic composition is limited. The correction coefficients become somewhat smaller when reflection is present.

A method to describe inelastic gamma field distribution in neutron gamma density logging.

Science.gov (United States)

Zhang, Feng; Zhang, Quanying; Liu, Juntao; Wang, Xinguang; Wu, He; Jia, Wenbao; Ti, Yongzhou; Qiu, Fei; Zhang, Xiaoyang

2017-11-01

Pulsed neutron gamma density logging (NGD) is of great significance for radioprotection and density measurement in LWD, however, the current methods have difficulty in quantitative calculation and single factor analysis for the inelastic gamma field distribution. In order to clarify the NGD mechanism, a new method is developed to describe the inelastic gamma field distribution. Based on the fast-neutron scattering and gamma attenuation, the inelastic gamma field distribution is characterized by the inelastic scattering cross section, fast-neutron scattering free path, formation density and other parameters. And the contribution of formation parameters on the field distribution is quantitatively analyzed. The results shows the contribution of density attenuation is opposite to that of inelastic scattering cross section and fast-neutron scattering free path. And as the detector-spacing increases, the density attenuation gradually plays a dominant role in the gamma field distribution, which means large detector-spacing is more favorable for the density measurement. Besides, the relationship of density sensitivity and detector spacing was studied according to this gamma field distribution, therefore, the spacing of near and far gamma ray detector is determined. The research provides theoretical guidance for the tool parameter design and density determination of pulsed neutron gamma density logging technique. Copyright © 2017 Elsevier Ltd. All rights reserved.

Universal features of multiplicity distributions

International Nuclear Information System (INIS)

Balantekin, A.B.; Washington Univ., Seattle, WA

1994-01-01

Universal features of multiplicity distributions are studied and combinants, certain linear combinations of ratios of probabilities, are introduced. It is argued that they can be a useful tool in analyzing multiplicity distributions of hadrons emitted in high energy collisions and large scale structure of galaxy distributions

Multiple small-angle neutron scattering studies of anisotropic materials

CERN Document Server

Allen, A J; Long, G G; Ilavsky, J

2002-01-01

Building on previous work that considered spherical scatterers and randomly oriented spheroidal scatterers, we describe a multiple small-angle neutron scattering (MSANS) analysis for nonrandomly oriented spheroids. We illustrate this with studies of the multi-component void morphologies found in plasma-spray thermal barrier coatings. (orig.)

Charge collection and space charge distribution in neutron-irradiated epitaxial silicon detectors

Energy Technology Data Exchange (ETDEWEB)

Poehlsen, Thomas

2010-04-15

In this work epitaxial n-type silicon diodes with a thickness of 100 {mu}m and 150 {mu}m are investigated. After neutron irradiation with fluences between 10{sup 14} cm{sup -2} and 4 x 10{sup 15} cm{sup -2} annealing studies were performed. CV-IV curves were taken and the depletion voltage was determined for different annealing times. All investigated diodes with neutron fluences greater than 2 x 10{sup 14} cm{sup -2} showed type inversion due to irradiation. Measurements with the transient current technique (TCT) using a pulsed laser were performed to investigate charge collection effects for temperatures of -40 C, -10 C and 20 C. The charge correction method was used to determine the effective trapping time {tau}{sub eff}. Inconsistencies of the results could be explained by assuming field dependent trapping times. A simulation of charge collection could be used to determine the field dependent trapping time {tau}{sub eff}(E) and the space charge distribution in the detector bulk. Assuming a linear field dependence of the trapping times and a linear space charge distribution the data could be described. Indications of charge multiplication were seen in the irradiated 100 {mu}m thick diodes for all investigated fluences at voltages above 800 V. The space charge distribution extracted from TCT measurements was compared to the results of the CV measurements and showed good agreement. (orig.)

Investigation of dose distribution in mixed neutron-gamma field of boron neutron capture therapy using N isopropylacrylamide gel

Energy Technology Data Exchange (ETDEWEB)

Bavarmegin, Elham; Sadremomtaz, Alireza [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Khalafi, Hossein; Kasesaz, Yaser [Dept. of Physics, University of Guilan, Rasht (Iran, Islamic Republic of); Khajeali, Azim [Medical Education Research Center, Tabriz (Iran, Islamic Republic of)

2017-02-15

Gel dosimeters have unique advantages in comparison with other dosimeters. Until now, these gels have been used in different radiotherapy techniques as a reliable dosimetric tool. Because dose distribution measurement is an important factor for appropriate treatment planning in different radiotherapy techniques, in this study, we evaluated the ability of the N-isopropylacrylamide (NIPAM) polymer gel to record the dose distribution resulting from the mixed neutron-gamma field of boron neutron capture therapy (BNCT). In this regard, a head phantom containing NIPAM gel was irradiated using the Tehran Research Reactor BNCT beam line, and then by a magnetic resonance scanner. Eventually, the R2 maps were obtained in different slices of the phantom by analyzing T2-weighted images. The results show that NIPAM gel has a suitable potential for recording three-dimensional dose distribution in mixed neutron-gamma field dosimetry.

Time dependent worldwide distribution of atmospheric neutrons and of their products. I, II, III.

Science.gov (United States)

Merker, M.; Light, E. S.; Verschell, H. J.; Mendell, R. B.; Korff, S. A.

1973-01-01

Review of the experimental results obtained in a series of measurements of the fast neutron cosmic ray spectrum by means of high-altitude balloons and aircraft. These results serve as a basis for checking a Monte Carlo calculation of the entire neutron distribution and its products. A calculation of neutron production and transport in the earth's atmosphere is then discussed for the purpose of providing a detailed description of the morphology of secondary neutron components. Finally, an analysis of neutron observations during solar particle events is presented. The Monte Carlo output is used to estimate the contribution of flare particles to fluctuations in the steady state neutron distributions.

Comparison of MCNP6 and experimental results for neutron counts, Rossi-{alpha}, and Feynman-{alpha} distributions

Energy Technology Data Exchange (ETDEWEB)

Talamo, A.; Gohar, Y. [Argonne National Laboratory, 9700 S. Cass Ave., Lemont, IL 60439 (United States); Sadovich, S.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C. [Joint Institute for Power and Nuclear Research-Sosny, 99 Academician A.K. Krasin Str., Minsk 220109 (Belarus)

2013-07-01

MCNP6, the general-purpose Monte Carlo N-Particle code, has the capability to perform time-dependent calculations by tracking the time interval between successive events of the neutron random walk. In fixed-source calculations for a subcritical assembly, the zero time value is assigned at the moment the neutron is emitted by the external neutron source. The PTRAC and F8 cards of MCNP allow to tally the time when a neutron is captured by {sup 3}He(n, p) reactions in the neutron detector. From this information, it is possible to build three different time distributions: neutron counts, Rossi-{alpha}, and Feynman-{alpha}. The neutron counts time distribution represents the number of neutrons captured as a function of time. The Rossi-a distribution represents the number of neutron pairs captured as a function of the time interval between two capture events. The Feynman-a distribution represents the variance-to-mean ratio, minus one, of the neutron counts array as a function of a fixed time interval. The MCNP6 results for these three time distributions have been compared with the experimental data of the YALINA Thermal facility and have been found to be in quite good agreement. (authors)

The effect of an iron plug on the neutron flux distributions in water

International Nuclear Information System (INIS)

Lotfi, A.; Maayouf, R.M.A.; Megahid, R.

1978-01-01

This work is concerned with studying both fast and thermal neutron fluxes distribution in water and its perturbation due to the presence of a cylindrical iron plug. The measurements were carried out using a collimated neutron beam emitted from one of the horizontal channels of the ET-RR-1 reactor. The fast neutron fluxes were measured using phosphorus activation detectors, while the thermal neutron ones were measured using fission fragment track detectors from glass. The results show that the presence of an iron plug causes a remarkable change in the intensities of both the fast and thermal neutron fluxes distribution in the water medium surrounding the iron plug. The flux intensities at the peaks, formed beyond the iron plug in case of thermal neutrons, are also compared with values calculated using the available emperical formula

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

International Nuclear Information System (INIS)

Merchant, A.C.

1986-03-01

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

Experimental Determination of the Neutron Radiation-Dose Distribution in the Human Phantom

Energy Technology Data Exchange (ETDEWEB)

Stipcic, Neda [Institute Rudjer Bogkovic, Zagreb, Yugoslavia (Serbia)

1967-01-15

The quality of the radiation delivering the radiation dose to the human phantom is quite different from that of the incident neutron beam. This paper describes the experimental investigation of the variation of neutron dose related to the variation of neutron fluence with depth in the human phantom. The distribution of neutron radiation was determined in the human phantom - a cube of paraffin wax 25 cm x 25 cm x 50 cm with a density of 0.92 cm{sup -3}. Po-Be and Ra-Be point sources were used as neutron sources. Neutron fluences were measured using different types of detector: scintillation detector, BF{sub 3} counter, and nuclear-track emulsions. Since the fluence measurements with these three types of detectors were carried out under the same experimental conditions, it was possible to separate and analyse each part of the radiation dose in the paraffin. From the investigations, the distribution of the total radiation dose was obtained as a function of the paraffin depth. The maximum value of this dose distribution is constant with respect to the distance between the source and the paraffin phantom. From the results obtained, some conclusions may be drawn concerning the amount of absorbed radiation dose in the human phantom. (author)

Time Evolving Fission Chain Theory and Fast Neutron and Gamma-Ray Counting Distributions

International Nuclear Information System (INIS)

Kim, K. S.; Nakae, L. F.; Prasad, M. K.; Snyderman, N. J.; Verbeke, J. M.

2015-01-01

Here, we solve a simple theoretical model of time evolving fission chains due to Feynman that generalizes and asymptotically approaches the point model theory. The point model theory has been used to analyze thermal neutron counting data. This extension of the theory underlies fast counting data for both neutrons and gamma rays from metal systems. Fast neutron and gamma-ray counting is now possible using liquid scintillator arrays with nanosecond time resolution. For individual fission chains, the differential equations describing three correlated probability distributions are solved: the time-dependent internal neutron population, accumulation of fissions in time, and accumulation of leaked neutrons in time. Explicit analytic formulas are given for correlated moments of the time evolving chain populations. The equations for random time gate fast neutron and gamma-ray counting distributions, due to randomly initiated chains, are presented. Correlated moment equations are given for both random time gate and triggered time gate counting. There are explicit formulas for all correlated moments are given up to triple order, for all combinations of correlated fast neutrons and gamma rays. The nonlinear differential equations for probabilities for time dependent fission chain populations have a remarkably simple Monte Carlo realization. A Monte Carlo code was developed for this theory and is shown to statistically realize the solutions to the fission chain theory probability distributions. Combined with random initiation of chains and detection of external quanta, the Monte Carlo code generates time tagged data for neutron and gamma-ray counting and from these data the counting distributions.

Correction factor for the experimental prompt neutron decay constant

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Y.; Sadovich, S.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.

2013-01-01

Highlights: • Definition of a spatial correction factor for the experimental prompt neutron decay constant. • Introduction of a MCNP6 calculation methodology to simulate Rossi-alpha distribution for pulsed neutron sources. • Comparison of MCNP6 results with experimental data for count rate, Rossi-alpha, and Feynman-alpha distributions. • Improvement of the comparison between numerical and experimental results by taking into account the dead-time effect. - Abstract: This study introduces a new correction factor to obtain the experimental effective multiplication factor of subcritical assemblies by the point kinetics formulation. The correction factor is defined as the ratio between the MCNP6 prompt neutron decay constant obtained in criticality mode and the one obtained in source mode. The correction factor mainly takes into account the longer neutron lifetime in the reflector region and the effects of the external neutron source. For the YALINA Thermal facility, the comparison between the experimental and computational effective multiplication factors noticeably improves after the application of the correction factor. The accuracy of the MCNP6 computational model of the YALINA Thermal subcritical assembly has been verified by reproducing the neutron count rate, Rossi-α, and Feynman-α distributions obtained from the experimental data

Observation of Neutron Skyshine from an Accelerator Based Neutron Source

Energy Technology Data Exchange (ETDEWEB)

Franklyn, C. B. [Radiation Science Department, Necsa, PO Box 582, Pretoria 0001 (South Africa)

2011-12-13

A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >10{sup 11} n{center_dot}s{sup -1}. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

2001-03-01

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

Leading neutron energy and p{sub T} distributions in deep inelastic scattering and photoproduction at HERA

Energy Technology Data Exchange (ETDEWEB)

Chekanov, S.; Derrick, M.; Magill, S. [Argonne National Laboratory, Argonne, IL (US)] (and others)

2007-02-15

The production of energetic neutrons in ep collisions has been studied with the ZEUS detector at HERA. The neutron energy and p{sub T}{sup 2} distributions were measured with a forward neutron calorimeter and tracker in a 40 pb{sup -1} sample of inclusive deep inelastic scattering (DIS) data and a 6 pb{sup -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{sub T}{sup 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.)

Neutronics analysis of Dalat Research Reactor

International Nuclear Information System (INIS)

Pham Van Lam; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Nguyen Manh Hung; Pham Hong Son; Tran Quoc Duong

2006-01-01

Many neutronics codes have been used to calculate for Dalat Research Reactor (DRR) from 1983 (the first critical of DRR in December, 1983). The purposes of all calculations are to know exactly many important parameters related to Reactor Physics and Neutron Physics in reactor core. The results from calculation play important role in core and fuel management for DRR. Especially basing on the results we can predict about fuel cycle, fuel burn up distribution and plan for using optimize remain fresh fuel assemblies of DRR. By using system neutronics code including transport codes, diffusion codes and Mote Carlo code, many characteristics of fuel assemblies and other parameters of whole core were received such as main features of VVR-M2 fuel assembly type, multiplication factor, neutron flux distribution, power distribution, burn up distribution, excess reactivity, control rods worth, neutron spectrum, temperature reactivity coefficient ect. In the paper, brief description all computer codes to being used in DRR and the calculation results from the codes above are presented. (author)

The measurements of thermal neutron flux distribution in a paraffin

Indian Academy of Sciences (India)

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

Comparison of experimental and calculated neutron emission spectra and angular distributions

International Nuclear Information System (INIS)

Gruppelaar, H.; Akkermans, J.M.

1980-06-01

Experimental and calculated neutron emission spectra and angular distributions have been intercompared for 14.6-MeV neutron-induced reactions. The experimental data, measured by Hermsdorf et al., cover 34 elements in a large mass range. To calculate the differential neutron scattering cross sections a unified model of preequilibrium neutron emission was used, in which the generalized master equation of Mantzouranis et al. was solved with a fast exact matrix method, recently introduced by Akkermans. For the scattering kernel a three-term Legendre polynomial representation was adopted, which was either derived from the differential free nucleon-nucleon scattering cross section or fitted to obtain optimal agreement with the set of experimental data of Hermsdorf et al. The results of the last-mentioned calculation are quite acceptable in view of the fact that only two global parameters have been to describe the angular distributions of all experimental data. The report contains tables and graphs of the calculated Legendre coefficients and graphs of energy-averaged angular distributions for all 34 elements. It is further shown that improvements in the energy and angular distributions could be obtained by means of adjustment of the level-density parameters of the individual residual nuclei. Finally a short discussion is devoted to the problems of fitting angular distributions at backward angles by varying the model parameters or the specification of the initial condition. It is indicated that the so-called preequilibrium phase of the nuclear reaction actually consists of two different stages, the first one generating the forward-peaked angular distributions and the second one showing angular distributions symmetric about 90 0

Calculation of neutron flux distribution of thermal neutrons from microtron converter in a graphite moderator with water reflector

International Nuclear Information System (INIS)

Andrejsek, K.

1977-01-01

The calculation is made of the thermal neutron flux in the moderator and reflector by solving the neutron diffusion equation using the four-group theory. The correction for neutron absorption in the moderator was carried out using the perturbation theory. The calculation was carried out for four groups with the following energy ranges: the first group 2 MeV to 3 keV, the second group 3 keV to 5 eV, the third group 5 eV to 0.025 eV and the fourth group 0.025 eV. The values of the macroscopic cross section of capture and scattering, of the diffusion coefficient, the macroscopic cross section of the moderator, of the neutron age and the extrapolation length for the water-graphite moderator used in the calculations are given. The spatial distribution of the thermal neutron flux is graphically represented for graphite of a 30, 40, and 50 cm radius and for graphite of a 30 and 40 cm radius with a 10 cm water reflector; a graphic comparison is made of the distribution of the thermal neutron flux in water and in graphite, both 40 cm in radius. The system of graphite with reflector proved to be the best and most efficient system for raising the flux density of thermal neutrons. (J.P.)

Neutron multiplicities as a measure for scission time scales and reaction violences

International Nuclear Information System (INIS)

Knoche, K.; Scobel, W.; Sprute, L.

1991-01-01

We discuss the temporal evolution of the fusion-fission reactions 32 S + 197 Au, 232 Th measured for 838 MeV projectiles by means of the neutron clock method. The results confirm existent precision lifetime versus fissility data. The total neutron multiplicity as a measure of the initial excitation energy E * is compared with the folding angle method. (author). 13 refs, 8 figs

Neutron distribution modeling based on integro-probabilistic approach of discrete ordinates method

International Nuclear Information System (INIS)

Khromov, V.V.; Kryuchkov, E.F.; Tikhomirov, G.V.

1992-01-01

In this paper is described the universal nodal method for the neutron distribution calculation in reactor and shielding problems, based on using of influence functions and factors of local-integrated volume and surface neutron sources in phase subregions. This method permits to avoid the limited capabilities of collision-probability method concerning with the detailed calculation of angular neutron flux dependence, scattering anisotropy and empty channels. The proposed method may be considered as modification of S n - method with advantage of ray-effects elimination. There are presented the description of method theory and algorithm following by the examples of method applications for calculation of neutron distribution in three-dimensional model of fusion reactor blanket and in highly heterogeneous reactor with empty channel

The γ-ray angular distribution in fast neutron inelastic scattering from iron

Science.gov (United States)

Beyer, Roland; Dietz, Mirco; Bemmerer, Daniel; Junghans, Arnd R.; Kögler, Toni; Massarczyk, Ralph; Müller, Stefan; Schmidt, Konrad; Schwengner, Ronald; Szücs, Tamás; Takács, Marcell P.; Wagner, Andreas

2018-04-01

The angular distribution of γ-rays emitted after inelastic scattering of fast neutrons from iron was determined at the n ELBE neutron time-of-flight facility. An iron sample of natural isotopic composition was irradiated by a continuous photo-neutron spectrum in the energy range from about 0.1 up to 10 MeV. The de-excitation γ-rays of the four lowest excited states of 56Fe and the first excited state of 54Fe were detected using a setup of five high-purity germanium (HPGe) detectors and five LaBr3 scintillation detectors positioned around the sample at 30°, 55°, 90°, 125° and 150° with respect to the incoming neutron beam. The resulting angular distributions were fitted by Legendre polynomials up to 4th order and the angular distribution coefficients a2 and a4 were extracted. The angular distribution coefficients of three transitions in 56Fe are reported here for the first time. The results are applied to a previous measurement of the inelastic scattering cross section determined using a single HPGe detector positioned at 125°. Using the updated γ-ray angular distribution, the previous cross section results are in good agreement with reference data.

Secondary neutron production from thick Pb target by light particle irradiation

CERN Document Server

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

1999-01-01

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

The multiple disk chopper neutron time-of-flight spectrometer at NIST

International Nuclear Information System (INIS)

Altorfer, F.B.; Cook, J.C.; Copley, J.R.D.

1995-01-01

A highly versatile multiple disk chopper neutron time-of-flight spectrometer is being installed at the Cold Neutron Research Facility of the National institute of Standards and Technology. This new instrument will fill an important gap in the portfolio of neutron inelastic scattering spectrometers in North America. It will be used for a wide variety of experiments such as studies of magnetic and vibrational excitations, tunneling spectroscopy, and quasielastic neutron scattering investigations of local and translational diffusion. The instrument uses disk choppers to monochromate and pulse the incident beam, and the energy changes of scattered neutrons are determined from their times-of-flight to a large array of detectors. The disks and the guide have been designed to make the instrument readily adaptable to the specific performance requirements of experimenters. The authors present important aspects of the design, as well as estimated values of the flux at the sample and the energy resolution for elastic scattering. The instrument should be operational in 1996

Sensitivity Analysis of Cf-252 (sf) Neutron and Gamma Observables in CGMF

Energy Technology Data Exchange (ETDEWEB)

Carter, Austin Lewis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Talou, Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stetcu, Ionel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kiedrowski, Brian Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jaffke, Patrick John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-12-06

CGMF is a Monte Carlo code that simulates the decay of primary fission fragments by emission of neutrons and gamma rays, according to the Hauser-Feshbach equations. As the CGMF code was recently integrated into the MCNP6.2 transport code, great emphasis has been placed on providing optimal parameters to CGMF such that many different observables are accurately represented. Of these observables, the prompt neutron spectrum, prompt neutron multiplicity, prompt gamma spectrum, and prompt gamma multiplicity are crucial for accurate transport simulations of criticality and nonproliferation applications. This contribution to the ongoing efforts to improve CGMF presents a study of the sensitivity of various neutron and gamma observables to several input parameters for Californium-252 spontaneous fission. Among the most influential parameters are those that affect the input yield distributions in fragment mass and total kinetic energy (TKE). A new scheme for representing Y(A,TKE) was implemented in CGMF using three fission modes, S1, S2 and SL. The sensitivity profiles were calculated for 17 total parameters, which show that the neutron multiplicity distribution is strongly affected by the TKE distribution of the fragments. The total excitation energy (TXE) of the fragments is shared according to a parameter RT, which is defined as the ratio of the light to heavy initial temperatures. The sensitivity profile of the neutron multiplicity shows a second order effect of RT on the mean neutron multiplicity. A final sensitivity profile was produced for the parameter alpha, which affects the spin of the fragments. Higher values of alpha lead to higher fragment spins, which inhibit the emission of neutrons. Understanding the sensitivity of the prompt neutron and gamma observables to the many CGMF input parameters provides a platform for the optimization of these parameters.

Approximate solution for the reactor neutron probability distribution

International Nuclear Information System (INIS)

Ruby, L.; McSwine, T.L.

1985-01-01

Several authors have studied the Kolmogorov equation for a fission-driven chain-reacting system, written in terms of the generating function G(x,y,z,t) where x, y, and z are dummy variables referring to the neutron, delayed neutron precursor, and detector-count populations, n, m, and c, respectively. Pal and Zolotukhin and Mogil'ner have shown that if delayed neutrons are neglected, the solution is approximately negative binomial for the neutron population. Wang and Ruby have shown that if the detector effect is neglected, the solution, including the effect of delayed neutrons, is approximately negative binomial. All of the authors assumed prompt-neutron emission not exceeding two neutrons per fission. An approximate method of separating the detector effect from the statistics of the neutron and precursor populations has been proposed by Ruby. In this weak-coupling limit, it is assumed that G(x,y,z,t) = H(x,y)I(z,t). Substitution of this assumption into the Kolmogorov equation separates the latter into two equations, one for H(x,y) and the other for I(z,t). Solution of the latter then gives a generating function, which indicates that in the weak-coupling limit, the detector counts are Poisson distributed. Ruby also showed that if the detector effect is neglected in the equation for H(x,y), i.e., the detector efficiency is set to zero, then the resulting equation is identical with that considered by Wang and Ruby. The authors present here an approximate solution for H(x,y) that does not set the detector efficiency to zero

Measurement of the angular distribution of neutron-proton scattering at 10 MeV

International Nuclear Information System (INIS)

Haight, R.C.; Bateman, F.B.; Grimes, S.M.; Brient, C.E.; Massey, T.N.; Wasson, O.A.; Carlson, A.D.; Zhou, H.

1995-01-01

The relative angular distribution of neutrons scattered from protons was measured at an incident neutron energy of 10 MeV at the Ohio University Accelerator Laboratory. An array of 11 detector telescopes at laboratory angles of 0 to 60 degrees was used to detect recoil protons from neutron interactions with a CH 2 (polypropylene) target. Data for 7 of these telescopes were obtained with one set of electronics and are presented here. These data, from 108 to 180 degrees for the center-of-mass scattering angles, have a small slope which agrees better with angular distributions predicted by the Arndt phase shifts than with the ENDF/B-VI angular distribution

A Monte Carlo evaluation of analytical multiple scattering corrections for unpolarised neutron scattering and polarisation analysis data

International Nuclear Information System (INIS)

Mayers, J.; Cywinski, R.

1985-03-01

Some of the approximations commonly used for the analytical estimation of multiple scattering corrections to thermal neutron elastic scattering data from cylindrical and plane slab samples have been tested using a Monte Carlo program. It is shown that the approximations are accurate for a wide range of sample geometries and scattering cross-sections. Neutron polarisation analysis provides the most stringent test of multiple scattering calculations as multiply scattered neutrons may be redistributed not only geometrically but also between the spin flip and non spin flip scattering channels. A very simple analytical technique for correcting for multiple scattering in neutron polarisation analysis has been tested using the Monte Carlo program and has been shown to work remarkably well in most circumstances. (author)

Effects of Neutron Emission on Fragment Mass and Kinetic Energy Distribution from Thermal Neutron-Induced Fission of 235U

International Nuclear Information System (INIS)

Montoya, M.; Rojas, J.; Saetone, E.

2007-01-01

The mass and kinetic energy distribution of nuclear fragments from thermal neutron-induced fission of 235 U(n th ,f) have been studied using a Monte-Carlo simulation. Besides reproducing the pronounced broadening in the standard deviation of the kinetic energy at the final fragment mass number around m = 109, our simulation also produces a second broadening around m = 125. These results are in good agreement with the experimental data obtained by Belhafaf et al. and other results on yield of mass. We conclude that the obtained results are a consequence of the characteristics of the neutron emission, the sharp variation in the primary fragment kinetic energy and mass yield curves. We show that because neutron emission is hazardous to make any conclusion on primary quantities distribution of fragments from experimental results on final quantities distributions

Time-correlated neutron analysis of a multiplying HEU source

Energy Technology Data Exchange (ETDEWEB)

Miller, E.C., E-mail: Eric.Miller@jhuapl.edu [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States); Kalter, J.M.; Lavelle, C.M. [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States); Watson, S.M.; Kinlaw, M.T.; Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID (United States); Noonan, W.A. [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States)

2015-06-01

The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated {sup 3}He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations.

Time-correlated neutron analysis of a multiplying HEU source

Science.gov (United States)

Miller, E. C.; Kalter, J. M.; Lavelle, C. M.; Watson, S. M.; Kinlaw, M. T.; Chichester, D. L.; Noonan, W. A.

2015-06-01

The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated 3He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations.

Time-correlated neutron analysis of a multiplying HEU source

International Nuclear Information System (INIS)

Miller, E.C.; Kalter, J.M.; Lavelle, C.M.; Watson, S.M.; Kinlaw, M.T.; Chichester, D.L.; Noonan, W.A.

2015-01-01

The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated 3 He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations

Energy distribution of neutrons from the (n,2n) reaction in 238U

International Nuclear Information System (INIS)

Misulovin, A.

1978-12-01

Energy distribution of the first and second neutrons from (n,2n) scattering event in 238 U was evaluated according to the consistent compound nucleus model recently proposed by Segev. The law for deriving the energy distribution of secondary neutrons from a (n,2n) scattering event, depends on whether the reaction is considered as a simultaneous emission of two neutrons from one compound nucleus, or a successive emission of neutrons from different compound nuclei. Segev has presented a means of calculating the energy distribution assuming the latter model. The laws presented in the ENDF/B data files suggest the former model. The evaluation was based on inelastic level excitation and evaporation data for 238 U and 237 U. Data was retrieved from ENDF/B files. The evaporation data for 237 U was based on (γ,n) reaction 238 U. The inelastic level excitation data for 237 U was evaluated at the Soreq Nuclear Research Centre. It is concluded from the application of Segev's model to 238 U, that the energetic spectrum of secondary neutrons, is harder in the high range of energy than the one predicted by the use of the distribution law presented in ENDF/B data files. The spectrum of secondary (n,2n) neutrons, resulting from the interaction of 14 MeV neutrons in 238 U calculated with Segev's model, is compared with the corresponding spectrum of the LLL library, ENDF/B library and the recent evaluation of BNWL. It is found that the spectrum evaluated by LLL and BNWL is harder than that evaluated with Segev's model

Modified model of neutron resonances widths distributions. Results of reduced neutron widths approximation for mass region 35 ≤ A ≤ 249

International Nuclear Information System (INIS)

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

2011-01-01

The distributions of the reduced neutron widths of s-, p- and d-resonances of nuclei of any type from nuclear mass region 35 ≤ A ≤ 249 were approximated with maximal precision by the model which presents experimental data set as a superposition of a maximum of four independent neutron amplitudes. Under the assumption that each of these amplitudes has the Gauss distribution with the unique maximum there were determined the most probable values of contribution of each amplitude in summary width distribution, their most probable mean values and dispersions. Comparison of the obtained χ 2 values with value χ 2 at description of the experimental data by one distribution of neutron amplitudes with best fitted parameters shows that all widths from more than 157 analyzed data sets can have different types of wave functions

Direct integration multiple collision integral transport analysis method for high energy fusion neutronics

International Nuclear Information System (INIS)

Koch, K.R.

1985-01-01

A new analysis method specially suited for the inherent difficulties of fusion neutronics was developed to provide detailed studies of the fusion neutron transport physics. These studies should provide a better understanding of the limitations and accuracies of typical fusion neutronics calculations. The new analysis method is based on the direct integration of the integral form of the neutron transport equation and employs a continuous energy formulation with the exact treatment of the energy angle kinematics of the scattering process. In addition, the overall solution is analyzed in terms of uncollided, once-collided, and multi-collided solution components based on a multiple collision treatment. Furthermore, the numerical evaluations of integrals use quadrature schemes that are based on the actual dependencies exhibited in the integrands. The new DITRAN computer code was developed on the Cyber 205 vector supercomputer to implement this direct integration multiple-collision fusion neutronics analysis. Three representative fusion reactor models were devised and the solutions to these problems were studied to provide suitable choices for the numerical quadrature orders as well as the discretized solution grid and to understand the limitations of the new analysis method. As further verification and as a first step in assessing the accuracy of existing fusion-neutronics calculations, solutions obtained using the new analysis method were compared to typical multigroup discrete ordinates calculations

Projectile fragmentation of neutron-rich nuclei on light target (momentum distribution and nucleon-removal cross section)

International Nuclear Information System (INIS)

Kobayashi, T.; Tanihata, I.; Suzuki, T.

1992-01-01

Transverse momentum distributions of the projectile fragments from β-unstable nuclei have been measured with various projectile and target combinations. The momentum correlation of two neutrons in the neutron halo is extracted from the P c t distribution of 9 Li and hat of the neutrons. It is found that the two neutrons are moving in the same direction on average and thus strongly suggests the formation of a di-neutron in 11 Li. (Author)

Directional Stand-off Detection of Fast Neutrons and Gammas Using Angular Scattering Distributions

Energy Technology Data Exchange (ETDEWEB)

Vanier P. e.; Dioszegi, I.; Salwen, C.; Forman, L.

2009-10-25

We have investigated the response of a DoubleScatter Neutron Spectrometer (DSNS) for sources at long distances (gr than 200 meters). We find that an alternative method for analyzing double scatter data avoids some uncertainties introduced by amplitude measurements in plastic scintillators.Time of flight is used to discriminate between gamma and neutron events, and the kinematic distributions of scattering angles are assumed to apply. Non-relativistic neutrons are most likely to scatter at 45°, while gammas with energies greater than 2 MeV are most likely to be forward scattered. The distribution of scattering angles of fission neutrons arriving from a distant point source generates a 45° cone, which can be back-projected to give the source direction. At the same time, the distribution of Compton-scattered gammas has a maximum in the forward direction, and can be made narrower by selecting events that deposit minimal energy in the first scattering event. We have further determined that the shape of spontaneous fission neutron spectra at ranges gr than 110 m is still significantly different from thecosmic ray background.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

Axial distribution of absorbed doses in fast neutron field at the RB reactor

International Nuclear Information System (INIS)

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

1988-11-01

The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

Measurement of angular distribution of neutron flux for the 6 MeV race-track microtron based pulsed neutron source

Energy Technology Data Exchange (ETDEWEB)

Patil, B.J., E-mail: bjp@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India)

2010-09-15

The 6 MeV race track microtron based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products, where the low neutron flux requirement is desirable. Electrons impinges on a e-{gamma} target to generate bremsstrahlung radiations, which further produces neutrons by photonuclear reaction in {gamma}-n target. The optimisation of these targets along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium at different scattering angles. Angular distribution of neutron flux indicates that the flux decreases with increase in the angle and are in good agreement with the FLUKA simulation.

Neutron multiplication and shielding problems in PWR spent-fuel shipping casks

International Nuclear Information System (INIS)

Devillers, C.

1976-01-01

In order to evaluate the degree of accuracy of computational methods used for the shield design of spent-fuel shipping casks, comparisons were made between biological dose rate calculations and measurements at the surface of a cask carrying three PWR fuel assemblies (the fuel being successively wet and dry). The experimental methods used provide ksub(eff) with an accuracy of 0.024. Neutron multiplication coefficients provided by the APOLLO and DOT-3 codes are located within the uncertainty range of the experimentally derived values. The APOLLO plus DOT codes for neutron source calculations and ANISN plus DOT codes for neutron transmission calculations provide neutron dose rate predictions in agreement with measurements to within 10%. The PEPIN 76 code used for deriving fission product γ-rays and the point kernel code MERCURE 4 treating the γ-ray transmission give γ dose rate predictions that generally differ from measurements by less than 25%

Neutron and PIMC determination of the longitudinal momentum distribution of HCP, BCC and normal liquid 4He

International Nuclear Information System (INIS)

Blasdell, R.C.; Ceperley, D.M.; Simmons, R.O.

1993-07-01

Deep inelastic neutron scattering has been used to measure the neutron Compton profile (NCP) of a series of condensed 4 He samples at densities from 28.8 atoms/nm 3 (essentially the minimum possible density in the solid phase) up to 39.8 atoms/nm 3 using a chopper spectrometer at the Argonne National Laboratory Intense Pulsed Neutron Source. At the lowest density, the NCP was measured along an isochore through the hcp, bcc, and normal liquid phases. Average atomic kinetic energies are extracted from each of the data sets and are compared to both published and new path integral Monte-Carlo (PIMC) calculations as well as other theoretical predictions. In this preliminary analysis of the data, account is taken of the effects of instrumental resolution, multiple scattering, and final-state interactions. Both our measurements and the PIMC theory show that there are only small differences in the kinetic energy and longitudinal momentum distribution of isochoric helium samples, regardless of their phase or crystal structure

Visualization of neutron flux and power distributions in TRIGA Mark II reactor as an educational tool

International Nuclear Information System (INIS)

Snoj, Luka; Ravnik, Matjaz; Lengar, Igor

2008-01-01

Modern Monte Carlo computer codes (e.g. MCNP) for neutron transport allow calculation of detailed neutron flux and power distribution in complex geometries with resolution of ∼1 mm. Moreover they enable the calculation of individual particle tracks, scattering and absorption events. With the use of advanced software for 3D visualization (e.g. Amira, Voxler, etc.) one can create and present neutron flux and power distribution in a 'user friendly' way convenient for educational purposes. One can view axial, radial or any other spatial distribution of the neutron flux and power distribution in a nuclear reactor from various perspectives and in various modalities of presentation. By visualizing the distribution of scattering and absorption events and individual particle tracks one can visualize neutron transport parameters (mean free path, diffusion length, macroscopic cross section, up-scattering, thermalization, etc.) from elementary point of view. Most of the people remember better, if they visualize the processes. Therefore the representation of the reactor and neutron transport parameters is a convenient modern educational tool for the (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. The visualization of neutron flux and power distributions in Jozef Stefan Institute TRIGA Mark II research reactor is treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. (authors)

Evaporation residue cross sections and average neutron multiplicities in the /sup 64/Ni+/sup 92/Zr and /sup 12/C+/sup 144/Sm reactions leading to /sup 156/Er

Energy Technology Data Exchange (ETDEWEB)

Janssens, R.V.F.; Holzmann, R.; Henning, W.; Khoo, T.L.; Lesko, K.T.; Stephans, G.S.F.; Radford, D.C.; Van den Berg, A.M.; Kuehn, W.; Ronningen, R.M.

1986-11-27

Evaporation residue cross sections and neutron multiplicity distributions have been measured for the /sup 12/C+/sup 144/Sm and /sup 64/Ni+/sup 92/Zr reactions leading to the same compound nucleous /sup 156/Er. Statistical model calculations can account for the data in the /sup 12/C-induced reaction. In contrast, the inhibition of neutron emission with respect to statistical model predictions seen in /sup 64/Ni+/sup 92/Zr cannot be explained even with the inclusion of the broad angular momentum distributions required to describe the fusion cross section data.

Correlation studies of neutron multiplicities in the 252Cf spontaneous fission

International Nuclear Information System (INIS)

Alkhazov, I.D.; Dmitriev, V.D.; Kovalenko, S.S.; Kuznetsov, A.V.; Malkin, L.Z.; Petrzhak, K.A.; Petrov, B.F.; Shpakov, V.I.

1988-01-01

Correlations between the numbers of neutrons emitted by the 252 Cf spontaneous fission fragments have been studied as a function of the fragment mass and total kinetic energy. Behaviour of the neutron number dispersions and covariances was studied for the region of symmetric fission. Parameters of the complementary fragment excitation energy distribution (mean values, dispersions, covariances) were determined. Various factors describing correlations between the complementary fragment excitation energies are considered

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-26

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

Measuring method for effective neutron multiplication factor upon containing irradiated fuel assembly

International Nuclear Information System (INIS)

Ueda, Makoto; Mitsuhashi, Ishi; Sasaki, Tomoharu.

1993-01-01

A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

Residual stress distribution in carbon steel pipe welded joint measured by neutron diffraction

International Nuclear Information System (INIS)

Hayashi, Makoto; Ishiwata, Masayuki; Morii, Yukio; Minakawa, Nobuaki

2000-01-01

In order to estimate crack growth behavior of fatigue and stress corrosion cracking in pipes, the residual stress distribution near the pipe weld region has to be measured through the wall thickness. Since the penetration depth of neutron is deep enough to pass through the thick pipe wall, the neutron diffraction technique for the residual stress measurement is effective for this purpose. At the first step the residual stress distribution near the weld region in a butt-welded carbon steel pipe was measured by the neutron diffraction. Significant stresses extended only to a distance of 30 mm from the center of the weld. The major tensile stresses occurred in the hoop direction in the fusion and heat affected zones of the weldment, and they attained a level greater than 200 MPa through the thickness. While the axial residual stress at the inside surface was 50 MPa, the stress at the outside surface was -100 MPa. The comparison of residual stress distributions measured by the neutron diffraction, the X-ray diffraction and the strain gauge method reveals that the neutron diffraction is the most effective for measuring the residual stress inside the structural components. (author)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Physics of neutron emission in fission

International Nuclear Information System (INIS)

Lemmel, H.D.

1989-06-01

The document contains the proceedings of the IAEA Consultants' Meeting on the Physics of Neutron Emission in Fission, Mito City (Japan), 24-27 May 1988. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers cover the following topics: Energy dependence of the number of fission neutrons ν-bar (3 papers), multiplicity distribution of fission neutrons (3 papers), competition between neutron and γ-ray emission (4 papers), the fission neutron yield in resonances (2 papers) and the energy spectrum of fission neutrons in experiment (9 papers), theory (4 papers) and evaluation (1 paper). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Solid-State Neutron Multiplicity Counting System Using Commercial Off-the-Shelf Semiconductor Detectors

Energy Technology Data Exchange (ETDEWEB)

Rozhdestvenskyy, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-08-09

This work iterates on the first demonstration of a solid-state neutron multiplicity counting system developed at Lawrence Livermore National Laboratory by using commercial off-the-shelf detectors. The system was demonstrated to determine the mass of a californium-252 neutron source within 20% error requiring only one-hour measurement time with 20 cm² of active detector area.

MCNP and visualization of neutron flux and power distributions

International Nuclear Information System (INIS)

Snoj, L.; Lengar, I.; Zerovnik, G.; Ravnik, M.

2009-01-01

The visualization of neutron flux and power distributions in two nuclear reactors (TRIG A type research reactor and typical PWR) and one thermonuclear reactor (tokamak type) are treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. The remembrance of most of the people is better, if they visualize a process. Therefore a representation of the reactor and neutron transport parameters is a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for core and irradiation planning. (authors)

Measurement of mass distribution of U-235 fission products in the intermediate neutron region

International Nuclear Information System (INIS)

Nakagomi, Yoshihiro; Kobayashi, Shohei; Yamamoto, Shuji; Kanno, Ikuo; Wakabayashi, Hiroaki.

1982-01-01

The mass distribution and the momentum distribution of U-235 fission products in the intermediate neutron region were measured by using a combination system of the Yayoi intermediate neutron column and an electron linear accelerator. The double energy measurement method was applied. A fission chamber, which consists of an enriched uranium target and two Si surface barrier detectors, was used for the measurement of the neutrons with energy above 1.3 eV. The linear accelerator was operated at the repetition rate of 100 Hz and the pulse width of 10 ns. The data obtained by the two-dimensional pulse height analysis were analyzed by the Schmitt's method. The preliminary results of the mass distribution and the momentum distribution of fission fragments were obtained. (Kato, T.)

Neutron angular distribution in a plasma focus obtained using nuclear track detectors.

Science.gov (United States)

Castillo-Mejía, F; Herrera, J J E; Rangel, J; Golzarri, J I; Espinosa, G

2002-01-01

The dense plasma focus (DPF) is a coaxial plasma gun in which a high-density, high-temperature plasma is obtained in a focused column for a few nanoseconds. When the filling gas is deuterium, neutrons can be obtained from fusion reactions. These are partially due to a beam of deuterons which are accelerated against the background hot plasma by large electric fields originating from plasma instabilities. Due to a beam-target effect, the angular distribution of the neutron emission is anisotropic, peaked in the forward direction along the axis of the gun. The purpose of this work is to illustrate the use of CR-39 nuclear track detectors as a diagnostic tool in the determination of the time-integrated neutron angular distribution. For the case studied in this work, neutron emission is found to have a 70% contribution from isotropic radiation and a 30% contribution from anisotropic radiation.

A one-dimensional, one-group absorption-production nodal method for neutron flux and power distributions calculations

International Nuclear Information System (INIS)

Ferreira, C.R.

1984-01-01

It is presented the absorption-production nodal method for steady and dynamical calculations in one-dimension and one group energy. It was elaborated the NOD1D computer code (in FORTRAN-IV language). Calculations of neutron flux and power distributions, burnup, effective multiplication factors and critical boron concentration were made with the NOD1D code and compared with results obtained through the CITATION code, which uses the finite difference method. The nuclear constants were produced by the LEOPARD code. (M.C.K.) [pt

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Multiplicity distributions for p-barp collisions

International Nuclear Information System (INIS)

Huang, D.W.; Yen, E.

1988-01-01

The multiplicity distribution for P-barP collisions at √s = 540 GeV is written as the sum of Poisson distributions at different impact parameter b. An energy independent relation between the variable z-bar and b is suggested. With this relation, the multiplicity distributions at √s = 200 and 900 GeV are described well. The distribution at √s = 1600 GeV is predicted

Fast Neutrons - LET Distributions and the Response of Mammalian Cells

Energy Technology Data Exchange (ETDEWEB)

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

1968-03-15

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

Development of a position-sensitive fission counter and measurement of neutron flux distributions

International Nuclear Information System (INIS)

Yamagishi, Hideshi; Soyama, Kazuhiko; Kakuta, Tsunemi

2001-08-01

A position-sensitive fission counter (PSFC) that operates in high neutron flux and high gamma-ray background such as at the side of a power reactor vessel has been developed. Neutron detection using the PSFC with a solenoid electrode is based on a delay-line method. The PSFC that has the outer diameter of 25 mm and the sensitive length of 1000 mm was manufactured for investigation of the performances. The PSFC provided output current pulses that were sufficiently higher than the alpha noise, though the PSFC has a solenoid electrode and large electrode-capacitance. The S/N ratio of PSFC outputs proved to be higher than that of ordinary fission counters with 200 mm sensitive length. A performance test to measure neutron flux distributions by a neutron measuring system with the PSFC was carried out by the side of a graphite pile, W2.4 x H1.4 x L1.2 m, with neutron sources, Am-Be 370 GBq x 2. It was confirmed that the neutron flux distribution was well measured with the system. (author)

Neutron production in interactions of relativistic protons and deuterons with lead targets

International Nuclear Information System (INIS)

Yurevich, V.I.; Amelin, N.S.; Yakovlev, R.M.; Nikolaev, V.A.; Lyapin, V.G.; Tsvetkov, I.O.

2005-01-01

Results on the neutron double-differential cross sections and yields obtained in the time-of-flight measurements with different lead targets and beams of protons and deuterons at an energy of about 2 GeV are discussed. The neutron spatial-energy distribution for an extended lead target was studied by the threshold detector method in the energy range of protons and deuterons 1-3.7 GeV. A dependence of the mean neutron multiplicity, energy of neutrons, and process of neutron multiplication in lead on the target dimension, and the type and energy of the beam particle is analyzed. (author)

Experimental Assessment of a New Passive Neutron Multiplication Counter for Partial Defect Verification of LWR Fuel Assemblies

International Nuclear Information System (INIS)

LaFleur, A.; Menlove, H.; Park, S.-H.; Lee, S. K.; Oh, J.-M.; Kim, H.-D.

2015-01-01

The development of non-destructive assay (NDA) capabilities to improve partial defect verification of spent fuel assemblies is needed to improve the timely detection of the diversion of significant quantities of fissile material. This NDA capability is important to the implementation of integrated safeguards for spent fuel verification by the International Atomic Energy Agency (IAEA) and would improve deterrence of possible diversions by increasing the risk of early detection. A new NDA technique called Passive Neutron Multiplication Counter (PNMC) is currently being developed at Los Alamos National Laboratory (LANL) to improve safeguards measurements of LightWater Reactor (LWR) fuel assemblies. The PNMC uses the ratio of the fast-neutron emission rate to the thermalneutron emission rate to quantify the neutron multiplication of the item. The fast neutrons versus thermal neutrons are measured using fission chambers (FC) that have differential shielding to isolate fast and thermal energies. The fast-neutron emission rate is directly proportional to the neutron multiplication in the spent fuel assembly; whereas, the thermalneutron leakage is suppressed by the fissile material absorption in the assembly. These FCs are already implemented in the basic Self-Interrogation Neutron Resonance Densitometry (SINRD) detector package. Experimental measurements of fresh and spent PWR fuel assemblies were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using a hybrid PNMC and SINRD detector. The results from these measurements provides valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. (author)

Apparatus and method for detecting contraband using fast neutron activation

International Nuclear Information System (INIS)

Gozani, T.; Sawa, Z.P.; Shea, P.M.

1992-01-01

This patent describes a method of detecting contraband within an object under investigation. It comprises: generating a beam of case neutrons; irradiating the object with the beam of fast neutrons, the fast neutrons interacting with atomic nuclei of the elements contained within the object to produce a gamma-ray spectrum having spectral lines characteristic of the elements contained within the object; measuring the spectral lines of the gamma-ray spectrum using a multiplicity of gamma-ray detectors judiciously positioned around the object; detecting the number of neutrons that pass through the object without interacting substantially with atomic nuclei within the object; determining the spatial and density distributions of the atomic nuclei of the elements contained within the object from the measured gamma-ray spectrum obtained from the multiplicity of gamma-ray detectors and the number of neutrons that pass through the object; comparing the measured spatial and density distributions of the atomic nuclei of the elements within the object with known spatial and density distributions of atomic nuclei for elements characteristic of contraband; and determining that contraband is present within the object when the comparison indicates a substantial match

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

International Nuclear Information System (INIS)

Duering, I.; Jahnke, U.

1993-01-01

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

Spallation neutron production and the current intra-nuclear cascade and transport codes

International Nuclear Information System (INIS)

Filges, D.; Goldenbaum, F.

2001-01-01

A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models. (orig.)

Spallation neutron production and the current intra-nuclear cascade and transport codes

Science.gov (United States)

Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.

A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.

Calculation of the distribution of the escaping from a fissionable sample neutrons number when introducing one fission neutron in that sample

International Nuclear Information System (INIS)

Dorlet, J.

1991-01-01

A describing algorithm furnishes the probabilities of having exactly N escaping neutrons in the descent of one fission neutron, using the punctual reactor model. Calculations can be performed even for N-values greater than 1000. Numerical results show that discrete neutrons counting is unfit to obtain the N mean value, even for very far subcritical devices. That mean value is still very used in the existing theoretical studies, because its obvious correlation with the device effective multiplication coefficient. For that reason modelling coincidence neutrons counting is not suitable using statistical moments approach, but only using probabilities it selves [fr

Production, Distribution, and Applications of Californium-252 Neutron Sources

International Nuclear Information System (INIS)

Balo, P.A.; Knauer, J.B.; Martin, R.C.

1999-01-01

The radioisotope 252 Cf is routinely encapsulated into compact, portable, intense neutron sources with a 2.6-year half-life. A source the size of a person's little finger can emit up to 10 11 neutrons/s. Californium-252 is used commercially as a reliable, cost-effective neutron source for prompt gamma neutron activation analysis (PGNAA) of coal, cement, and minerals, as well as for detection and identification of explosives, laud mines, and unexploded military ordnance. Other uses are neutron radiography, nuclear waste assays, reactor start-up sources, calibration standards, and cancer therapy. The inherent safety of source encapsulations is demonstrated by 30 years of experience and by U.S. Bureau of Mines tests of source survivability during explosions. The production and distribution center for the U. S Department of Energy (DOE) Californium Program is the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). DOE sells 252 Cf to commercial reencapsulators domestically and internationally. Sealed 252 Cf sources are also available for loan to agencies and subcontractors of the U.S. government and to universities for educational, research, and medical applications. The REDC has established the Californium User Facility (CUF) for Neutron Science to make its large inventory of 252 Cf sources available to researchers for irradiations inside uncontaminated hot cells. Experiments at the CUF include a land mine detection system, neutron damage testing of solid-state detectors, irradiation of human cancer cells for boron neutron capture therapy experiments, and irradiation of rice to induce genetic mutations

Measurements and applications of neutron multiple scattering in resonance region

International Nuclear Information System (INIS)

Ohkubo, Makio

1977-02-01

Capture yield of neutrons impinging on a thick material is complicated due to self-shielding and multiple scattering, especially in the resonance region. When the incident neutron energy is equal to a resonance energy of the material, capture probability of the neutron increases with sample thickness and reaches a saturation value P sub(CO). There is a simple relation between P sub(CO) and GAMMA sub(n)/GAMMA and the recoil energy by the Monte-Carlo calculation. To examine validity of the relation, P sub(CO) was measured for 19 resonances in 12 nuclides with thick samples, using a JAERI linac time-of-flight spectrometer with Moxon-Rae type gamma ray detector and transmission type neutron flux monitor. Results of the measurements confirmed the validity. With this relation, the GAMMA sub(n)/GAMMA or GAMMA sub(γ)/GAMMA value can be obtained from the measured P sub(CO), and also the level spins be determined by combining the transmission data. Because of the definition of P sub(CO), determination of the resonance parameters is not sensitive to the sample thickness as far as it is sufficiently thick. (auth.)

Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.

Science.gov (United States)

Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y

2003-10-01

A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

Incident energy and target dependence of interaction cross sections and density distribution of neutron drip-line nuclei

International Nuclear Information System (INIS)

Shimoura, S.

1992-01-01

The relation between nuclear density distribution and interaction cross section is discussed in terms of Glauber model. Based on the model, density distribution of neutron drip-line nucleus 11 Be and 11 Li is determined experimentally from incident energy dependence of interaction cross sections of 11 Be and 11 Li on light targets. The obtained distributions have long tails corresponding to neutron halos of loosely bound neutrons. (Author)

Evaluation of the neutron and gamma-ray production cross-sections for 55Mn

International Nuclear Information System (INIS)

Takahashi, H.

1974-11-01

The evaluation of neutron and gamma production cross sections for manganese-55 from 1.0 (10) -5 eV to 20.0 MeV for ENDF/ B-IV is summarized. Included are resonance parameters, neutron cross sections, angular and energy distribution of secondary neutrons, gamma multiplicities and transition probability array, gamma angular and energy distributions, nuclear model calculations, uncertainty estimates of cross sections, and evaluated cross sections. (U.S.)

The Chain-Length Distribution in Subcritical Systems

International Nuclear Information System (INIS)

Nolen, Steven Douglas

2000-01-01

The individual fission chains that appear in any neutron multiplying system provide a means, via neutron noise analysis, to unlock a wealth of information regarding the nature of the system. This work begins by determining the probability density distributions for fission chain lengths in zero-dimensional systems over a range of prompt neutron multiplication constant (K) values. This section is followed by showing how the integral representation of the chain-length distribution can be used to obtain an estimate of the system's subcritical prompt multiplication (MP). The lifetime of the chains is then used to provide a basis for determining whether a neutron noise analysis will be successful in assessing the neutron multiplication constant, k, of the system in the presence of a strong intrinsic source. A Monte Carlo transport code, MC++, is used to model the evolution of the individual fission chains and to determine how they are influenced by spatial effects. The dissertation concludes by demonstrating how experimental validation of certain global system parameters by neutron noise analysis may be precluded in situations in which the system K is relatively low and in which realistic detector efficiencies are simulated

The Chain-Length Distribution in Subcritical Systems

Energy Technology Data Exchange (ETDEWEB)

Nolen, Steven Douglas [Texas A & M Univ., College Station, TX (United States)

2000-06-01

The individual fission chains that appear in any neutron multiplying system provide a means, via neutron noise analysis, to unlock a wealth of information regarding the nature of the system. This work begins by determining the probability density distributions for fission chain lengths in zero-dimensional systems over a range of prompt neutron multiplication constant (K) values. This section is followed by showing how the integral representation of the chain-length distribution can be used to obtain an estimate of the system's subcritical prompt multiplication (MP). The lifetime of the chains is then used to provide a basis for determining whether a neutron noise analysis will be successful in assessing the neutron multiplication constant, k, of the system in the presence of a strong intrinsic source. A Monte Carlo transport code, MC++, is used to model the evolution of the individual fission chains and to determine how they are influenced by spatial effects. The dissertation concludes by demonstrating how experimental validation of certain global system parameters by neutron noise analysis may be precluded in situations in which the system K is relatively low and in which realistic detector efficiencies are simulated.

Radiation dose distribution monitoring at neutron radiography facility area, Nuclear Energy Unit, Malaysia

International Nuclear Information System (INIS)

Abdul Razak Daud

1995-01-01

One experiment was carried out to get the distribution of radiation doses at the neutron radiography facilities, Nuclear Energy Unit, Malaysia. The analysis was done to evaluate the safety level of the area. The analysis was used in neutron radiography work

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

International Nuclear Information System (INIS)

Jammes, Christian; Filliatre, Philippe; Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan

2015-01-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

Energy Technology Data Exchange (ETDEWEB)

Jammes, Christian; Filliatre, Philippe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan [Division of Applied Nuclear Physics, Uppsala University, SE-75120 Uppsala, (Sweden)

2015-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

Neutron generators with size scalability, ease of fabrication and multiple ion source functionalities

Science.gov (United States)

Elizondo-Decanini, Juan M

2014-11-18

A neutron generator is provided with a flat, rectilinear geometry and surface mounted metallizations. This construction provides scalability and ease of fabrication, and permits multiple ion source functionalities.

Sensitivity of the 252Cf(sf neutron observables to the FREYA input yield functions Y(A, Z, TKE

Directory of Open Access Journals (Sweden)

Randrup Jørgen

2017-01-01

Full Text Available Within the framework of the fission event generator FREYA, we are studying the sensitivity of various neutron observables to the yield distribution Y (A,Z,TKE used as input to the code. Concentrating on spontaneous fission of 252Cf, we have sampled a large number of different input yield functions based on χ2 fits to the experimental data on Y (A and Y (TKE|A. For each of these input yield distributions, we then use FREYA to generate a large sample of complete fission events from which we extract a variety of neutron observables, including the multiplicity distribution, the associated correlation coefficients, and its factorial moments, the dependence of the mean neutron multiplicity on the total fragment kinetic energy TKE and on the fragment mass number A, the neutron energy spectrum, and the two-neutron angular correlation function. In this way, we can determine the variation of these observables resulting from the uncertainties in the experimental mesurements. The imposition of a constraint on the resulting mean neutron multiplicity reduces the variation of the calculated neutron observables and provides a means for shrinking the uncertainties associated with the measured data.

Sensitivity of the 252Cf(sf) neutron observables to the FREYA input yield functions Y(A, Z, TKE)

Science.gov (United States)

Randrup, Jørgen; Talou, Patrick; Vogt, Ramona

2017-09-01

Within the framework of the fission event generator FREYA, we are studying the sensitivity of various neutron observables to the yield distribution Y (A,Z,TKE) used as input to the code. Concentrating on spontaneous fission of 252Cf, we have sampled a large number of different input yield functions based on χ2 fits to the experimental data on Y (A) and Y (TKE|A). For each of these input yield distributions, we then use FREYA to generate a large sample of complete fission events from which we extract a variety of neutron observables, including the multiplicity distribution, the associated correlation coefficients, and its factorial moments, the dependence of the mean neutron multiplicity on the total fragment kinetic energy TKE and on the fragment mass number A, the neutron energy spectrum, and the two-neutron angular correlation function. In this way, we can determine the variation of these observables resulting from the uncertainties in the experimental mesurements. The imposition of a constraint on the resulting mean neutron multiplicity reduces the variation of the calculated neutron observables and provides a means for shrinking the uncertainties associated with the measured data.

Scaling neutron absorbed dose distributions from one medium to another

International Nuclear Information System (INIS)

Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

1982-11-01

Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone TE-solutions, mineral oil and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. the OAR's measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. It is recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. A table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS

International Nuclear Information System (INIS)

Verbeke, Jerome M.

1999-01-01

A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only

Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS

Energy Technology Data Exchange (ETDEWEB)

Verbeke, Jerome M.

1999-12-14

A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only.

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

Spatial fluxes and energy distributions of reactor fast neutrons in two types of heat resistant concretes

International Nuclear Information System (INIS)

Akki, T.S.; Benayad, S.A.; Megahid, R.M.

1992-01-01

Measurements have been carried out to study the spatial fluxes and energy distributions of reactor fast neutrons transmitted through two types of heat resistant concretes, serpentine concrete and magnetic lemonite concrete. The physical, chemical and mechanical properties of these concretes were checked by well known techniques. In addition, the effect of heating at temperatures up to 500deg C on the crystaline water content was checked by the method of differential thermal analysis. Measurements were performed using a collimated beam of reactor neutrons emitted from a 10 MW research reactor. The neutron spectra transmitted through concrete barriers of different thickness were measured by a scintillation spectrometer with NE-213 liquid organic scintillator. Discrimination against undesired pulses due to gamma-rays was achieved by a method based on pulse shape discrimination technique. The operating principle of this technique is based on the comparison of two weighted time integrals of the detector signal. The measured pulse amplitude distribution was converted to neutron energy distribution by a computational code based on double differentiation technique. The spectrometer workability and the accuracy of the unfolding technique were checked by measuring the neutron spectra of neutrons from Pu-α-Be and 252 Cf neutron sources. The obtained neutron spectra for the two concretes were used to derive the total cross sections for neutrons of different energies. (orig.)

The study on neutron and photon distribution of AP1000 reactor by MCNP code

International Nuclear Information System (INIS)

Chen Defeng; Shen Mingqi

2014-01-01

The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)

Research on amplification multiple of source neutron number for ADS

International Nuclear Information System (INIS)

Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qingbiao; Ding Dazhao

1998-01-01

NJOY-91.91 and MILER code systems was applied to process and generate 44 group cross sections in AMPX master library format from CENDL-2 and ENDF/B-6. It is important an ADS (Accelerator-Driven System) assembly spectrum is used as the weighting spectrum for generating multi-group constants. Amplification multiples of source neutron number for several fast assemblies were calculated

A training and educational tool for neutron coincidence measurements

International Nuclear Information System (INIS)

Huszti, J.; Bagi, J.; Langner, D.

2009-01-01

Neutron coincidence counting techniques are widely used for nuclear safeguards inspection. They are based on the detection of time correlated neutrons created from spontaneous or induced fission of plutonium and some other actinides. IAEA inspectors are trained to know and to use this technique, but it is not easy to illustrate and explain the basics of the neutron coincidence counting. The traditional shift registers or multiplicity counters give only multiplicity distributions and the singles, doubles and triples count rates. Using the list mode method for the recording and evaluation of neutron coincidence data makes it easier to teach this technique. List mode acquisition is a relatively new way to collect data in neutron coincidence counting. It is based on the recording of the follow-up times of neutron pulses originating from a neutron detector into a file. The recorded pulse train can be evaluated with special software after the measurement. Hardware and software for list mode neutron coincidence acquisition have been developed in the Institute of Isotopes and is called a Pulse Train Reader. A system called Virtual Source for replaying pulse trains registered with the list mode device has also been developed. The list mode device and the pulse train 're-player' together build a good educational tool for teaching the basics of neutron coincidence counting. Some features of the follow-up time, multiplicity and Rossi-alpha distributions can be well demonstrated by replaying artificially generated or pre-recorded pulse trains. The choice of real sources is stored on DVD. There is no need to transport and maintain real sources for the training. Virtual sources also give the possibility of investigating rare sources that trainees would not have access to otherwise. (authors)

Energy and angular distributions of neutrons from 252Cf spontaneous fission

International Nuclear Information System (INIS)

Vasil'ev, Yu.A.; Sidorov, L.V.; Vasil'eva, N.K.

1982-01-01

Some results from a first series of measurements of energy and angular distributions of neutrons from 252 Cf spontaneous fission using a spectrometer with high neutron detection efficiency, i.e. a 4π neutron time-of-flight spectrometer, were already presented. Subsequently, a second series of measurements was performed using a more sophisticated technique. For this second series, we used a more intense 252 Cf layer (25,000 spontaneous fissions per second). The angular resolution was improved by a factor of 2-3 by combining the hexahedral counter modules, placed at the same angle with respect to the direction of motion of the fragments, in new panoramic counters. The neutron counters were calibrated against the average 252 Cf neutron spectrum at several positions of the axis of the fragment detector with respect to the neutron counters. In the spectrum measurements and calibration work, the scattered neutron background was not determined theoretically, as in the first series of measurements, but experimentally using four extra scintillation counters with scatter cones; the counters were set up at 60 deg., 80 deg., 100 deg., and 120 deg. to the direction of separation of the fragments

Study of the neutron flux distribution in acylindrical reactor

Directory of Open Access Journals (Sweden)

A. Vidal-Ferràndiz

2017-08-01

Full Text Available In the Energy Engineering Degree of the Universitat Politècnica de València, the students attend to the Nuclear Technology course, in which the basic knowledge of this technology is presented. A main objective of this technology is to obtain neutron population distribution inside a reactor core, in order to maintain the fission reaction chain. As this activity cannot be experimentally developed, mathematical modelling is of great importance to achieve such objective.  One of the computer laboratories proposed consists in the neutron flux determination analytically and numerically in a cylindrical geometry. The analytical solution makes use of the Bessel functions and is a good example of their applications. Alternatively, a numerical solution based on finite differences is used to obtain an approximate solution of the neutron flux. In this work, different discretizations of the cylindrical geometry are implemented and their results are compared.

Study of the ferrimagnetic and paramagnetic phases of magnetite measured by multiple neutron diffraction

International Nuclear Information System (INIS)

Mazzocchi, V.L.

1992-01-01

Structural parameters of the ferrimagnetic and paramagnetic phases of magnetite have been refined from neutron multiple diffraction data. Experimental multiple diffraction patterns used in the refinement, were obtained by measuring the 111 primary reflection of a natural single crystal of this compound, at room temperature for the ferrimagnetic phase and 703 0 C for the paramagnetic phase. Corresponding theoretical patterns for both phases have been calculated by the program MULTI which uses the iterative method for the intensity calculations in neutron multiple diffraction. In this method intensities are calculated as Taylor series expansions summed up to a order sufficient for a good approximation. A step by step process has been used in the refinements according to the parameter-shift method. Both isotropic and anisotropic thermal parameters were used in the calculation of the temperature factor. (author)

Distribution of the thermal neutron field around the graphite reflector of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Thermal neutron flux distributions around the graphite reflector of the Dalat Nuclear Research Reactor are determined by the method for neutron activating Cu foils. The major results are as follows: a/The axial distributions at the inner and outer margins of the graphite reflector have unsymmetrical shapes, similar to axial distributions in the core. There is a dissimilarity between the distribution curves at the inner margin and those at the outer margin of the reflector. b/ The radial distribution on the upper surface of the graphite reflector is measured and is described by the two-group neutron diffusion theory. The maximal value of the curve lies at the position of R{sub m}ax = 22.5 cm. c/ The distribution in the twenty water irradiation holes around the rotary specimen rack is obtained. (author). 3 refs., 5 figs., 1 tab.

Superclusters and hadronic multiplicity distributions

International Nuclear Information System (INIS)

Shih, C.C.; Carruthers, P.

1986-01-01

The multiplicity distribution is expressed in terms of supercluster production in hadronic processes at high energy. This process creates unstable clusters at intermediate stages and hadrons in final stage. It includes Poisson-transform distributions (with the partially coherent distribution as a special case) and is very flexible for phenomenological analyses. The associated Koba, Nielson, and Olesen limit and the behavior of cumulant moments are analyzed in detail for finite and/or infinite cluster size and particle size per cluster. In general, a supercluster distribution does not need to be equivalent to a negative binomial distribution to fit experimental data well. Furthermore, the requirement of such equivalence leads to many solutions, in which the average size of the cluster is not logarithmic: e.g., it may show a power behavior instead. Superclustering is defined as a two-or multi-stage process underlying observed global multiplicity distributions. At the first stage of the production process, individual clusters are produced according to a given statistical law. For example, the clustering distribution may be described by partially coherent (oreven sub-Poissonian distribution models. At the second stage, the clusters are considered as the sources of particle production. The corresponding distribution may then be as general as the clustering distribution just mentioned. 8 refs

Flux distribution in phantom for biomedical use of beam-type thermal neutrons

International Nuclear Information System (INIS)

Aoki, Kazuhiko; Kobayashi, Tooru; Kanda, Keiji; Kimura, Itsuro

1985-01-01

For boron neutron capture therapy, the thermal neutron beam is worth using as therapeutic neutron irradiation without useless and unfavorable exposure of normal tissues around tumor and for microanalysis system to measure ppm-order 10 B concentrations in tissue and to search for the location of the metastasis of tumor. In the present study, the thermal neutron flux distribution in a phantom, when beam-type thermal neutrons were incident on it, was measured at the KUR Neutron Guide Tube. The measurements were carried out by two different methods using indium foil. The one is an ordinary foil activation technique by using the 115 In(n, γ) 116m 1 In reactions, while the other is to detect γ-rays from the 115 In(n, γ) 116m 2 In reactions during neutron irradiations with a handy-type Ge detector. The calculations with DOT 3.5 were performed to examine thermal neutron flux in the phantom for various beam size and phantom size. The experimental and calculated results are in good agreement and it is shown that the second type measurement has a potential for practical application as a new monitoring system of the thermal neutron flux in a living body for boron neutron capture therapy. (author)

Angular distribution of scission neutrons studied with time-dependent Schrödinger equation

Science.gov (United States)

Wada, Takahiro; Asano, Tomomasa; Carjan, Nicolae

2018-03-01

We investigate the angular distribution of scission neutrons taking account of the effects of fission fragments. The time evolution of the wave function of the scission neutron is obtained by integrating the time-dependent Schrodinger equation numerically. The effects of the fission fragments are taken into account by means of the optical potentials. The angular distribution is strongly modified by the presence of the fragments. In the case of asymmetric fission, it is found that the heavy fragment has stronger effects. Dependence on the initial distribution and on the properties of fission fragments is discussed. We also discuss on the treatment of the boundary to avoid artificial reflections

Neutron chain length distributions in subcritical systems

International Nuclear Information System (INIS)

Nolen, S.D.; Spriggs, G.

1999-01-01

In this paper, the authors present the results of the chain-length distribution as a function of k in subcritical systems. These results were obtained from a point Monte Carlo code and a three-dimensional Monte Carlo code, MC++. Based on these results, they then attempt to explain why several of the common neutron noise techniques, such as the Rossi-α and Feynman's variance-to-mean techniques, are difficult to perform in highly subcritical systems using low-efficiency detectors

Neutron distribution in the central cell and a peripheral cell of the Fontenay-aux-Roses pile

International Nuclear Information System (INIS)

Roullier, F.

1958-01-01

The distribution of the neutron density has been determined in the central cell and a peripheral cell of the pile at Fontenay-aux-Roses. This measurement was carried out by the autoradiographic method with manganese detectors. The neutron density distribution in the uranium rod has already been studied. The measurement was completed by the study of the neutron density in the complete cell by means of detectors placed in the uranium and in the heavy water. (author) [fr

Neutron detection and multiplicity counting using a boron-loaded plastic scintillator/bismuth germanate phoswich detector array

International Nuclear Information System (INIS)

Miller, M.C.

1998-03-01

Neutron detection and multiplicity counting has been investigated using a boron-loaded plastic scintillator/bismuth germanate phoswich detector array. Boron-loaded plastic combines neutron moderation (H) and detection ( 10 B) at the molecular level, thereby physically coupling increasing detection efficiency and decreasing die-away time with detector volume. Both of these characteristics address a fundamental limitation of thermal-neutron multiplicity counters, where 3 He proportional counters are embedded in a polyethylene matrix. Separation of the phoswich response into its plastic scintillator and bismuth germanate components was accomplished on a pulse-by-pulse basis using custom integrator and timing circuits. In addition, a custom time-tag module was used to provide a time for each detector event. Analysis of the combined energy and time event stream was performed by calibrating each detector's response and filtering based on the presence of a simultaneous energy deposition corresponding to the 10 B(n,alpha) reaction products in the plastic scintillator (93 keV ee ) and the accompanying neutron-capture gamma ray in the bismuth germanate (478 keV). Time-correlation analysis was subsequently performed on the filtered event stream to obtain shift-register-type singles and doubles count rates. Proof-of-principle measurements were conducted with a variety of gamma-ray and neutron sources including 137 Cs, 54 Mn, AmLi, and 252 Cf. Results of this study indicate that a neutron-capture probability of ∼10% and a die-away time of ∼10 micros are possible with a 4-detector array with a detector volume of 1600 cm 3 . Simulations were performed that indicate neutron-capture probabilities on the order of 50% and die-away times of less than 4 micros are realistically achievable. While further study will be required for practical application of such a detection system, the results obtained in this investigation are encouraging and may lead to a new class of high

Spin assignments of nuclear levels above the neutron binding energy in $^{88}$Sr

CERN Multimedia

Neutron resonances reveal nuclear levels in the highly excited region of the nucleus around the neutron binding energy. Nuclear level density models are therefore usually calibrated to the number of observed levels in neutron-induced reactions. The gamma-ray cascade from the decay of the highly excited compound nucleus state to the ground state show dierences dependent on the initial spin. This results in a dierence in the multiplicity distribution which can be exploited. We propose to use the 4${\\pi}$ total absorption calorimeter (TAC) at the n TOF facility to determine the spins of resonances formed by neutrons incident on a metallic $^{87}$Sr sample by measuring the gamma multiplicity distributions for the resolved resonances. In addition we would like to use the available enriched $^{87}$Sr target for cross section measurements with the C$\\scriptscriptstyle{6}$D$\\scriptscriptstyle{6}$ detector setup.

Neutron detection time distributions of multisphere LiI detectors and AB rem meter at a 20 MeV electron linac

International Nuclear Information System (INIS)

Liu, J.C.; Rokni, S.; Vylet, V.; Arora, R.; Semones, E.; Justus, A.

1997-01-01

Neutron detection time distribution is an important factor for the dead-time correction for moderator type neutron detectors used in pulsed radiation fields. Measurements of the neutron detection time distributions of multisphere LiL detectors (2''3'' , 5'', 8'', 10'' and 12'' in diameter) and an AB rem meter were made inside an ANL 20 MeV electron linac room. Calculations of the neutron detection time distributions were also made using Monte Carlo codes. The first step was to calculate the neutron energy spectra at the target and detector positions, using a coupled EGS4-MORSE code with a giant-resonant photoneutron generation scheme. The calculated detector spectrum was found in agreement with the multisphere measurements. Then, neutrons hitting the detector surface were scored as a function of energy and the travel time in the room using MCNP. Finally, the above neutron fluence as a function of energy and travel time was used as the source term, and the neutrons detected by 6 Li or 10 B in the sensor were scored as a function of detection time for each detector using MCNP. The calculations of the detection time distributions agree with the measurements. The results also show that the detection time distributions of detectors with large moderators depend mainly on the moderator thickness and neutron spectrum. However, for small detectors, the neutron travel time in the field is also crucial. Therefore, all four factors (neutron spectrum, neutron travel time in the field, detector moderator thickness and detector response function) may play inter-related roles in the detection time distribution of moderator type detectors. (Author)

Goodness of isospin in neutron rich systems from the fission fragment distribution

Science.gov (United States)

Garg, Swati; Jain, Ashok Kumar

2017-09-01

We present the results of our calculations for the relative yields of neutron-rich fission fragments emitted in 208Pb (18O, fission) reaction by using the concept of the conservation of isospin and compare with the experimental data. We take into account a range of isospin values allowed by the isospin algebra and assume that the fission fragments are formed in isobaric analog states. We also take into account the neutron multiplicity data for various neutron-emission channels in each partition, and use them to obtain the weight factors in calculating the yields. We then calculate the relative yields of the fission fragments. Our calculated results are able to reproduce the experimental trends reasonably well. This is the first direct evidence of the isospin conservation in neutron-rich systems and may prove a very useful tool in their studies.

Determination of nuclear friction in strongly damped reactions from prescission neutron multiplicities

NARCIS (Netherlands)

Wilczynski, J; SiwekWilczynska, K; Wilschut, HW

Nonfusion, fissionlike reactions in collisions of four heavy systems (well below the fusion extra-push energy threshold), Mr which Hinde and co-workers had measured the prescission neutron multiplicities, have been analyzed in terms of the deterministic dynamic model of Feldmeier coupled to a

Angular Distributions of Neutrons from (p,n)-Reactions in some Mirror Nuclei

Energy Technology Data Exchange (ETDEWEB)

Stroemberg, L G; Wiedling, T; Holmqvist, B

1964-04-15

The angular distributions of neutrons from the reactions {sup 13}C (p,n) {sup 13}N and {sup 19}F (p, n) {sup 19}Ne have been measured for some energies close to the reactions threshold. For the reaction {sup 9}B (p, n) {sup 9}C angular distributions have been measured at several proton energies below the reaction threshold of the neutrons to the first excited state in {sup 11}C . A 5.5 MeV Van de Graaff has been used for the experiments. The neutrons were detected with a long counter. The measurements were carried out for 16 energies for the B (p,n) reaction, 3 energies for the C (p, n) reaction, and for 7 energies for the F 19(p, n) reaction. One of the main reasons for investigating these (p, n) reactions was to check whether the direct reaction process is important at low proton energies as well as close to reaction thresholds in nuclei consisting of closed shells of neutrons and protons either with an extra nucleon outside the closed shell or a nucleon hole. Comparisons with a theory proposed by Bloom, Glendenning, and Moszkowski have been performed.

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

Monte carlo sampling of fission multiplicity.

Energy Technology Data Exchange (ETDEWEB)

Hendricks, J. S. (John S.)

2004-01-01

Two new methods have been developed for fission multiplicity modeling in Monte Carlo calculations. The traditional method of sampling neutron multiplicity from fission is to sample the number of neutrons above or below the average. For example, if there are 2.7 neutrons per fission, three would be chosen 70% of the time and two would be chosen 30% of the time. For many applications, particularly {sup 3}He coincidence counting, a better estimate of the true number of neutrons per fission is required. Generally, this number is estimated by sampling a Gaussian distribution about the average. However, because the tail of the Gaussian distribution is negative and negative neutrons cannot be produced, a slight positive bias can be found in the average value. For criticality calculations, the result of rejecting the negative neutrons is an increase in k{sub eff} of 0.1% in some cases. For spontaneous fission, where the average number of neutrons emitted from fission is low, the error also can be unacceptably large. If the Gaussian width approaches the average number of fissions, 10% too many fission neutrons are produced by not treating the negative Gaussian tail adequately. The first method to treat the Gaussian tail is to determine a correction offset, which then is subtracted from all sampled values of the number of neutrons produced. This offset depends on the average value for any given fission at any energy and must be computed efficiently at each fission from the non-integrable error function. The second method is to determine a corrected zero point so that all neutrons sampled between zero and the corrected zero point are killed to compensate for the negative Gaussian tail bias. Again, the zero point must be computed efficiently at each fission. Both methods give excellent results with a negligible computing time penalty. It is now possible to include the full effects of fission multiplicity without the negative Gaussian tail bias.

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)

Passive assay of plutonium metal plates using a fast-neutron multiplicity counter

Energy Technology Data Exchange (ETDEWEB)

Di Fulvio, A., E-mail: difulvio@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Shin, T.H.; Jordan, T.; Sosa, C.; Ruch, M.L.; Clarke, S.D. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Pozzi, S.A. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

2017-05-21

We developed a fast-neutron multiplicity counter based on organic scintillators (EJ-309 liquid and stilbene). The system detects correlated photon and neutron multiplets emitted by fission reactions, within a gate time of tens of nanoseconds. The system was used at Idaho National Laboratory to assay a variety of plutonium metal plates. A coincidence counting strategy was used to quantify the {sup 240}Pu effective mass of the samples. Coincident neutrons, detected within a 40-ns coincidence window, show a monotonic trend, increasing with the {sup 240}Pu-effective mass (in this work, we tested the 0.005–0.5 kg range). After calibration, the system estimated the {sup 240}Pu effective mass of an unknown sample ({sup 240}Pu{sub eff} >50 g) with an uncertainty lower than 1% in a 4-min assay time.

Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, La Reina, Santiago (Chile); Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, La Reina, Santiago (Chile)

2016-07-07

Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work, we present the unfolding results using the EM algorithm.

Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

Science.gov (United States)

Faghihi, F.; Mehdizadeh, S.; Hadad, K.

Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

Experimental determination of neutron temperature distribution in reactor cell; Eksperimentalno odredjivanje raspodele neutronske temperature u celiji reaktorske resetke

Energy Technology Data Exchange (ETDEWEB)

Bosevski, T [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-12-15

This paper describes theoretical preparation of the experiment for measuring neutron temperature distribution at the RB reactor by activation foils. Due to rather low neutron flux Cu and Lu foil were irradiated for 4 days. Special natural uranium fuel element was prepared to enable easy removal of foils after irradiation. Experimental device was placed in the reactor core at half height in order to measure directly the mean neutron density. Experimental data of neutron temperature distribution for square lattice pitch 16 cm are presented with mean values of neutron temperature in the moderator, in the fuel and on the fuel element surface.

Moments of the very high multiplicity distributions

International Nuclear Information System (INIS)

Nechitailo, V.A.

2004-01-01

In experiment, the multiplicity distributions of inelastic processes are truncated due to finite energy, insufficient statistics, or special choice of events. It is shown that the moments of such truncated multiplicity distributions possess some typical features. In particular, the oscillations of cumulant moments at high ranks and their negative values at the second rank can be considered as ones most indicative of the specifics of these distributions. They allow one to distinguish between distributions of different type

A diffusion-theoretical method to calculate the neutron flux distribution in multisphere configurations

International Nuclear Information System (INIS)

Schuerrer, F.

1980-01-01

For characterizing heterogene configurations of pebble-bed reactors the fine structure of the flux distribution as well as the determination of the macroscopic neutronphysical quantities are of interest. When calculating system parameters of Wigner-Seitz-cells the usual codes for neutron spectra calculation always neglect the modulation of the neutron flux by the influence of neighbouring spheres. To judge the error arising from that procedure it is necessary to determinate the flux distribution in the surrounding of a spherical fuel element. In the present paper an approximation method to calculate the flux distribution in the two-sphere model is developed. This method is based on the exactly solvable problem of the flux determination of a point source of neutrons in an infinite medium, which contains a spherical perturbation zone eccentric to the point source. An iteration method allows by superposing secondary fields and alternately satisfying the conditions of continuity on the surface of each of the two fuel elements to advance to continually improving approximations. (orig.) 891 RW/orig. 892 CKA [de

The neutron star mass distribution

Energy Technology Data Exchange (ETDEWEB)

Kiziltan, Bülent [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Kottas, Athanasios; De Yoreo, Maria [Department of Applied Mathematics and Statistics, University of California, Santa Cruz, CA 95064 (United States); Thorsett, Stephen E., E-mail: bkiziltan@cfa.harvard.edu [Department of Astronomy and Astrophysics, University of California and UCO/Lick Observatory, Santa Cruz, CA 95064 (United States)

2013-11-20

In recent years, the number of pulsars with secure mass measurements has increased to a level that allows us to probe the underlying neutron star (NS) mass distribution in detail. We critically review the radio pulsar mass measurements. For the first time, we are able to analyze a sizable population of NSs with a flexible modeling approach that can effectively accommodate a skewed underlying distribution and asymmetric measurement errors. We find that NSs that have evolved through different evolutionary paths reflect distinctive signatures through dissimilar distribution peak and mass cutoff values. NSs in double NS and NS-white dwarf (WD) systems show consistent respective peaks at 1.33 M {sub ☉} and 1.55 M {sub ☉}, suggesting significant mass accretion (Δm ≈ 0.22 M {sub ☉}) has occurred during the spin-up phase. The width of the mass distribution implied by double NS systems is indicative of a tight initial mass function while the inferred mass range is significantly wider for NSs that have gone through recycling. We find a mass cutoff at ∼2.1 M {sub ☉} for NSs with WD companions, which establishes a firm lower bound for the maximum NS mass. This rules out the majority of strange quark and soft equation of state models as viable configurations for NS matter. The lack of truncation close to the maximum mass cutoff along with the skewed nature of the inferred mass distribution both enforce the suggestion that the 2.1 M {sub ☉} limit is set by evolutionary constraints rather than nuclear physics or general relativity, and the existence of rare supermassive NSs is possible.

Coherent deeply virtual Compton scattering off 3He and neutron generalized parton distributions

Directory of Open Access Journals (Sweden)

Rinaldi Matteo

2014-06-01

Full Text Available It has been recently proposed to study coherent deeply virtual Compton scattering (DVCS off 3He nuclei to access neutron generalized parton distributions (GPDs. In particular, it has been shown that, in Impulse Approximation (IA and at low momentum transfer, the sum of the quark helicity conserving GPDs of 3He, H and E, is dominated by the neutron contribution. This peculiar result makes the 3He target very promising to access the neutron information. We present here the IA calculation of the spin dependent GPD H See Formula in PDF of 3He. Also for this quantity the neutron contribution is found to be the dominant one, at low momentum transfer. The known forward limit of the IA calculation of H See Formula in PDF , yielding the polarized parton distributions of 3He, is correctly recovered. The extraction of the neutron information could be anyway non trivial, so that a procedure, able to take into account the nuclear effects encoded in the IA analysis, is proposed. These calculations, essential for the evaluation of the coherent DVCS cross section asymmetries, which depend on the GPDs H,E and H See Formula in PDF , represent a crucial step for planning possible experiments at Jefferson Lab.

Properties of truncated multiplicity distributions

International Nuclear Information System (INIS)

Lupia, S.

1995-01-01

Truncation effects on multiplicity distributions are discussed. Observables sensitive to the tail, like factorial moments, factorial cumulants and their ratio, are shown to be strongly affected by truncation. A possible way to overcome this problem by looking at the head of the distribution is suggested. (author)

Properties of truncated multiplicity distributions

Energy Technology Data Exchange (ETDEWEB)

Lupia, S. [Turin Univ. (Italy). Dipt. di Fisica

1995-12-31

Truncation effects on multiplicity distributions are discussed. Observables sensitive to the tail, like factorial moments, factorial cumulants and their ratio, are shown to be strongly affected by truncation. A possible way to overcome this problem by looking at the head of the distribution is suggested. (author)

Three-dimensional neutron dose distribution in the environment around a 1-GeV electron synchrotron facility at INS

International Nuclear Information System (INIS)

Uwamino, Y.; Nakamura, T.

1987-01-01

The three-dimensional (surface and altitude) skyshine neutron-dose-equivalent distribution around the 1-GeV electron synchrotron (ES) of the Institute for Nuclear Study, University of Tokyo, was measured with a high-sensitivity dose-equivalent counter. The neutron spectrum in the environment was also measured with a multimoderator spectrometer incorporating a 3 He counter. The dose-equivalent distribution and the leakage neutron spectrum at the surface of the ES building were measured with a Studsvik 2202D counter and the multimoderator spectrometer, including an indium activation detector. Skyshine neutron transport calculations, beginning with the photoneutron spectrum and yielding the dose-equivalent distribution in the environment, were performed with the DOT3.5 code and two Monte Carlo codes, MMCR-2 and MMCR-3, using the DLC-87/HILO group cross sections. The calculated neutron spectra at the top surface of the concrete ceiling and at a point 111 m from the ES agreed well with the measured results, and the calculated three-dimensional dose-equivalent distribution also agreed. The dose value increased linearly with altitude, and the slope was estimated for neutron-producing facilities. (author)

The radial distribution of the neutron field in the core of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Determination of the radial distribution of the thermal neutron field in the core of the Dalat reactor is done by the Cu foil activation method. The measured data are fitted by the least square method to determine several physical parameters of the reactor, as follows: 1. Buckling B{sub r}{sup 2}=(84.6{+-}5.5)10{sup -4}/cm{sup 2}. 2. The effective radius R{sub eff}=(27.6{+-}1.0)cm. 3. The extrapolation distance {lambda}=(8.7{+-}1.0)cm. 4. The unequal coefficient of the effective multiplication K{sub r}=1.77{+-}0.11. (author). 2 refs., 4 figs., 1 tab.

Investigation of distribution of elements in a Korean ginseng by using a neutron activation method

Energy Technology Data Exchange (ETDEWEB)

Lee, Yu Na; Sun, Gwang Min; Chung, Yong Sam; Kim, Young Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-10-15

The Distinction of production areas of Korean ginsengs has been tried by using neutron activation techniques such as an instrumental neutron activation analysis (INAA) and a prompt gamma activation analysis (PGAA). This study was done as a part of those efforts. As is well known, the distribution of elements varies according to the part of plant due to the difference of enrichment effect and influence from a soil where the plants have been grown. So a correlation study between plants and soil is an important issue. In this study, the distribution of trace elements within a Korean ginseng was investigated by using an instrumental neutron activation analysis.

Accelerator-driven neutron sources for materials research

International Nuclear Information System (INIS)

Jameson, R.A.

1990-01-01

Particle accelerators are important tools for materials research and production. Advances in high-intensity linear accelerator technology make it possible to consider enhanced neutron sources for fusion material studies or as a source of spallation neutrons. Energy variability, uniformity of target dose distribution, target bombardment from multiple directions, time-scheduled dose patterns, and other features can be provided, opening new experimental opportunities. New designs have also been used to ensure hands-on maintenance on the accelerator in these factory-type facilities. Designs suitable for proposals such as the Japanese Energy-Selective Intense Neutron Source, and the international Fusion Materials Irradiation Facility are discussed

Neutron multicounter detector for investigation of content and spatial distribution of fission materials in large volume samples

International Nuclear Information System (INIS)

Swiderska-Kowalczyk, M.; Starosta, W.; Zoltowski, T.

1998-01-01

The experimental device is a neutron coincidence well counter. It can be applied for passive assay of fissile - especially for plutonium bearing - materials. It consist of a set of 3 He tubes placed inside a polyethylene moderator; outputs from the tubes, first processed by preamplifier/amplifier/discriminator circuits, are then analysed using neutron correlator connected with a PC, and correlation techniques implemented in software. Such a neutron counter allows for determination of plutonium mass ( 240 Pu effective mass) in nonmultiplying samples having fairly big volume (up to 0.14 m 3 ). For determination of neutron sources distribution inside the sample, the heuristic methods based on hierarchical cluster analysis are applied. As an input parameters, amplitudes and phases of two-dimensional Fourier transformation of the count profiles matrices for known point sources distributions and for the examined samples, are taken. Such matrices are collected by means of sample scanning by detection head. During clustering process, counts profiles for unknown samples fitted into dendrograms using the 'proximity' criterion of the examined sample profile to standard samples profiles. Distribution of neutron sources in an examined sample is then evaluated on the basis of comparison with standard sources distributions. (author)

Novel applications of Lattice QCD: Parton Distributions, proton charge radius and neutron electric dipole moment

Directory of Open Access Journals (Sweden)

Alexandrou Constantia

2017-01-01

Full Text Available We briefly discuss the current status of lattice QCD simulations and review selective results on nucleon observables focusing on recent developments in the lattice QCD evaluation of the nucleon form factors and radii, parton distribution functions and their moments, and the neutron electric dipole moment. Nucleon charges and moments of parton distribution functions are presented using simulations generated at physical values of the quark masses, while exploratory studies are performed for the parton distribution functions and the neutron electric dipole moment at heavier than physical value of the pion mass.

Proposal for a new method of reactor neutron flux distribution determination

Energy Technology Data Exchange (ETDEWEB)

Popic, V R [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-01-15

A method, based on the measurements of the activity produced in a medium flowing with variable velocity through a reactor, for the determination of the neutron flux distribution inside a reactor is considered theoretically (author)

Calculation of neutron radiation energy deposition distribution in subcellular parts of tissue using recombination chamber microdosimetry

International Nuclear Information System (INIS)

Golnik, N.; Zielczynski, M.

1999-01-01

Recombination chamber microdosimetry was used as an instrument for determination of local neutron radiation energy deposition distribution. The method allows to simulate of subcellular regions of tissue of the order of 70 nm in size. The results obtained qualitatively correspond to relationship between biological efficiency and neutron energy, and show regular differences of distributions achieved by the recombination method and distributions measured using tissue equivalent proportional counters (TEPC), which simulates greater tissue regions of 1 μm in size

Parametric analysis of neutron streaming through major penetrations in the 0.914 m TFTR test cell floor

International Nuclear Information System (INIS)

Ku, L.P.; Liew, S.L.; Kolibal, J.G.

1985-09-01

Neutron streaming through penetrations in the 0.914 m TFTR test cell floor has two distinct features: (1) the oblique angle of incidence; and (2) the high order of anisotropy in the angular distribution for incident neutrons with energies > 10 keV. The effects of these features on the neutron streaming into the TFTR basement were studied parametrically for isolated penetrations. Variations with respect to the source energies, angular distributions, and sizes of the penetrations were made. The results form a data base from which the spatial distribution of the neutron flux in the basement due to multiple penetrations may be evaluated

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

International Nuclear Information System (INIS)

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

1983-01-01

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

A Monte Carlo Study on the Effect of Various Neutron Capturers on Dose Distribution in Brachytherapy with 252Cf Source

Directory of Open Access Journals (Sweden)

Firoozabadi M. M.

2017-03-01

Full Text Available Background: In neutron interaction with matter and reduction of neutron energy due to multiple scatterings to the thermal energy range, increasing the probability of thermal neutron capture by neutron captures makes dose enhancement in the tumors loaded with these materials. Objective: The purpose of this study is to evaluate dose distribution in the presence of 10B, 157Gd and 33S neutron capturers and to determine the effect of these materials on dose enhancement rate for 252Cf brachytherapy source. Methods: Neutron-ray flux and energy spectra, neutron and gamma dose rates and dose enhancement factor (DEF are determined in the absence and presence of 10B, 157Gd and 33S using Monte Carlo simulation. Results: The difference in the thermal neutron flux rate in the presence of 10B and 157Gd is significant, while the flux changes in the fast and epithermal energy ranges are insensible. The dose enhancement factor has increased with increasing distance from the source and reached its maximum amount equal to 258.3 and 476.1 cGy/h/µg for 157Gd and 10B, respectively at about 8 cm distance from the source center. DEF for 33S is equal to one. Conclusion: Results show that the magnitude of dose augmentation in tumors containing 10B and 157Gd in brachytherapy with 252Cf source will depend not only on the capture product dose level, but also on the tumor distance from the source. 33S makes dose enhancement under specific conditions that these conditions depend on the neutron energy spectra of source, the 33S concentration in tumor and tumor distance from the source.

Discerning the neutron density distribution of 208Pb from nucleon elastic scattering

International Nuclear Information System (INIS)

Karataglidis, S.; Amos, K.; University of Melbourne, VIC; Brown, B.A.; Deb, P.K.

2001-01-01

We seek a measure of the neutron density of 208 Pb from analyses of intermediate energy nucleon elastic scattering. The pertinent model for such analyses is based on coordinate space nonlocal optical potentials obtained from model nuclear ground state densities. As a calibration of the use of Skyrme-Hartree-Fock models the elastic scattering from 40 Cawas considered as well. Those potentials give predictions of integral observables and of angular distributions which show sensitivity to the neutron density. When compared with experiment, and correlated with analyses of electron scattering data, the results suggest that 208 Pb has a neutron skin thickness ∼ 0.17 fm

Neutron monitor generated data distributions in quantum variational Monte Carlo

Science.gov (United States)

Kussainov, A. S.; Pya, N.

2016-08-01

We have assessed the potential applications of the neutron monitor hardware as random number generator for normal and uniform distributions. The data tables from the acquisition channels with no extreme changes in the signal level were chosen as the retrospective model. The stochastic component was extracted by fitting the raw data with splines and then subtracting the fit. Scaling the extracted data to zero mean and variance of one is sufficient to obtain a stable standard normal random variate. Distributions under consideration pass all available normality tests. Inverse transform sampling is suggested to use as a source of the uniform random numbers. Variational Monte Carlo method for quantum harmonic oscillator was used to test the quality of our random numbers. If the data delivery rate is of importance and the conventional one minute resolution neutron count is insufficient, we could always settle for an efficient seed generator to feed into the faster algorithmic random number generator or create a buffer.

Simulation of neutron multiplicity measurements using Geant4. Open source software for nuclear arms control

Energy Technology Data Exchange (ETDEWEB)

Kuett, Moritz

2016-07-07

Nuclear arms control, including nuclear safeguards and verification technologies for nuclear disarmament typically use software as part of many different technological applications. This thesis proposes to use three open source criteria for such software, allowing users and developers to have free access to a program, have access to the full source code and be able to publish modifications for the program. This proposition is presented and analyzed in detail, together with the description of the development of ''Open Neutron Multiplicity Simulation'', an open source software tool to simulate neutron multiplicity measurements. The description includes physical background of the method, details of the developed program and a comprehensive set of validation calculations.

Violence of heavy-ion reactions from neutron multiplicity: 11 to 20A MeV /sup 20/Ne+ /sup 238/U

International Nuclear Information System (INIS)

Jahnke, U.; Ingold, G.; Hilscher, D.; Lehmann, M.; Schwinn, E.; Zank, P.

1986-01-01

The suitability of the neutron multiplicity as a gauge for the violence of medium-energy heavy-ion reactions is investigated for the first time. For this purpose the number of neutrons emitted from fission reactions induced by 220-, 290-, and 400-MeV /sup 20/Ne on /sup 238/U is registered event-by-event with a large 4π scintillator tank. It is shown that the neutron multiplicity is indeed closely related to the two quantities characterizing the violence: the induced total intrinsic excitation and the linear momentum transfer

Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

Science.gov (United States)

Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

2010-10-01

The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

Negative binomial multiplicity distribution from binomial cluster production

International Nuclear Information System (INIS)

Iso, C.; Mori, K.

1990-01-01

Two-step interpretation of negative binomial multiplicity distribution as a compound of binomial cluster production and negative binomial like cluster decay distribution is proposed. In this model we can expect the average multiplicity for the cluster production increases with increasing energy, different from a compound Poisson-Logarithmic distribution. (orig.)

Personnel neutron dosimetry applications of track-size distributions on electrochemically etched CR-39 foils

International Nuclear Information System (INIS)

Hankins, D.E.; Homann, S.G.; Westermark, J.

1988-01-01

The track-size distribution on electrochemically etched CR-39 foils can be used to obtain some limited information on the incident neutron spectra. Track-size distributions on CR-39 foils can also be used to determine if the tracks were caused by neutrons or if they are merely background tracks (which have a significantly different track-size distribution). Identifying and discarding the high-background foils reduces the number of foils that must be etched. This also lowers the detection limit of the dosimetry system. We have developed an image analyzer program that can more efficiently determine the track density and track-size distribution, as well as read the laser-cut identification numbers on each foil. This new image analyzer makes the routine application of track-size distributions on CR-39 foils feasible. 2 refs., 3 figs

Sensitivity analysis of physical/operational parameters in neutron multiplicity counting

International Nuclear Information System (INIS)

Peerani, P.; Marin Ferrer, M.

2007-01-01

In this paper, we perform a sensitivity study on the influence of various physical and operational parameters on the results of neutron multiplicity counting. The purpose is to have a better understanding of the importance of each component and its contribution to the measurement uncertainty. Then we will be able to determine the optimal conditions for the operational parameters and for detector design and as well to point out weaknesses in the knowledge of critical fundamental nuclear data

Multiple scattering and attenuation corrections in Deep Inelastic Neutron Scattering experiments

International Nuclear Information System (INIS)

Dawidowski, J; Blostein, J J; Granada, J R

2006-01-01

Multiple scattering and attenuation corrections in Deep Inelastic Neutron Scattering experiments are analyzed. The theoretical basis of the method is stated, and a Monte Carlo procedure to perform the calculation is presented. The results are compared with experimental data. The importance of the accuracy in the description of the experimental parameters is tested, and the implications of the present results on the data analysis procedures is examined

Uranium mass and neutron multiplication factor estimates from time-correlation coincidence counts

Energy Technology Data Exchange (ETDEWEB)

Xie, Wenxiong [China Academy of Engineering Physics, Center for Strategic Studies, Beijing 100088 (China); Li, Jiansheng [China Academy of Engineering Physics, Institute of Nuclear Physics and Chemistry, Mianyang 621900 (China); Zhu, Jianyu [China Academy of Engineering Physics, Center for Strategic Studies, Beijing 100088 (China)

2015-10-11

Time-correlation coincidence counts of neutrons are an important means to measure attributes of nuclear material. The main deficiency in the analysis is that an attribute of an unknown component can only be assessed by comparing it with similar known components. There is a lack of a universal method of measurement suitable for the different attributes of the components. This paper presents a new method that uses universal relations to estimate the mass and neutron multiplication factor of any uranium component with known enrichment. Based on numerical simulations and analyses of 64 highly enriched uranium components with different thicknesses and average radii, the relations between mass, multiplication and coincidence spectral features have been obtained by linear regression analysis. To examine the validity of the method in estimating the mass of uranium components with different sizes, shapes, enrichment, and shielding, the features of time-correlation coincidence-count spectra for other objects with similar attributes are simulated. Most of the masses and multiplications for these objects could also be derived by the formulation. Experimental measurements of highly enriched uranium castings have also been used to verify the formulation. The results show that for a well-designed time-dependent coincidence-count measuring system of a uranium attribute, there are a set of relations dependent on the uranium enrichment by which the mass and multiplication of the measured uranium components of any shape and size can be estimated from the features of the source-detector coincidence-count spectrum.

Contributions to the theory of fission neutron emission

International Nuclear Information System (INIS)

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

1990-03-01

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

On the spherical harmonic expansion of the neutron angular distribution function

Energy Technology Data Exchange (ETDEWEB)

Depken, Sven

1959-03-15

The neutron (one-velocity) angular distribution function is expanded in terms of spherical harmonic tensors. The solution to the equations of the moments is given explicitly and the result is applied to the plane, spherical and cylinder symmetrical cases.

On the spherical harmonic expansion of the neutron angular distribution function

International Nuclear Information System (INIS)

Depken, Sven

1959-03-01

The neutron (one-velocity) angular distribution function is expanded in terms of spherical harmonic tensors. The solution to the equations of the moments is given explicitly and the result is applied to the plane, spherical and cylinder symmetrical cases

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

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1993-01-01

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

Used Fuel Cask Identification through Neutron Profile

Energy Technology Data Exchange (ETDEWEB)

Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-20

Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

Measurement and analysis of neutron flux distribution of STACY heterogeneous core by position sensitive proportional counter. Contract research

CERN Document Server

Murazaki, M; Uno, Y

2003-01-01

We have measured neutron flux distribution around the core tank of STACY heterogeneous core by position sensitive proportional counter (PSPC) to develop the method to measure reactivity for subcritical systems. The neutron flux distribution data in the position accuracy of +-13 mm have been obtained in the range of uranium concentration of 50g/L to 210g/L both in critical and in subcritical state. The prompt neutron decay constant, alpha, was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutron flux and sup 3 He reaction rates at the location of PSPC by using continuous energy Monte Carlo code MCNP. The measurement data was compared with the calculation results. As results of comparison, calculated values agreed generally with measurement data of PSPC with Cd cover in the region above half of solution height, but the difference between calculated value and measurement data was large in the region below half of solution height. On the other hand, ...

Sensitivity Analysis of Core Neutronic Parameters in Electron Accelerator-driven Subcritical Advanced Liquid Metal Reactor

Directory of Open Access Journals (Sweden)

Marziye Ebrahimkhani

2016-02-01

Full Text Available Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (Ee and source multiplication coefficient (ks, has been investigated. A Monte Carlo code (MCNPX_2.6 has been used to calculate neutronic parameters such as effective multiplication coefficient (keff, net neutron multiplication (M, neutron yield (Yn/e, energy constant gain (G0, energy gain (G, importance of neutron source (φ∗, axial and radial distributions of neutron flux, and power peaking factor (Pmax/Pave in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current (Ie have been decreased in the highest case of ks, but G and φ∗ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing Ee from 100 MeV up to 1 GeV, Yn/e and G improved by 91.09% and 10.21%, and Ie and Pacc decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np–Pu assemblies on the periphery allows for a consistent keff because the Np–Pu assemblies experience less burn-up.

Bench-mark experiments to study the neutron distribution in a heterogeneous reactor shielding

International Nuclear Information System (INIS)

Bolyatko, V.V.; Vyrskij, M.Yu.; Mashkovich, V.P.; Nagaev, R.Kh.; Prit'mov, A.P.; Sakharov, V.K.; Troshin, V.S.; Tikhonov, E.G.

1981-01-01

The bench-mark experiments performed at the B-2 facility of the BR-10 reactor to investigate the spatial and energy neutron distributions are described. The experimental facility includes the neutron beam channel with a slide, a mo shielding composition investigated consisted of sequential layers of steel (1KH18N9T) and graphite slabs. The neutron spectra were measured by activation method, a set of treshold and resonance detectors having been used. The detectors made it possible to obtain the absolute neutron spectra in the 1.4 eV-10 MeV range. The comparison of calculations with the results of the bench-mark experiments made it possible to prove the neutron transport calculational model realized in the ROZ-9 and ARAMAKO-2F computer codes and evaluate the validity of the ARAMAKO constants for the class of shielding compositions in question [ru

Prompt neutron emission from fragments in spontaneous fission of {sup 244,248}Cm and {sup 252}Cf

Energy Technology Data Exchange (ETDEWEB)

Vorobyev, A. S.; Shcherbakov, O. A. [Petersburg Nuclear Physics Institute, Gatchina, Leningrad district, 188300 (Russian Federation); Dushin, V. N.; Jakovlev, V. A.; Kalinin, V. A.; Petrov, B. F. [V.G. Khlopin Radium Institute, St. Petersburg, 194021 (Russian Federation); Hambsch, F.J [EC-JRC-Institute for Reference Materials and Measurements Retieseweg 111, B-2440 Geel (Belgium); Laptev, A. B. [Petersburg Nuclear Physics Institute, Gatchina, Leningrad district, 188300 (Russian Federation); Japan Nuclear Cycle Development Institute, Tokai-mura, Naka-gun, Ibaraki 319-1194 (Japan)

2005-07-01

Neutrons emitted in fission were measured separately for each complementary fragment in correlation with fission fragment energies. Two high efficiency Gd-loaded liquid scintillator tanks were used for neutron registration. Fission fragment energies were measured using a twin Frisch gridded ionization chamber with a pin-hole collimator. The neutron multiplicity distributions were obtained for each value of the fission fragment mass and energy and corrected for neutron registration efficiency, background and pile-up. The dependencies of these distributions on fragment mass and energy for different energy and mass bins, as well as the mass and energy distribution of the fission fragments are presented and discussed. (authors)

Optimization of neutron flux distribution in Isotope Production Reactor

International Nuclear Information System (INIS)

Valladares, G.L.

1988-01-01

In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely the Vmod / Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO 2 ) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author) [pt

Structure and Spatial Distribution of Ge Nanocrystals Subjected to Fast Neutron Irradiation

Directory of Open Access Journals (Sweden)

Alexander N. Ionov

2011-07-01

Full Text Available The influence of fast neutron irradiation on the structure and spatial distribution of Ge nanocrystals (NC embedded in an amorphous SiO2 matrix has been studied. The investigation was conducted by means of laser Raman Scattering (RS, High Resolution Transmission Electron Microscopy (HR-TEM and X-ray photoelectron spectroscopy (XPS. The irradiation of Ge- NC samples by a high dose of fast neutrons lead to a partial destruction of the nanocrystals. Full reconstruction of crystallinity was achieved after annealing the radiation damage at 8000C, which resulted in full restoration of the RS spectrum. HR-TEM images show, however, that the spatial distributions of Ge-NC changed as a result of irradiation and annealing. A sharp decrease in NC distribution towards the SiO2 surface has been observed. This was accompanied by XPS detection of Ge oxides and elemental Ge within both the surface and subsurface region.

Determining neutron multiplication factor in the infinite system by reactivity dependence on one dimension of the reactor core

International Nuclear Information System (INIS)

Pesic, M.

1975-01-01

The objective of this task was to apply Fermi age theory for determining τ and neutron multiplication factor in infinite medium by measuring reactivity coefficient of heavy water in heterogeneous mixed reactor lattice. Basis of experiment is the measurement of stable reactor period. Measurement of heavy water reactivity coefficient by measuring the stable reactor period is described for chosen overcritical heavy water levels. Calculated values of infinite multiplication factor for measured neutron age data are presented and they are compared to expected theoretical values

Measurement of energy deposition distributions produced in cylindrical geometry by irradiation with 15 MeV neutrons

International Nuclear Information System (INIS)

Brandan, M.E.

1979-01-01

Cellular survival experiments have shown that the biological damage induced by radiation depends on the density of energy deposition along the trajectory of the ionizing particle. The quantity L is defined to measure the density of energy transfer along a charged particle's trajectory. It is equal to sigma/l, where sigma is the energy transferred to a medium and l is the path length along which the transfer takes place. L is the stochastic quantity whose mean value is the unrestricted linear energy transfer, L/sub infinity/. Measurements of the distribution of L in a thin medium by secondary charged particles from fast neutron irradiation were undertaken. A counter operating under time coincidence between two coaxial cylindrical detectors was designed and built for this purpose. Secondary charged particles enter a gas proportional counter and deposit some energy sigma. Those particles traversing the chamber along a radial trajectory strike a CsI scintillator. A coincidence between both detectors' signals selects a known path length for these events, namely the radius of the cavity. Measurements of L distributions for l = 1 μm in tissue were obtained for 3 and 15 MeV neutron irradiation of a tissue-equivalent target wall and for 15 MeV neutron irradiation of a graphite wall. Photon events were corrected for by measurements with a Pb target wall and 15 MeV neutron irradiation as well as exposure to a pure photon field. The measured TE wall distributions with 15 MeV neutron bombardment show contributions from protons, α-particles, 9 Be and 12 C recoils. The last three comprise the L distribution for irradiation of the graphite wall. The proton component of the measured L distributions at 3 and 15 MeV was compared to calculated LET distributions

Determination of energy distribution for photon and neutron microdosimetry

International Nuclear Information System (INIS)

Todo, A.S.

1989-01-01

This work was undertaken to provide basic physical data for use in both microdosimetry and dosimetry of high energy photons and also in the neutron radiation field. It is described the formalism to determine the initial electron energy spectra in water irradiated by photons with energies up to 1 GeV. Calculations were performed with a Monte Carlo computer code, PHOEL-3, which is also described. The code treats explicitly the production of electron-positron pairs, Compton scattering, photoelectric absorption, and the emission of Auger electrons following the occurrence of K-shell vacancies in oxygen. The tables give directly the information needed to specify the absolute single-collision kerma in water, which approximates tissue, at each photon energy. Results for continuous photon energy spectra can be obtained by using linear interpolation with the tables. The conditions under which first-collision kerma approximate absorbed dose are discussed. A formula is given for estimating bremsstrahlung energy loss, one of the principal differences between kerma and absorbed dose in practical case. A study has been carried out, on the use of cylindrical, energy-proportional pulse-height detector for determining microdosimetric quantities, as neutron fractional dose spectra, D (L), in function of linear energy transfer, TLE. In the present study the Hurst detector was used and this device satisfies the requirement of the Bragg-Gray principle. It is developed a Monte Carlo Method to obtain the D(L) spectrum from a measured pulse-height spectrum H(h), and the knowledge of the distribution of recoil-particle track lenght, P(T) in the sensitive volume of the detector. These developed programs to find P(T) and D(L) are presented. The distribution of D(L) in LET were obtained using a known distribution of P(T) and the measured H(h) spectrum from sup(252)Cf neutron source. All the results are discussed and the conclusions are presented. (author)

Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

International Nuclear Information System (INIS)

Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2013-01-01

Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

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

Uncovering the Density of Matter from Multiplicity Distribution

International Nuclear Information System (INIS)

Bialas, A.

2010-01-01

Multiplicity distributions in the form of superposition of Poisson distributions which are observed in multiparticle production are interpreted as reflection of a two-step nature of this process: the creation and evolution of the strongly interacting fluid, followed by its uncorrelated decay into observed hadrons. A method to uncover the density of the fluid from the observed multiplicity distribution is described. (author)

Charged particle multiplicity distributions in Au-Au collisions at RHIC-BNL energies (BRAHMS Experiment)

International Nuclear Information System (INIS)

Argintaru, D.; Bearden, I.G.; Beavis, D.

2002-01-01

The BRAHMS Experiment (Broad RAnge Hadronic Magnetic Spectrometers) takes place at the Relativistic Heavy Ion Collider (RHIC) from Brookhaven National Laboratory and searches for a transition of matter into a new phase called quark-gluon plasma (QGP), a phase transition that appears in extreme conditions of nuclear matter densities and temperatures. Important signals for this transition are related to dependencies of the charged particle multiplicity distributions on the pseudorapidity range. The charged particle multiplicity distributions in Au-Au collisions at RHIC energies are obtained from the global detector measurements. These detectors are: - An array of Si strip detectors and scintillation tiles placed around the interaction region; they cover the range η < 2 in pseudorapidity, measuring the majority of charged particles; - Two systems of Cerenkov detectors (Beam-Beam Counters) placed both sides of the nominal interaction point at 220 cm and cover the range 3 < η < 4.3 in pseudorapidity. These detectors are used for vertex determination and supply a level zero trigger for the entire experiment; - Zero Degree Calorimeters placed at zero degree with respect to the beam axis, both sides of the vertex, measuring the spectator neutrons from the nuclear reactions. These detectors supplies information about the reaction centrality and could estimate the interaction vertex. The paper presents some results on charged particle multiplicities in different pseudorapidity ranges at different impact parameters. Interesting dependencies of the average charged particle multiplicities on the pseudorapidity range, impact parameters and total available energy in the centre of mass system. Some comparisons with the simulation codes predictions and theoretical model estimations are included, too. (authors)

The influence of the energy distribution of workplace fields on neutron personal dosemeter reading

International Nuclear Information System (INIS)

Tanner, R.J.; Thomas, D.J.; Bartlett, D.T.; Hager, L.G.; Horwood, N.A.

2002-01-01

Variations in the energy dependence of response of neutron personal dosemeters cause systematic errors in the readings obtained in workplace fields. The magnitude of these errors has been determined theoretically by folding measured and calculated workplace energy distributions with dosemeter response functions, to determine the response of a given personal dosemeter in that field. These results have been analysed with consideration of the dosemeter response to various calibration spectra, and with reference to different workplaces. The dosemeters in the study are discussed in terms of the workplaces for which they can be suitably calibrated. Deficiencies in the published neutron energy distributions are identified

Neutron detector for detecting rare events of spontaneous fission

International Nuclear Information System (INIS)

Ter-Akop'yan, G.M.; Popeko, A.G.; Sokol, E.A.; Chelnokov, L.P.; Smirnov, V.I.; Gorshkov, V.A.

1981-01-01

The neutron detector for registering rare events of spontaneous fission by detecting multiple neutron emission is described. The detector represents a block of plexiglas of 550 mm diameter and 700 mm height in the centre of which there is a through 160 mm diameter channel for the sample under investigation. The detector comprises 56 3 He filled counters (up to 7 atm pressure) with 1% CO 2 addition. The counters have a 500 mm length and a 32 mm diameter. The sampling of fission events is realized by an electron system which allows determining the number of detected neutrons, numbers of operated counters, signal amplitude and time for fission event detecting. A block diagram of a neutron detector electron system is presented and its operation principle is considered. For protection against cosmic radiation the detector is surronded by a system of plastic scintillators and placed behind the concrete shield of 6 m thickness. The results of measurements of background radiation are given. It has been found that the background radiation of single neutron constitutes about 150 counts per hour, the detecting efficiency of single neutron equals 0.483 +- 0.005, for a 10l detector sensitive volume. By means of the detector described the parameters of multiplicity distribution of prompt neutrons for 256 Fm spontaneous fission are measured. The average multiplicity equals 3.59+-0.06 the dispersion being 2.30+-0.65

Study of particle size distribution and formation mechanism of radioactive aerosols generated in high-energy neutron fields

CERN Document Server

Endo, A; Noguchi, H; Tanaka, S; Iida, T; Furuichi, S; Kanda, Y; Oki, Y

2003-01-01

The size distributions of sup 3 sup 8 Cl, sup 3 sup 9 Cl, sup 8 sup 2 Br and sup 8 sup 4 Br aerosols generated by irradiations of argon and krypton gases containing di-octyl phthalate (DOP) aerosols with 45 MeV and 65 MeV quasi-monoenergetic neutrons were measured in order to study the formation mechanism of radioactive particles in high energy radiation fields. The effects of the size distribution of the radioactive aerosols on the size of the added DOP aerosols, the energy of the neutrons and the kinds of nuclides were studied. The observed size distributions of the radioactive particles were explained by attachment of the radioactive atoms generated by the neutron-induced reactions to the DOP aerosols. (author)

Mass distributions in monoenergetic-neutron-induced fission of 232Th

International Nuclear Information System (INIS)

Glendenin, L.E.; Gindler, J.E.; Ahmad, I.; Henderson, D.J.; Meadows, J.W.

1980-01-01

Fission product yields for 38 masses were determined for the fission of 232 Th with essentially monoenergetic neutrons of 2.0, 3.0, 4.0, 5.9, 6.4, 6.9, 7.6, and 8.0 MeV. Fission product activities were measured by Ge(Li) γ-ray spectrometry of irradiated 232 Th foils and by chemical separation of the fission product elements followed by β counting. The mass yield data for 232 Th(n,f ) show a sensitive increase of fission yields in the near-symmetric mass region (valley) with increasing incident neutron energy E/sub n/ and a pronounced dip in yield at the onset of second-chance fission just above the neutron binding energy (at approx. 6 MeV) where the excitation energy is lowered by competition with neutron evaporation prior to fission. The effect of second-chance fission is also seen in the yields of asymmetric peak products. A distinct third peak is observed at symmetry in the valley of the mass distribution, and enhanced yields are observed in the asymmetric peaks at masses associated with even Z (proton pairing effect). The fission yeilds of 232 Th(n,f ) are compared with those of 238 U(n,f ) and 232 Th

Neutron emission effects on final fragments mass and kinetic energy distribution from low energy fission of 234U

International Nuclear Information System (INIS)

Montoya, M.; Rojas, J.; Lobato, I.

2008-01-01

The standard deviation of the final kinetic energy distribution (σ e ) as a function of mass of final fragments (m) from low energy fission of 234 U, measured with the Lohengrin spectrometer by Belhafaf et al., presents a peak around m = 109 and another around m = 122. The authors attribute the first peak to the evaporation of a large number of neutrons around the corresponding mass number, i.e. there is no peak on the standard deviation of the primary kinetic energy distribution (σ E ) as a function of primary fragment mass (A). The second peak is attributed to a real peak on σ E (A). However, theoretical calculations related to primary distributions made by H.R. Faust and Z. Bao do not suggest any peak on σ E (A). In order to clarify this apparent controversy, we have made a numerical experiment in which the masses and the kinetic energy of final fragments are calculated, assuming an initial distribution of the kinetic energy without structures on the standard deviation as function of fragment mass. As a result we obtain a pronounced peak on σ e (m) curve around m = 109, a depletion from m = 121 to m = 129, and an small peak around m = 122, which is not as great as that measured by Belhafaf et al. Our simulation also reproduces the experimental results on the yield of the final mass Y(m), the average number of emitted neutrons as a function of the provisional mass (calculated from the values of the final kinetic energy of the complementary fragments) and the average value of fragment kinetic energy as a function of the final mass. From our results we conclude that there are no peaks on the σ E (A) curve, and the observed peaks on σ e (m) are due to the emitted neutron multiplicity and the variation of the average fragment kinetic energy as a function of primary fragment mass. (Author)

Effect of double false pulses in calibrated neutron coincidence collar during measuring time-correlated neutrons from PuBe neutron sources

Energy Technology Data Exchange (ETDEWEB)

Nguyen, Tam Cong, E-mail: tam.nguyen.cong@energia.mta.hu; Huszti, Jozsef; Nguyen, Quan Van

2015-09-01

Effect of double false pulses of preamplifiers in neutron coincidence collar was investigated to explain non-parallel shape of calibrated D/S–M{sub Pu} curves of two commercial neutron coincidence collars, JCC-31 and JCC-13. Two curves, which were constructed from D/S ratio (doubles and singles count rate), and Pu content M{sub Pu}, of the same set of secondary standard PuBe neutron sources, should be parallel. Non-parallelism rises doubt about usability of the method based on this curve for determination of Pu content in PuBe neutron sources. We have shown in three steps that the problem originates from double false pulses of preamplifiers in JCC-13. First we used a pulse train diagram for analyzing the non-parallel shape, second we used Rossi-Alpha distribution measured by pulse train recorder developed in our institute and finally, we investigated the effect of inserted noise pulses. This implies a new type of QA test option in traditional multiplicity shift registers for excluding presence of double false pulses.

Restoration of γ-ray multiplicity distributions from experiments with low efficiency multiplicity filters

International Nuclear Information System (INIS)

Bellia, G.; Del Zoppo, A.; Migneco, E.; Russo, G.; Istituto Nazionale di Fisica Nucleare, Catania

1984-01-01

The restoration of γ-ray multiplicity distributions from experimental p-fold coincidence distributions is discussed. It is shown that the restoration of the multiplicity from measurements with low total detection efficiency is an 'incorrectly posed problem'. While in the literature the analysis of the experimental data has been attempted only in terms of the lowest central moments of the multiplicity distribution, in this paper an unfolding method based on the minimization of the directioned discrepancies in the probability space is used. The method is found to work very well even if the total efficiency Ω <= 0.1. Realistic tests and a comparison with the usual method of analysis are presented. (orig.)

On a possible use of multiple Bragg reflections for high-resolution monochromatization of neutrons

Czech Academy of Sciences Publication Activity Database

Mikula, Pavol; Vrána, Miroslav; Wagner, V.

2004-01-01

Roč. 350, - (2004), e667-e670 ISSN 0921-4526 R&D Projects: GA ČR GA202/03/0891 Keywords : neutron diffraction * multiple reflections * higg-resolution monochromator Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.679, year: 2004

Flux distribution by neutrons semi-conductors detectors during the startup of the EL4 reactor

International Nuclear Information System (INIS)

Fuster, S.; Tarabella, A.

1967-01-01

The Cea developed neutron semi-conductors detectors which allows a quasi-instantaneous monitoring of neutrons flux distribution, when placed in a reactor during the tests. These detectors have been experimented in the EL4 reactor. The experiment and the results are presented and compared with reference mappings. (A.L.B.)

Measurement and analysis of neutron flux distribution of STACY heterogeneous core by position sensitive proportional counter. Contract research

Energy Technology Data Exchange (ETDEWEB)

Murazaki, Minoru; Uno, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

We have measured neutron flux distribution around the core tank of STACY heterogeneous core by position sensitive proportional counter (PSPC) to develop the method to measure reactivity for subcritical systems. The neutron flux distribution data in the position accuracy of {+-}13 mm have been obtained in the range of uranium concentration of 50g/L to 210g/L both in critical and in subcritical state. The prompt neutron decay constant, {alpha}, was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutron flux and {sup 3}He reaction rates at the location of PSPC by using continuous energy Monte Carlo code MCNP. The measurement data was compared with the calculation results. As results of comparison, calculated values agreed generally with measurement data of PSPC with Cd cover in the region above half of solution height, but the difference between calculated value and measurement data was large in the region below half of solution height. On the other hand, calculated value agreed well with measurement data of PSPC without Cd cover. (author)

Dynamic radial distribution function from inelastic neutron scattering

International Nuclear Information System (INIS)

McQueeney, R.J.

1998-01-01

A real-space, local dynamic structure function g(r,ω) is defined from the dynamic structure function S(Q,ω), which can be measured using inelastic neutron scattering. At any particular frequency ω, S(Q,ω) contains Q-dependent intensity oscillations which reflect the spatial distribution and relative displacement directions for the atoms vibrating at that frequency. Information about local and dynamic atomic correlations is obtained from the Fourier transform of these oscillations g(r,ω) at the particular frequency. g(r,ω) can be formulated such that the elastic and frequency-summed limits correspond to the average and instantaneous radial distribution function, respectively, and is thus called the dynamic radial distribution function. As an example, the dynamic radial distribution function is calculated for fcc nickel in a model which considers only the harmonic atomic displacements due to phonons. The results of these calculations demonstrate that the magnitude of the atomic correlations can be quantified and g(r,ω) is a well-defined correlation function. This leads to a simple prescription for investigating local lattice dynamics. copyright 1998 The American Physical Society

Determination of the hexagonal network parameters of the quartz β using neutron multiple diffraction

International Nuclear Information System (INIS)

Campos, L.C.; Parente, C.B.R.; Mazzocchi, V.L.; Helene, O.

2000-01-01

In this work, neutron multiple diffraction is employed for the determination of the parameters a and c of the β-quartz hexagonal cell. This crystalline phase of silica (SiO 2 ) occurs in temperatures between ca. 846 and 1143 K. A β-quartz neutron multiple diffraction pattern has been used in the determinations. This pattern was obtained with a natural quartz single crystal heated to 1003 K. During the indexing of the pattern it was verified that most of the pairs of secondary reflections, which are responsible for the formation of peaks, could be classified as 'good for the determination of a' or 'good for the determination of c'. With this classification, it became possible to employ an iterative method for the determination of both parameters. After 8 cycles of iteration the values found for the parameters were a = 4.9964 +- 0.0018 and c = 5.46268 +- 0.00052 A. (author)

Determination of the neutron energy and spatial distributions of the neutron beam from the TSR-II in the large beam shield

International Nuclear Information System (INIS)

Clifford, C.E.; Muckenthaler, F.J.

1976-01-01

The TSR-II reactor of the ORNL Tower Shielding Facility has recently been relocated within a new, fixed shield. A principal feature of the new shield is a beam port of considerably larger area than that of its predecessor. The usable neutron flux has thereby been increased by a factor of approximately 200. The bare beam neutron spectrum behind the new shield has been experimentally determined over the energy range from 0.8 to 16 MeV. A high level of fission product gamma ray background prevented measurement of bare beam spectra below 0.8 MeV, however neutron spectra in the energy range from 8 keV to 1.4 MeV were obtained for two simple, calculable shielding configurations. Also measured in the present work were weighted integral flux distributions and fast neutron dose rates

Estimation of the {alpha} particles and neutron distribution generated during a fusion reaction; Evaluation de la distribution des particules {alpha} et des neutrons issus de la reaction de fusion

Energy Technology Data Exchange (ETDEWEB)

Dellacherie, S.

1997-12-01

The respective distributions (or density probabilities) of {alpha} particles and neutrons have been modeled using a Monte-Carlo method for the thermonuclear fusion reaction D + T {yields} {alpha} + n + 17.6 MeV. (N.T.).

Multiplicity distributions in inelastic reactions on nuclei

CERN Document Server

Caneschi, L; Schwimmer, A

1976-01-01

The multiplicity distribution of the number of knocked-out nucleons and the correlation of the former with the multiplicity of the produced mesons, in inelastic particle-nucleus scattering, are computed.

A verification scenario of nuclear plus interference scattering effects using neutron incident angle distribution to the wall in beam-injected deuterium plasmas

International Nuclear Information System (INIS)

Sugiyama, Shota; Matsuura, Hideaki; Uchiyama, Daisuke; Sawada, Daisuke; Watanabe, Tsuguhiro; Goto, Takuya; Mitarai, Osamu

2015-01-01

A verification scenario of knock-on tail formation in the deuteron distribution function due to nuclear plus interference scattering is presented by observing the incident angle distribution of neutrons in a vacuum vessel. Assuming a knock-on tail created in a "3He-beam-injected deuterium plasma, the incident angle distribution and energy spectra of the neutrons produced by fusion reactions between 1-MeV and thermal deuterons are evaluated. The relation between the neutron incident angle to the vacuum vessel and neutron energy is examined in the case of anisotropic neutron emission due to knock-on tail formation in neutral-beam-injected plasmas. (author)

Medium dependence of multiplicity distributions in MLLA

International Nuclear Information System (INIS)

Armesto, Nestor; Pajares, Carlos; Quiroga-Arias, Paloma

2009-01-01

We study the modification of the multiplicity distributions in MLLA due to the presence of a QCD medium. The medium is introduced through a multiplicative constant (f med ) in the soft infrared parts of the kernels of the QCD evolution equations. Using the asymptotic ansatz for mean multiplicities of the quark and gluons, left angle n G right angle =e γy and left angle n Q right angle =r -1 e γy , respectively, we study two cases: fixed γ as previously considered in the literature, and fixed α s . We find opposite behaviors of the dispersion of the multiplicity distributions with increasing f med in both cases. For fixed γ the dispersion decreases, while for fixed α s it increases. (orig.)

Extensive Air Showers Detected by Aragats Neutron Monitor

International Nuclear Information System (INIS)

Badalyan, A.; Chilingarian, A.; Hovsepyan, G.; Grigoryan, A.; Khanikyants, Y.; Manukyan, A.; Pokhsraryan, D.; Soghomonyan, S.

2017-01-01

Extensive Air Shower (EAS) duration as registered by the surface particle detectors does not exceed a few tens of nanosecond. However, Neutron monitors containing plenty of absorbing matter can respond to EAS core traversal during 1 ∼ms by registering secondary slow neutrons born by EAS hadrons in the soil, walls of buildings and in the matter of detector itself. Thus, the time distribution of the pulses from the proportional counters of the neutron monitor after EAS propagation extends to ∼l ms, ∼5 orders of magnitude larger than the EAS passing time. The Aragats Neutron Monitor (ArNM) has a special option for the EAS core detection. In general, the dead time of NM is ∼1 ms that provides the one-to-one relation of incident hadrons and detector counts. The pulses generated by the neutrons possibly entering the proportional chamber after the first one will be neglected. In ArNM, we use several “electronic” dead times, and with the shortest one, 400 ns, the detector counts all pulses that enter the proportional chambers. If ArNM one-second time series corresponding to the shortest dead time contain much more signals (a neutron burst) than with l-ms dead time, then we conclude that the EAS core hits the detector. We assume that he distribution of registered burst multiplicities is proportional to the energy of the primary particle. The primary cosmic ray energy spectrum was obtained by the frequency analysis through the counting frequencies of the multiplicities of different magnitudes and relating them to the integral energy spectrum measured by the MAKET array at the same place several years ago. (author)

Neutron Production by Muon Spallation I: Theory

International Nuclear Information System (INIS)

Luu, T; Hagmann, C

2006-01-01

We describe the physics and codes developed in the Muon Physics Package. This package is a self-contained Fortran90 module that is intended to be used with the Monte Carlo package MCNPX. We calculate simulated energy spectra, multiplicities, and angular distributions of direct neutrons and pions from muon spallation

The influence of a spatial displacement of hydrogen on the reactivity and neutron flux density distribution in a ZrH-moderated reactor

International Nuclear Information System (INIS)

Doehler, J.; Bartsch, G.

1975-08-01

The effect of changes of the hydrogen concentration in uranium zirconium hydride resulting from spatially varying temperatures on the reactivity and neutron flux distribution of the BER-II core (power 2.2 MW) are shown. Furthermore, in general, the influence of the hydrogen concentration on important reactor parameters of a fuel cell of BER-II is calculated and presented. A comparison of the diffusion calculation with spatially constant hydrogen concentrations shows a decrease of the thermal neutron flux density in regions with a low hydrogen content (high temperature) and inversely an increase for high hydrogen concentrations. Furthermore, a change of the effective multiplication factor by 0.6% was found in the case of a spatially varying hydrogen concentration as compared with the calculation for a constant concentration. (orig.) [de

Determination and analysis of neutron flux distribution on radial Piercing beam port for utilization of Kartini research reactor

International Nuclear Information System (INIS)

Widarto

2002-01-01

Determination and analysis of neutron flux measurements on radial piercing beam port have been done as completion experimental data document and progressing on utilization of the Kartini research reactor purposes. The analysis and determination of the neutron flux have been carried out by using Au foils detector neutron activation analysis method which put on the radius of cross section (19 cm) and a long of radial piercing beam port (310 cm) Based on the calculation, distribution of the thermal neutron flux is around (8.3 ± 0.9) x 10 5 ncm -2 s -1 to (6.8 ± 0.5) x 10 7 ncm -2 s -1 and fast neutron is (5.0 ± 0.2) x 10 5 ncm -2 s -1 to (1.43 ± 0.6) x 10 7 ncm -2 s -1 . Analyzing by means of curve fitting method could be concluded that the neutron flux distribution on radial piercing beam port has profiled as a polynomial curve. (author)

Measurements of {sup 237}Np secondary neutron spectra

Energy Technology Data Exchange (ETDEWEB)

Kornilov, N.V.

1997-03-01

The activities carried out during the first year of the project are summarized. The main problems for Np spectra measurements arise from high intrinsic gamma-ray activity of the sample and admixture of the oxygen and iron nuclei. The inelastically scattered neutrons and the fission neutrons spectra for {sup 237}Np were measured by time-of-flight spectrometer of the IPPE at incident neutron energies {approx_equal}1.5 MeV, and {approx_equal}0.5 MeV. A solid tritium target and a Li-metallic target were used as neutron sources. The neutron scattering on C sample (C(n,n) standard reaction) was measured to normalize the Np data. The experimental data should be simulated by Monte Carlo method to correct the experimental data for oxygen and iron admixture as well as for multiple scattering of the neutrons in the sample. Therefore the response function of the spectrometer, and the neutron energy distribution from the source were investigated in detail. (author)

Measurement of angular distribution of neutrons emitted from plasma focus using NTD

International Nuclear Information System (INIS)

Antanasijevic, R.; Maric, Z.; Banjanac, R.; Dragic, A.; Stanojevic, J.; Dordevic, D.; Joksimovic, D.; Udovicic, V.; Vukovic, J.

1999-01-01

Mica detectors together with thick uranium foils were used for detection and angular distribution measurement of the neutrons emitted from the d-plasma focus (DPF). The distance between detectors and plasma pinch was 4 cm. For reason of detector protection from the thermal shock, they were covered with the mica layer of 1 mm thickness. Annealing of the detectors were measured after the irradiation with neutrons from an Am-Be source and treated with thermal shock H-plasma focus (HPF). Geometry of detectors in both experiments was the same

Measurement of angular distribution of neutrons emitted from plasma focus using NTD

Energy Technology Data Exchange (ETDEWEB)

Antanasijevic, R.; Maric, Z.; Banjanac, R.; Dragic, A.; Stanojevic, J.; Dordevic, D.; Joksimovic, D.; Udovicic, V.; Vukovic, J

1999-06-01

Mica detectors together with thick uranium foils were used for detection and angular distribution measurement of the neutrons emitted from the d-plasma focus (DPF). The distance between detectors and plasma pinch was 4 cm. For reason of detector protection from the thermal shock, they were covered with the mica layer of 1 mm thickness. Annealing of the detectors were measured after the irradiation with neutrons from an Am-Be source and treated with thermal shock H-plasma focus (HPF). Geometry of detectors in both experiments was the same.

Measurement of angular distribution of neutrons emitted from plasma focus using NTD

CERN Document Server

Antanasijevic, R; Banjanac, R; Dragic, A; Stanojevic, J; Dordevic, D; Joksimovic, D; Udovicic, V; Vukovic, J

1999-01-01

Mica detectors together with thick uranium foils were used for detection and angular distribution measurement of the neutrons emitted from the d-plasma focus (DPF). The distance between detectors and plasma pinch was 4 cm. For reason of detector protection from the thermal shock, they were covered with the mica layer of 1 mm thickness. Annealing of the detectors were measured after the irradiation with neutrons from an Am-Be source and treated with thermal shock H-plasma focus (HPF). Geometry of detectors in both experiments was the same.

Analysis of the neutron generation from a D-Li neutron source

International Nuclear Information System (INIS)

Gomes, I.

1994-02-01

The study of the neutron generation from the D-Li reaction is an important issue to define the optimum combination of the intervening parameters during the design phase of a D-Li neutron source irradiation facility. The major players in defining the neutron yield from the D-Li reaction are the deuteron incident energy and the beam current, provided that the lithium target is thick enough to stop all incident deuterons. The incident deuteron energy also plays a role on the angular distribution of the generated neutrons, on the energy distribution of the generated neutrons, and on the maximum possible energy of the neutrons. The D-Li reaction produces neutrons with energies ranging from eV's to several MeV's. The angular distribution of these neutrons is dependent on the energy of both, incident deuterons and generated neutrons. The deuterons lose energy interacting with the lithium target material in such a way that the energy of the deuterons inside the lithium target varies from the incident deuteron energy to essentially zero. The first part of this study focuses in analyzing the neutron generation rate from the D-Li reaction as a function of the intervening parameters, in defining the source term, in terms of the energy and angular distributions of the generated neutrons, and finally in providing some insights of the impact of varying input parameters on the generation rate and correlated distributions. In the second part an analytical description of the Monte Carlo sampling procedure of the neutron from the D-Li reaction is provided with the aim at further Monte Carlo transport of the D-Li neutrons

Scaling neutron absorbed dose distributions from one medium to another

International Nuclear Information System (INIS)

Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

1983-01-01

Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone tissue-equivalent (TE) solutions, mineral oil, and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU Report No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. The OARs measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. Therefore, neutron beam CADDs and OARs may be measured in either TE solution (USA practice) or water (European practice), and having determined the respective scaling lengths, all measurements may be scaled from one medium to any other. It is recommended that for general treatment planning purposes, scaling be made to TE muscle with a density of 1.04 g cm -3 , since this value represents muscle and other soft tissues better than TE solution of density 1.07 g cm -3 . For such a transformation, relative measurements made in water are found to require very small corrections. Hence, it is further recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. Finally, a table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

SLAROM, Neutron Flux Distribution and Spectra in Lattice Cell

International Nuclear Information System (INIS)

Nakagawa, M.; Tsuchihashi, K.

2002-01-01

1 - Description of program or function: SLAROM solves the neutron integral transport equations to determine flux distribution and spectra in a lattice and calculates cell averaged effective cross sections. 2 - Method of solution: Collision probability method for cell calculation and 1D diffusion for core calculation. 3 - Restrictions on the complexity of the problem: Variable dimensions are used throughout the program so that computer core requirements depend on a variety of program parameters

Jet multiplicity distributions: medium dependence in MLLA

International Nuclear Information System (INIS)

Armesto, Nestor; Pajares, Carlos; Quiroga-Arias, Paloma

2009-01-01

We study the medium dependence of the multiplicity distributions in the modified leading logarithmic approximation. We focus in the enhancement in the number of branchings as the partons travel trough a dense medium created in a heavy-ion collision. We study the effect of a higher number of splittings in some jet observables by introducing the medium as a constant (f med ) in the splitting functions. Having as our ansatz for the quark and gluon jets mean multiplicities left angle n G right angle =e γy and left angle n Q right angle =r -1 e γy , we study in an analytic approach the dependence with the medium (f med ) of the anomalous dimension (γ), the multiplicity ratio (r), and so the mean multiplicities. We also obtain the higher-order moments of the multiplicity distribution, what allows us to study its dispersion. (orig.)

Shielding implications for secondary neutrons and photons produced within the patient during IMPT

Energy Technology Data Exchange (ETDEWEB)

DeMarco, J.; Kupelian, P.; Santhanam, A.; Low, D. [UCLA Department of Radiation Oncology, University of California Los Angeles, Los Angeles, California 90095 (United States)

2013-07-15

Purpose: Intensity modulated proton therapy (IMPT) uses a combination of computer controlled spot scanning and spot-weight optimized planning to irradiate the tumor volume uniformly. In contrast to passive scattering systems, secondary neutrons and photons produced from inelastic proton interactions within the patient represent the major source of emitted radiation during IMPT delivery. Various published studies evaluated the shielding considerations for passive scattering systems but did not directly address secondary neutron production from IMPT and the ambient dose equivalent on surrounding occupational and nonoccupational work areas. Thus, the purpose of this study was to utilize Monte Carlo simulations to evaluate the energy and angular distributions of secondary neutrons and photons following inelastic proton interactions within a tissue-equivalent phantom for incident proton spot energies between 70 and 250 MeV.Methods: Monte Carlo simulation methods were used to calculate the ambient dose equivalent of secondary neutrons and photons produced from inelastic proton interactions in a tissue-equivalent phantom. The angular distribution of emitted neutrons and photons were scored as a function of incident proton energy throughout a spherical annulus at 1, 2, 3, 4, and 5 m from the phantom center. Appropriate dose equivalent conversion factors were applied to estimate the total ambient dose equivalent from secondary neutrons and photons.Results: A reference distance of 1 m from the center of the patient was used to evaluate the mean energy distribution of secondary neutrons and photons and the resulting ambient dose equivalent. For an incident proton spot energy of 250 MeV, the total ambient dose equivalent (3.6 Multiplication-Sign 10{sup -3} mSv per proton Gy) was greatest along the direction of the incident proton spot (0 Degree-Sign -10 Degree-Sign ) with a mean secondary neutron energy of 71.3 MeV. The dose equivalent decreased by a factor of 5 in the

Medium dependence of multiplicity distributions in MLLA

Energy Technology Data Exchange (ETDEWEB)

Armesto, Nestor; Pajares, Carlos; Quiroga-Arias, Paloma [Universidade de Santiago de Compostela, Departamento de Fisica de Particulas and IGFAE, Santiago de Compostela (Spain)

2009-06-15

We study the modification of the multiplicity distributions in MLLA due to the presence of a QCD medium. The medium is introduced through a multiplicative constant (f{sub med}) in the soft infrared parts of the kernels of the QCD evolution equations. Using the asymptotic ansatz for mean multiplicities of the quark and gluons, left angle n{sub G} right angle =e{sup {gamma}}{sup y} and left angle n{sub Q} right angle =r {sup -1}e{sup {gamma}}{sup y}, respectively, we study two cases: fixed {gamma} as previously considered in the literature, and fixed {alpha}{sub s}. We find opposite behaviors of the dispersion of the multiplicity distributions with increasing f{sub med} in both cases. For fixed {gamma} the dispersion decreases, while for fixed {alpha}{sub s} it increases. (orig.)

Measurements of neutron flux distributions in the core of the Ljubljana TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Rant, J.; Ravnik, M.; Mele, I.; Dimic, V.

2008-01-01

Recently the Ljubljana TRIGA Mark II Reactor has been refurbished and upgraded to pulsed operation. To verify the core design calculations using TRIGAP and PULSTR1 codes and to obtain necessary data for future irradiation and neutron beam experiments, an extensive experimental program of neutron flux mapping and neutron field characterization was carried out. Using the existing neutron measuring thimbles complete axial and radial distributions in two radial directions were determined for two different core configurations. For one core configuration the measurements were also carried out in the pulsed mode. For flux distributions thin Cu (relative measurements) and diluted Au wires (absolute values) were used. For each radial position the cadmium ratio was determined in two axial levels. The core configuration was rather uniform, well defined (fresh fuel of a single type, including fuelled followers) and compact (no irradiation channels or gaps), offering unique opportunity to test the computer codes for TRIGA reactor calculations. The neutron flux measuring procedures and techniques are described and the experimental results are presented. The agreement between the predicted and measured power peaking factors are within the error limits of the measurements (<±5%) and calculations (±10%). Power peaking occurs in the B ring, and in the A ring (centre) there is a significant flux depression. (authors)

Monte Carlo analyses of the source multiplication factor of the YALINA booster facility

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto; Gohar, Y.; Kondev, F.; Aliberti, Gerardo [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Bolshinsky, I. [Idaho National Laboratory, P. O. Box 2528, Idaho Falls, Idaho 83403 (United States); Kiyavitskaya, Hanna; Bournos, Victor; Fokov, Yury; Routkovskaya, Christina; Serafimovich, Ivan [Joint Institute for Power and Nuclear Research-Sosny, National Academy of Sciences, Minsk, acad. Krasin, 99, 220109 (Belarus)

2008-07-01

The multiplication factor of a subcritical assembly is affected by the energy spectrum and spatial distribution of the neutron source. In a critical assembly, neutrons emerge from the fission reactions with an average energy of approx2 MeV; in a deuteron accelerator driven subcritical assembly, neutrons emerge from the fusion target with a fixed energy of 2.45 or 14.1 MeV, from the Deuterium-Deuterium (D-D) and Deuterium-Tritium (D-T) reactions respectively. This study aims at generating accurate neutronics models for the YALINA Booster facility, based on the use of different Monte Carlo neutron transport codes, at defining the facility key physical parameters, and at comparing the neutron multiplication factor for three different neutron sources: fission, D-D and D-T. The calculated values are compared with the experimental results. (authors)

Monte Carlo analyses of the source multiplication factor of the YALINA booster facility

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Y.; Kondev, F.; Aliberti, Gerardo; Bolshinsky, I.; Kiyavitskaya, Hanna; Bournos, Victor; Fokov, Yury; Routkovskaya, Christina; Serafimovich, Ivan

2008-01-01

The multiplication factor of a subcritical assembly is affected by the energy spectrum and spatial distribution of the neutron source. In a critical assembly, neutrons emerge from the fission reactions with an average energy of ∼2 MeV; in a deuteron accelerator driven subcritical assembly, neutrons emerge from the fusion target with a fixed energy of 2.45 or 14.1 MeV, from the Deuterium-Deuterium (D-D) and Deuterium-Tritium (D-T) reactions respectively. This study aims at generating accurate neutronics models for the YALINA Booster facility, based on the use of different Monte Carlo neutron transport codes, at defining the facility key physical parameters, and at comparing the neutron multiplication factor for three different neutron sources: fission, D-D and D-T. The calculated values are compared with the experimental results. (authors)

Neutron emission effects on final fragments mass and kinetic energy distribution from low energy fission of {sup 234}U

Energy Technology Data Exchange (ETDEWEB)

Montoya, M.; Rojas, J. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Lobato, I. [Facultad de Ciencias, Universidad Nacional de Ingenieria, Av. Tupac Amaru 210, Apartado Postal 31-139, Lima (Peru)]. e-mail: mmontoya@ipen.gob.pe

2008-07-01

The standard deviation of the final kinetic energy distribution ({sigma}{sub e}) as a function of mass of final fragments (m) from low energy fission of {sup 234}U, measured with the Lohengrin spectrometer by Belhafaf et al., presents a peak around m = 109 and another around m = 122. The authors attribute the first peak to the evaporation of a large number of neutrons around the corresponding mass number, i.e. there is no peak on the standard deviation of the primary kinetic energy distribution ({sigma}{sub E}) as a function of primary fragment mass (A). The second peak is attributed to a real peak on {sigma}{sub E}(A). However, theoretical calculations related to primary distributions made by H.R. Faust and Z. Bao do not suggest any peak on {sigma}{sub E}(A). In order to clarify this apparent controversy, we have made a numerical experiment in which the masses and the kinetic energy of final fragments are calculated, assuming an initial distribution of the kinetic energy without structures on the standard deviation as function of fragment mass. As a result we obtain a pronounced peak on {sigma}{sub e} (m) curve around m = 109, a depletion from m = 121 to m = 129, and an small peak around m = 122, which is not as great as that measured by Belhafaf et al. Our simulation also reproduces the experimental results on the yield of the final mass Y(m), the average number of emitted neutrons as a function of the provisional mass (calculated from the values of the final kinetic energy of the complementary fragments) and the average value of fragment kinetic energy as a function of the final mass. From our results we conclude that there are no peaks on the {sigma}{sub E} (A) curve, and the observed peaks on {sigma}{sub e} (m) are due to the emitted neutron multiplicity and the variation of the average fragment kinetic energy as a function of primary fragment mass. (Author)

Jet multiplicity distributions: medium dependence in MLLA

Energy Technology Data Exchange (ETDEWEB)

Armesto, Nestor; Pajares, Carlos; Quiroga-Arias, Paloma [Universidade de Santiago de Compostela, Departamento de Fisica de Particulas and IGFAE, Santiago de Compostela (Spain)

2009-07-15

We study the medium dependence of the multiplicity distributions in the modified leading logarithmic approximation. We focus in the enhancement in the number of branchings as the partons travel trough a dense medium created in a heavy-ion collision. We study the effect of a higher number of splittings in some jet observables by introducing the medium as a constant (f{sub med}) in the splitting functions. Having as our ansatz for the quark and gluon jets mean multiplicities left angle n{sub G} right angle =e{sup {gamma}}{sup y} and left angle n{sub Q} right angle =r{sup -1}e{sup {gamma}}{sup y}, we study in an analytic approach the dependence with the medium (f{sub med}) of the anomalous dimension ({gamma}), the multiplicity ratio (r), and so the mean multiplicities. We also obtain the higher-order moments of the multiplicity distribution, what allows us to study its dispersion. (orig.)

New method for exact measurement of thermal neutron distribution in elementary cell

International Nuclear Information System (INIS)

Takac, S.M.; Krcevinac, S.B.

1966-06-01

Exact measurement of thermal neutron density distribution in an elementary cell necessitates the knowledge of the perturbations involved in the cell by the measuring device. A new method has been developed in which a special stress is made to evaluate these perturbations by measuring the response from the perturbations introduced in the elementary cell. The unperturbed distribution was obtained by extrapolation to zero perturbation. The final distributions for different lattice pitches were compared with a THERMOS-type calculation. As a pleasing fact a very good agreement has been reached, which dissolves the long existing disagreement between THERMOS calculations and measured density distribution (author)

New method for exact measurement of thermal neutron distribution in elementary cell

Energy Technology Data Exchange (ETDEWEB)

Takac, S M; Krcevinac, S B [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1966-06-15

Exact measurement of thermal neutron density distribution in an elementary cell necessitates the knowledge of the perturbations involved in the cell by the measuring device. A new method has been developed in which a special stress is made to evaluate these perturbations by measuring the response from the perturbations introduced in the elementary cell. The unperturbed distribution was obtained by extrapolation to zero perturbation. The final distributions for different lattice pitches were compared with a THERMOS-type calculation. As a pleasing fact a very good agreement has been reached, which dissolves the long existing disagreement between THERMOS calculations and measured density distribution (author)

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

International Nuclear Information System (INIS)

Aredes, Vitor Ottoni Garcia

2014-01-01

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

Multiplicity distribution and forward-backward correlations in the hydrodynamic theory of multiple production

International Nuclear Information System (INIS)

Tarasov, Yu.A.

1992-01-01

A hydrodynamic model for the collisions of gluon clusters is used to calculate the charge-particle multiplicity distributions in collisions of nucleons at ISR and collider energies. The separation temperature of the hydrodynamic system is calculated as a function of the rapidity [T k (γ 1 )] for each value of the inelasticity coefficient K. In the central region, this temperature is higher at collider energies than at the ISR energy. The average number of resonance clusters which decay into various (fixed) numbers of charged hadrons is found for each value of K. The number of these clusters fluctuates in accordance with a Poisson distribution. A hadron multiplicity distribution which incorporates these fluctuations is found. This distribution is averaged over the inelasticity coefficient. The distributions p(n ch ) and the KNO functions Ψ(z) are calculated for the overall and central regions of the rapidity, |y| ≤ 1.5. The broadening of the distributions and the violation of KNO scaling at collider energies results from increased contributions from the decays of resonances. The front-back multiplicity correlations are also studied; the decay of resonances is taken into account. The distributions and slope coefficients of the correlation function which are found for the various energies agree with experimental data

Isospin Conservation in Neutron Rich Systems of Heavy Nuclei

Science.gov (United States)

Jain, Ashok Kumar; Garg, Swati

2018-05-01

It is generally believed that isospin would diminish in its importance as we go towards heavy mass region due to isospin mixing caused by the growing Coulomb forces. However, it was realized quite early that isospin could become an important and useful quantum number for all nuclei including heavy nuclei due to neutron richness of the systems [1]. Lane and Soper [2] also showed in a theoretical calculation that isospin indeed remains quite good in heavy mass neutron rich systems. In this paper, we present isospin based calculations [3, 4] for the fission fragment distributions obtained from heavy-ion fusion fission reactions. We discuss in detail the procedure adopted to assign the isospin values and the role of neutron multiplicity data in obtaining the total fission fragment distributions. We show that the observed fragment distributions can be reproduced rather reasonably well by the calculations based on the idea of conservation of isospin. This is a direct experimental evidence of the validity of isospin in heavy nuclei, which arises largely due to the neutron-rich nature of heavy nuclei and their fragments. This result may eventually become useful for the theories of nuclear fission and also in other practical applications.

Dose distributions in thorax inhomogeneity for fast neutron beam from NIRS cyclotron

International Nuclear Information System (INIS)

Kutsutani-Nakamura, Yuzuru; Furukawa, Shigeo; Iinuma, T.A.; Kawashima, Katsuhiro; Hoshino, Kazuo; Hiraoka, Takeshi; Maruyama, Takashi; Sakashita, Kunio; Tsunemoto, Hiroshi

1990-01-01

The power law tissue-air ratio (TAR) method developed by Batho appears to be practical use for inhomogeneity corrections to the dose calculated in a layered media for photon beam therapy. The validity was examined in applying the modified power law TAR and the isodose shift methods to the dose calculation in thorax tissue inhomogeneity containing the boundary region for fast neutron beam. The neutron beam is produced by bombarding a thick beryllium target with 30 MeV deuterons. Lung phantom was made of granulated tissue equivalent plastic, which resulted in density of 0.30 and 0.60 g/cm 3 . Depth dose distributions for neutron beam were measured in thorax phantom by an air-filled cylindrical ionization chamber with TE plastic wall. The power law TAR method considering TAR of zero depth at boundary was compared with the measured data and a good result was obtained that the calculated dose was within ±3 % against the measured. But the isodose shift method is not so good for dose calculation in thorax tissue inhomogeneity using fast neutron beam. (author)

Multiplicity distributions in the dual parton model

International Nuclear Information System (INIS)

Batunin, A.V.; Tolstenkov, A.N.

1985-01-01

Multiplicity distributions are calculated by means of a new mechanism of production of hadrons in a string, which was proposed previously by the authors and takes into account explicitly the valence character of the ends of the string. It is shown that allowance for this greatly improves the description of the low-energy multiplicity distributions. At superhigh energies, the contribution of the ends of the strings becomes negligibly small, but in this case multi-Pomeron contributions must be taken into account

Entrance channel systematics of pre-scission neutron multiplicities

International Nuclear Information System (INIS)

Shareef, M.; Prasad, E.; Chatterjee, A.

2016-01-01

Statistical model analysis has been performed for the available neutron multiplicity (ν_p_r_e) data in the literature. Larger ν_p_r_e values for more symmetric reactions have been observed in comparison with asymmetric reactions forming the same compound nucleus, in most cases. A reverse trend has also been noticed in a few cases. A systematic entrance channel dependence of fission timescale is brought out in this work. Fission timescales calculated using the experimental ν_p_r_e values fall into two distinct groups according to the entrance channel mass asymmetry of the reaction with respect to the Businaro-Gallone critical mass asymmetry. The difference in the delay between these two groups ranges between 20 and 100 zs, which is larger than that reported in some cases. (orig.)

Neutronic design of a 22 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

2008-01-01

The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (authors)

Neutronic design of a 22 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

2006-12-01

The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (author)

A QUANTITATIVE EVALUATION OF THE WATER DISTRIBUTION IN A SOIL SAMPLE USING NEUTRON IMAGING

Directory of Open Access Journals (Sweden)

Jan Šácha

2016-10-01

Full Text Available This paper presents an empirical method by Kang et al. recently proposed for correcting two-dimensional neutron radiography for water quantification in soil. The method was tested on data from neutron imaging of the water infiltration in a soil sample. The raw data were affected by neutron scattering and by beam hardening artefacts. Two strategies for identifying the correction parameters are proposed in this paper. The method has been further developed for the case of three-dimensional neutron tomography. In a related experiment, neutron imaging is used to record ponded-infiltration experiments in two artificial soil samples. Radiograms, i.e., two-dimensional projections of the sample, were acquired during infiltration. A calculation was made of the amount of water and its distribution within the radiograms, in the form of two-dimensional water thickness maps. Tomograms were reconstructed from the corrected and uncorrected water thickness maps to obtain the 3D spatial distribution of the water content within the sample. Without the correction, the beam hardening and the scattering effects overestimated the water content values close to the perimeter of the sample, and at the same time underestimated the values close to the centre of the sample. The total water content of the entire sample was the same in both cases. The empirical correction method presented in this study is a relatively accurate, rapid and simple way to obtain the quantitatively determined water content from two-dimensional and three-dimensional neutron images. However, an independent method for measuring the total water volume in the sample is needed in order to identify the correction parameters.

Comparison of multiplicity distributions to the negative binomial distribution in muon-proton scattering

International Nuclear Information System (INIS)

Arneodo, M.; Ferrero, M.I.; Peroni, C.; Bee, C.P.; Bird, I.; Coughlan, J.; Sloan, T.; Braun, H.; Brueck, H.; Drees, J.; Edwards, A.; Krueger, J.; Montgomery, H.E.; Peschel, H.; Pietrzyk, U.; Poetsch, M.; Schneider, A.; Dreyer, T.; Ernst, T.; Haas, J.; Kabuss, E.M.; Landgraf, U.; Mohr, W.; Rith, K.; Schlagboehmer, A.; Schroeder, T.; Stier, H.E.; Wallucks, W.

1987-01-01

The multiplicity distributions of charged hadrons produced in the deep inelastic muon-proton scattering at 280 GeV are analysed in various rapidity intervals, as a function of the total hadronic centre of mass energy W ranging from 4-20 GeV. Multiplicity distributions for the backward and forward hemispheres are also analysed separately. The data can be well parameterized by binomial distributions, extending their range of applicability to the case of lepton-proton scattering. The energy and the rapidity dependence of the parameters is presented and a smooth transition from the binomial distribution via Poissonian to the ordinary binomial is observed. (orig.)

The number of degrees of freedom for statistical distribution of s wave reduced neutron width for several nuclei

International Nuclear Information System (INIS)

Zhixiang, Z.

1983-01-01

The least squares fit has been performed using chi-squared distribution function for all available evaluated data for s-wave reduced neutron width of several nuclei. The number of degrees of freedom and average value have been obtained. The missing levels of weak s-wave resonances and extra p-wave levels have been taken into account, if any. For 75 As and 103 Rh, s-wave population has been separated by Bayes' theorem before making fit. The results thus obtained are consistent with Porter-Thomas distribution, i.e., chi-squared distribution with γ=1, as one would expect. It has not been found in this work that the number of degrees of freedom for the distribution of s-wave reduced neutron width might be greater than one as reported by H.C.Sharma et al. (1976) at the international conference on interactions of neutrons with nuclei. (Auth.)

Perturbation method for experimental determination of neutron spatial distribution in the reactor cell

International Nuclear Information System (INIS)

Takac, S.M.

1972-01-01

The method is based on perturbation of the reactor cell from a few up to few tens of percent. Measurements were performed for square lattice calls of zero power reactors Anna, NORA and RB, with metal uranium and uranium oxide fuel elements, water, heavy water and graphite moderators. Character and functional dependence of perturbations were obtained from the experimental results. Zero perturbation was determined by extrapolation thus obtaining the real physical neutron flux distribution in the reactor cell. Simple diffusion theory for partial plate cell perturbation was developed for verification of the perturbation method. The results of these calculation proved that introducing the perturbation sample in the fuel results in flattening the thermal neutron density dependent on the amplitude of the applied perturbation. Extrapolation applied for perturbed distributions was found to be justified

Study of the neutron-photon competition during fission fragment de-excitation

International Nuclear Information System (INIS)

Min, Dong Pil.

1976-01-01

A program was developed to study in detail the competition between neutron and photon emissions during the different stages of the nucleus de-excitation. The main conclusions of this work are the following: the neutron-photon competition fairly depends of the initial spin of the primary fragment. It has a strong effect on the mean number of emitted neutrons, on the photon energy, and to a lower degree, on the mean energy per neutron. A relation between the mean initial spin of the heavy fragment for the almost symmetrical fission, the mean initial spin of the heavy fragment for a very asymmetric fission and the corresponding values of the mean number of emitted neutrons is given. The mean initial excitation energy must increase of about 9MeV for the nucleus to emit one more neutron. Two reasons are given to explain the fact that the measured neutron multiplicity variance is higher for the heavy fragment than for the light one: either the existence of a covariance between spin and excitation energy distribution, or a dispersion of the values of the mean number of emitted neutrons due to the mass and charge distribution resulting from experimental incertitudes. The mean energy per neutron calculated with the program is in good agreement with measured values [fr

Calculations of charged-particle recoils, slowing-down spectra, LET and event-size distributions for fast neutrons and comparisons with measurements

International Nuclear Information System (INIS)

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

1979-01-01

A rapid system has been developed for computing charged-particle distributions generated in tissue by any neutron spectra less than 4 MeV. Oxygen and carbon recoils were derived from R-matrix theory, and hydrogen recoils were obtained from cross-section evaluation. Application to two quite different fission-neutron spectra demonstrates the flexibility of this method for providing spectral details of the different types of charged-particle recoils. Comparisons have been made between calculations and measurements of event-size distributions for a sphere of tissue 1 μm in diameter irradiated by these two neutron spectra. LET distributions have been calculated from computed charged-particle recoils and also derived from measurements using the conventional approximation that all charged particles traverse the chamber. The limitations of the approximation for these neutron spectra are discussed. (author)

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

International Nuclear Information System (INIS)

Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

2008-01-01

The Molten Salt Reactor (MSR), one of the 'Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition. (authors)

Radical distributions in ammonium tartrate single crystals exposed to photon and neutron beams

International Nuclear Information System (INIS)

Marrale, M.; Longo, A.; Brai, M.; Barbon, A.; Brustolon, M.

2014-01-01

The radiation therapy carried out by means of heavy charged particles (such as carbon ions) and neutrons is rapidly becoming widespread worldwide. The success of these radiation therapies relies on the high density of energy released by these particles or by secondary particles produced after primary interaction with matter. The biological damages produced by ionising radiations in tissues and cells depend more properly on the energy released per unit pathlength, which is the linear energy transfer and which determines the radiation quality. To improve the therapy effectiveness, it is necessary to grasp the mechanisms of free radical production and distribution after irradiation with these particles when compared with the photon beams. In this work some preliminary results on the analysis of the spatial distributions of the free radicals produced after exposure of ammonium tartrate crystals to various radiation beams ( 60 Co gamma photons and thermal neutrons) were reported. Electron spin resonance analyses were performed by the electron spin echo technique, which allows the determination of local spin concentrations and by double electron-electron resonance technique, which is able to measure the spatial distance distribution (range 1.5-8 nm) among pairs of radicals in solids. The results of these analyses are discussed on the basis of the different distributions of free radicals produced by the two different radiation beams used. This paper extends to the single crystal case, a similar work done on AT powder irradiated with different beams, with assessment of microscopic radical concentration by determining the amount of ID contribution and obtaining the inter-radical distance distributions by double microwave irradiation. In this paper single crystals of AT have been exposed to 60 Co photons and neutrons. The results confirm that advanced pulse EPR techniques allow the direct measurement of the local free radical concentration and provide information about the

Neutron flux distribution measurement in the elementary cell of the RB reactor

Energy Technology Data Exchange (ETDEWEB)

Takac, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1963-04-15

The distribution of thermal neutrons was measured in the elementary cell with dysprosium foils for different lattice pitches and the results obtained were given. From the distributions measured average fluxes were determined for every single zone. By using the published data concerning effective absorption cross sections thermal utilization factors were calculated and their changes given as functions of lattice pitches: l = 8.0 cm; 11.3 cm and 17.9 cm. (author)

Experimental evaluation of the extended Dytlewski-style dead time correction formalism for neutron multiplicity counting

Science.gov (United States)

Lockhart, M.; Henzlova, D.; Croft, S.; Cutler, T.; Favalli, A.; McGahee, Ch.; Parker, R.

2018-01-01

Over the past few decades, neutron multiplicity counting has played an integral role in Special Nuclear Material (SNM) characterization pertaining to nuclear safeguards. Current neutron multiplicity analysis techniques use singles, doubles, and triples count rates because a methodology to extract and dead time correct higher order count rates (i.e. quads and pents) was not fully developed. This limitation is overcome by the recent extension of a popular dead time correction method developed by Dytlewski. This extended dead time correction algorithm, named Dytlewski-Croft-Favalli(DCF), is detailed in reference Croft and Favalli (2017), which gives an extensive explanation of the theory and implications of this new development. Dead time corrected results can then be used to assay SNM by inverting a set of extended point model equations which as well have only recently been formulated. The current paper discusses and presents the experimental evaluation of practical feasibility of the DCF dead time correction algorithm to demonstrate its performance and applicability in nuclear safeguards applications. In order to test the validity and effectiveness of the dead time correction for quads and pents, 252Cf and SNM sources were measured in high efficiency neutron multiplicity counters at the Los Alamos National Laboratory (LANL) and the count rates were extracted up to the fifth order and corrected for dead time. In order to assess the DCF dead time correction, the corrected data is compared to traditional dead time correction treatment within INCC. The DCF dead time correction is found to provide adequate dead time treatment for broad range of count rates available in practical applications.

Vertical distribution of elements in non-polluted estuarine sediments determined by neutron induced prompt gamma-ray and instrumental neutron activation analyses

International Nuclear Information System (INIS)

Kuno, A.; Sampei, K.; Matsuo, M.; Sawahata, H.

1999-01-01

Neutron induced prompt gamma-ray analysis (PGA) and instrumental neutron activation analysis (INAA) have been applied to the sediments collected from the Yasaka River estuary in Oita Prefecture, Japan. The vertical distribution of 33 elements in the sediments has been determined and compared with that in more polluted estuarine sediments. While the S content increased with increasing depth because of a sulphide accumulation under reducing condition, the increase in sulphide-forming elements such as Ag, Cd, Co and Zn was not observed in the deeper section of the Yasaka River estuarine sediments. (author)

Proton multiplicity distributions in high-energy hadron-nuclei collisions

International Nuclear Information System (INIS)

Strugalski, Z.

1979-01-01

The fast proton emission process is analyzed in high-energy hadron-nuclei collisions. The formula describing the proton multiplicity distributions is derived. It describes well enough the proton multiplicity distribution of pion-nuclei and proton-nuclei collisions at 200 and 400 GeV

Hyper-thermal neutron irradiation field for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1994-01-01

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

Timelike Compton scattering off the neutron and generalized parton distributions

Energy Technology Data Exchange (ETDEWEB)

Boer, M.; Guidal, M. [CNRS-IN2P3, Universite Paris-Sud, Institut de Physique Nucleaire d' Orsay, Orsay (France); Vanderhaeghen, M. [Johannes Gutenberg Universitaet, Institut fuer Kernphysik and PRISMA Cluster of Excellence, Mainz (Germany)

2016-02-15

We study the exclusive photoproduction of an electron-positron pair on a neutron target in the Jefferson Lab energy domain. The reaction consists of two processes: the Bethe-Heitler and the Timelike Compton Scattering. The latter process provides potentially access to the Generalized Parton Distributions (GPDs) of the nucleon. We calculate all the unpolarized, single- and double-spin observables of the reaction and study their sensitivities to GPDs. (orig.)

Energy–angle correlation of neutrons and gamma-rays emitted from an HEU source

Energy Technology Data Exchange (ETDEWEB)

Miloshevsky, G., E-mail: gennady@purdue.edu; Hassanein, A.

2014-06-01

Special Nuclear Materials (SNM) yield very unique fission signatures, namely correlated neutrons and gamma-rays. A major challenge is not only to detect, but also to rapidly identify and recognize SNM with certainty. Accounting for particle multiplicity and correlations is one of standard ways to detect SNM. However, many parameter data such as joint distributions of energy, angle, lifetime, and multiplicity of neutrons and gamma-rays can lead to better recognition of SNM signatures in the background radiation noise. These joint distributions are not well understood. The Monte Carlo simulations of the transport of neutrons and gamma-rays produced from spontaneous and interrogation-induced fission of SNM are carried out using the developed MONSOL computer code. The energy spectra of neutrons and gamma-rays from a bare Highly Enriched Uranium (HEU) source are investigated. The energy spectrum of gamma-rays shows spectral lines by which HEU isotopes can be identified, while those of neutrons do not show any characteristic lines. The joint probability density function (JPDF) of the energy–angle association of neutrons and gamma-rays is constructed. Marginal probability density functions (MPDFs) of energy and angle are derived from JPDF. A probabilistic model is developed for the analysis of JPDF and MPDFs. This probabilistic model is used to evaluate mean values, standard deviations, covariance and correlation between the energy and angle of neutrons and gamma-rays emitted from the HEU source. For both neutrons and gamma-rays, it is found that the energy–angle variables are only weakly correlated.

Fast neutron forward distributions from C, Be and U thick targets bombarded by deuterons

International Nuclear Information System (INIS)

Menard, S.; Clapier, F.; Pauwels, N.; Proust, J.; Donzaud, C.; Guillemaud-Mueller, D.; Lhenry, I.; Mueller, A.C.; Scarpaci, J.A.; Sorlin, O.; Mirea, M.

1999-01-01

In principle, to produce neutron rich radioactive beams with sufficient intensities, a source of isotopes far from the valley of β--stability can be obtained through the fission of 238 U induced by fast neutrons. A very promising way to assess the feasibility of these very intense neutron beams is to break an intense 2 H beam in a dedicated converter. The main objective of SPIRAL and PARRNe R - D projects is the investigation of the optimum parameters for a neutron rich isotope source in accordance with the scheme presented above. In such conditions, the charge particle energy loss can prevent the destruction of the fission target. In the frame of these project, a special attention is dedicated to the energetic and angular distributions of the neutrons emerging from a set of converters at a series of 2 H incident energies. Deuteron beams at energies less than 30 MeV are particularly interesting because it is expected that, after the decay in the 238 U target, the neutron rich radioactive fission products are cold enough, thus avoiding the evaporation of a too large number of neutrons. For such purposes, one needs experimental angular distributions at given energies for different types of converters and to elaborate a theoretical tool in order to estimate accurately the characteristics of the secondary neutron beam. In this paper, the experimental results were obtained with 17, 20 and 28 MeV deuteron energies on Be, C and U converters using the time of flight method. These data are compared to results given by a model valid at higher energy in order to obtain pertinent simulations in a large range of incident energies. Many theoretical tools were developed to characterize the properties of the neutron beams emerging from thick targets. In this contribution the Serber's model, considered with its improvements which account for the Coulomb deflection and the mean straggling of the beam in the material, is compared to experimental data in order to verify the validity

Neutron activation for logging the distribution of gold in bore-hole cores

International Nuclear Information System (INIS)

Rahmanian, H.; Watterson, J.I.W.

1992-01-01

A new method for the non-destructive determination of gold in bore-hole cores has been developed using instrumental neutron activation analysis with a 252 Cf source. The procedure obtains the distribution and concentration of gold along the longitudinal axis of the core i.e. a log of the gold concentration. The accuracy of the method is comparable to fire assay at a level of 2 ppm and has a detection limit of 1 ppm under the conditions used. The assay of the gold is carried out by employing a novel variation of the conventional comparator method using gold wires as both standard and flux monitor. A method is described for logging gold in bore-hole cores using neutron activation with a 160 μg 252 Cf neutron source. The method has a limit of detection of about 1 ppm under the described conditions. (author)

Multiplicity distributions in high energy collisions

International Nuclear Information System (INIS)

Giovannini, A.; Lupia, S.; Ugoccioni, R.

1992-01-01

We discuss the important phases in the evolution of our understanding of multiplicity distributions in high energy collisions with particular emphasis to intermittent behavior and shoulder structure problem. (orig.)

Multiplicity distributions in high energy collisions

Energy Technology Data Exchange (ETDEWEB)

Giovannini, A.; Lupia, S.; Ugoccioni, R. (Dipt. di Fisica Teorica, Univ. Turin (Italy) INFN, Turin (Italy))

1992-03-01

We discuss the important phases in the evolution of our understanding of multiplicity distributions in high energy collisions with particular emphasis to intermittent behavior and shoulder structure problem. (orig.).

Assessing cadmium distribution applying neutron radiography in moss trophical levels, located in Szarvasko, Hungary

Energy Technology Data Exchange (ETDEWEB)

Varga, J.; Koroesi, F. E-mail: fkorosi@hotmail.com; Balasko, M. E-mail: balasko@sunserv.kfki.hu; Naar, Z

2004-10-01

The measuring station of the 10 MW VVR-SM research reactor at the Budapest Neutron Centre (Hungary) was used to perform dynamic neutron radiography (DNR), which was, to our best knowledge, the first time, in a Tortella tortuosa biotope. In the conducted study, two trophical levels, moss and spider Thomisidae sp. juv., were examined. Cadmium penetration routes, distribution and accumulation zones were visualized in the leafy gametophyte life cycle of Tortella tortuosa and in the organs of the spider.

The secondary neutrons spectra of 235U, 238U for incident energy range 1-2.5 MeV

International Nuclear Information System (INIS)

Kornilov, N.V.; Kagalenko, A.B.; Balitsky, A.V.; Baryba, V.Ja.; Androsenko, P.A.; Androsenko, A.A.

1993-01-01

Spectra of inelastic scattered neutrons and fission neutrons were measured with neutron time of flight spectrometer. The solid tritium target was used as a neutron source. The energy distribution of neutrons on the sample was calculated with Monte-Carlo code, taking into account interaction income protons inside target and reaction kinematics. The detector efficiency was determined with 252 Cf source. The multiple scattering and absorption corrections were calculated with codes packet BRAND. Our results confirm ENDF/B-6 data library. (author)

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)

Negative binomial distribution for multiplicity distributions in e/sup +/e/sup -/ annihilation

International Nuclear Information System (INIS)

Chew, C.K.; Lim, Y.K.

1986-01-01

The authors show that the negative binomial distribution fits excellently the available charged-particle multiplicity distributions of e/sup +/e/sup -/ annihilation into hadrons at three different energies √s = 14, 22 and 34 GeV

Comparisons of calculated and measured spectral distributions of neutrons from a 14-MeV neutron source inside the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.

1985-12-01

A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions

Enhanced finite difference scheme for the neutron diffusion equation using the importance function

International Nuclear Information System (INIS)

Vagheian, Mehran; Vosoughi, Naser; Gharib, Morteza

2016-01-01

Highlights: • An enhanced finite difference scheme for the neutron diffusion equation is proposed. • A seven-step algorithm is considered based on the importance function. • Mesh points are distributed through entire reactor core with respect to the importance function. • The results all proved that the proposed algorithm is highly efficient. - Abstract: Mesh point positions in Finite Difference Method (FDM) of discretization for the neutron diffusion equation can remarkably affect the averaged neutron fluxes as well as the effective multiplication factor. In this study, by aid of improving the mesh point positions, an enhanced finite difference scheme for the neutron diffusion equation is proposed based on the neutron importance function. In order to determine the neutron importance function, the adjoint (backward) neutron diffusion calculations are performed in the same procedure as for the forward calculations. Considering the neutron importance function, the mesh points can be improved through the entire reactor core. Accordingly, in regions with greater neutron importance, density of mesh elements is higher than that in regions with less importance. The forward calculations are then performed for both of the uniform and improved non-uniform mesh point distributions and the results (the neutron fluxes along with the corresponding eigenvalues) for the two cases are compared with each other. The results are benchmarked against the reference values (with fine meshes) for Kang and Rod Bundle BWR benchmark problems. These benchmark cases revealed that the improved non-uniform mesh point distribution is highly efficient.

The Development of Advanced Processing and Analysis Algorithms for Improved Neutron Multiplicity Measurements

International Nuclear Information System (INIS)

Santi, P.; Favalli, A.; Hauck, D.; Henzl, V.; Henzlova, D.; Ianakiev, K.; Iliev, M.; Swinhoe, M.; Croft, S.; Worrall, L.

2015-01-01

One of the most distinctive and informative signatures of special nuclear materials is the emission of correlated neutrons from either spontaneous or induced fission. Because the emission of correlated neutrons is a unique and unmistakable signature of nuclear materials, the ability to effectively detect, process, and analyze these emissions will continue to play a vital role in the non-proliferation, safeguards, and security missions. While currently deployed neutron measurement techniques based on 3He proportional counter technology, such as neutron coincidence and multiplicity counters currently used by the International Atomic Energy Agency, have proven to be effective over the past several decades for a wide range of measurement needs, a number of technical and practical limitations exist in continuing to apply this technique to future measurement needs. In many cases, those limitations exist within the algorithms that are used to process and analyze the detected signals from these counters that were initially developed approximately 20 years ago based on the technology and computing power that was available at that time. Over the past three years, an effort has been undertaken to address the general shortcomings in these algorithms by developing new algorithms that are based on fundamental physics principles that should lead to the development of more sensitive neutron non-destructive assay instrumentation. Through this effort, a number of advancements have been made in correcting incoming data for electronic dead time, connecting the two main types of analysis techniques used to quantify the data (Shift register analysis and Feynman variance to mean analysis), and in the underlying physical model, known as the point model, that is used to interpret the data in terms of the characteristic properties of the item being measured. The current status of the testing and evaluation of these advancements in correlated neutron analysis techniques will be discussed

Evidence for a maximum mass cut-off in the neutron star mass distribution and constraints on the equation of state

Science.gov (United States)

Alsing, Justin; Silva, Hector O.; Berti, Emanuele

2018-04-01

We infer the mass distribution of neutron stars in binary systems using a flexible Gaussian mixture model and use Bayesian model selection to explore evidence for multi-modality and a sharp cut-off in the mass distribution. We find overwhelming evidence for a bimodal distribution, in agreement with previous literature, and report for the first time positive evidence for a sharp cut-off at a maximum neutron star mass. We measure the maximum mass to be 2.0M⊙ sharp cut-off is interpreted as the maximum stable neutron star mass allowed by the equation of state of dense matter, our measurement puts constraints on the equation of state. For a set of realistic equations of state that support >2M⊙ neutron stars, our inference of mmax is able to distinguish between models at odds ratios of up to 12: 1, whilst under a flexible piecewise polytropic equation of state model our maximum mass measurement improves constraints on the pressure at 3 - 7 × the nuclear saturation density by ˜30 - 50% compared to simply requiring mmax > 2M⊙. We obtain a lower bound on the maximum sound speed attained inside the neutron star of c_s^max > 0.63c (99.8%), ruling out c_s^max c/√{3} at high significance. Our constraints on the maximum neutron star mass strengthen the case for neutron star-neutron star mergers as the primary source of short gamma-ray bursts.

Fission cross section and fission fragment angular distribution for oriented nucleus fission by intermediate energy neutrons (epsilon < or approximately 1 Mev)

International Nuclear Information System (INIS)

Barabanov, A.L.; Grechukhin, D.P.

1985-01-01

General analysis is conducted, and formulae for fission cross section and angular distribution of fission fragments of oriented nuclei by fast neutrons are presented. Geometrical coefficients making up the formulae permitting to carry out calculations for target nuclei with spins I=3/2, 5/2, 7/2 at interaction energies epsilon < or approximately 1 MeV are tabulated. Results of demonstrative calculation of fission fragment angular distribution of oriented sup(235)U nuclei by 0.1 <= epsilon <= 1.0 MeV neutrons reveal that angular distribution weakly depends on the set of permeability factors of neutron waves applied in the calculations

Measurement of the Neutron Slowing-Down Time Distribution at 1.46 eV and its Space Dependence in Water

Energy Technology Data Exchange (ETDEWEB)

Moeller, E

1965-12-15

The use of the time dependent reaction rate method for the measurement of neutron slowing-down time distributions in hydrogen has been analyzed and applied to the case of sloping down in water. Neutrons with energies of about 1 MeV were slowed down, and the time-dependent neutron density at 1.46 eV and its space dependence was measured with a time resolution of 0.042 {mu}s. The results confirm the well known theory for time-dependent slowing down in hydrogen. The space dependence of the distributions is well described by the P{sub 1}-calculations by Claesson.

Measurement of the Neutron Slowing-Down Time Distribution at 1.46 eV and its Space Dependence in Water

International Nuclear Information System (INIS)

Moeller, E.

1965-12-01

The use of the time dependent reaction rate method for the measurement of neutron slowing-down time distributions in hydrogen has been analyzed and applied to the case of sloping down in water. Neutrons with energies of about 1 MeV were slowed down, and the time-dependent neutron density at 1.46 eV and its space dependence was measured with a time resolution of 0.042 μs. The results confirm the well known theory for time-dependent slowing down in hydrogen. The space dependence of the distributions is well described by the P 1 -calculations by Claesson

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

Institute of Scientific and Technical Information of China (English)

ZHANG Da-Lin; QIU Sui-Zheng; LIU Chang-Liang; SU Guang-Hui

2008-01-01

The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.

Monte Carlo calculations of energy and angular distributions of transmitted and backscattered neutrons of 15 MeV incident energy

International Nuclear Information System (INIS)

Gaber, M.; Faied, A.

1994-01-01

The Monte Carlo technique was used to generate both energy and angular distributions of transmitted and backscattered neutrons incident on infinite graphite slabs of thicknesses ranging from 1-90 cm. Point isotropic and parallel beams of 15 MeV neutrons were used. A computer program was developed to simulate collisions by fast neutrons. (author)

Application of the modified neutron source multiplication method for a measurement of sub-criticality in AGN-201K reactor

International Nuclear Information System (INIS)

Myung-Hyun Kim

2010-01-01

Measurement of sub-criticality is a challenging and required task in nuclear industry both for nuclear criticality safety and physics test in nuclear power plant. A relatively new method named as Modified Neutron Source Multiplication Method (MNSM) was proposed in Japan. This method is an improvement of traditional Neutron Source Multiplication (NSM) Method, in which three correction factors are applied additionally. In this study, MNSM was tested in calculation of rod worth using an educational reactor in Kyung Hee University, AGN-201K. For this study, a revised nuclear data library and a neutron transport code system TRANSX-PARTISN were used for the calculation of correction factors for various control rod positions and source locations. Experiments were designed and performed to enhance errors in NSM from the location effects of source and detectors. MNSM can correct these effects but current results showed not much correction effects. (author)

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)

Study of the α and β phases of quartz by neutron multiple diffraction

International Nuclear Information System (INIS)

Mazzocchi, V.L.

1984-01-01

Crystal structures of α and β phases of quartz are studied, employing neutron multiple diffraction as a method of analysis. Theoretical multiple diffraction patterns in a many-beam case were determined by a computer program which calculates intensities of beams as sums of Taylor's series expansions, retaining terms up to a order n. Experimental 'umweg' and transmitted beam patterns were obtained for the 00.1 primary reflection of α and β phases of quartz. To calculate α - quartz multiple diffraction intensities it was necessary to determine the Dauphine twinning fraction for the crystal after having passed by the β-phase. For the two models of β-quartz a better agreement between experimental and calculated integrated intensities was found for the disordered structure model based on split-half-oxigen positions. (Author) [pt

Distributions of neutron and gamma doses in phantom under a mixed field

International Nuclear Information System (INIS)

Beraud-Sudreau, E.

1982-06-01

A calculation program, based on Monte Carlo method, allowed to estimate the absorbed doses relatives to the reactor primary radiation, in a water cubic phantom and in cylindrical phantoms modelized from tissue compositions. This calculation is a theoretical approach of gamma and neutron dose gradient study in an animal phantom. PIN junction dosimetric characteristics have been studied experimentally. Air and water phantom radiation doses measured by PIN junction and lithium 7 fluoride, in reactor field have been compared to doses given by dosimetry classical techniques as tissue equivalent plastic and aluminium ionization chambers. Dosimeter responses have been employed to evaluate neutron and gamma doses in plastinaut (tissue equivalent plastic) and animal (piglet). Dose repartition in the piglet bone medulla has been also determined. This work has been completed by comparisons with Doerschell, Dousset and Brown results and by neutron dose calculations; the dose distribution related to lineic energy transfer in Auxier phantom has been also calculated [fr

Application of probability generating function to the essentials of nondestructive nuclear materials assay system using neutron correlation

International Nuclear Information System (INIS)

Hosoma, Takashi

2017-01-01

In the previous research (JAEA-Research 2015-009), essentials of neutron multiplicity counting mathematics were reconsidered where experiences obtained at the Plutonium Conversion Development Facility were taken into, and formulae of multiplicity distribution were algebraically derived up to septuplet using a probability generating function to make a strategic move in the future. Its principle was reported by K. Böhnel in 1985, but such a high-order expansion was the first case due to its increasing complexity. In this research, characteristics of the high-order correlation were investigated. It was found that higher-order correlation increases rapidly in response to the increase of leakage multiplication, crosses and leaves lower-order correlations behind, when leakage multiplication is > 1.3 that depends on detector efficiency and counter setting. In addition, fission rates and doubles count rates by fast neutron and by thermal neutron in their coexisting system were algebraically derived using a probability generating function again. Its principle was reported by I. Pázsit and L. Pál in 2012, but such a physical interpretation, i.e. associating their stochastic variables with fission rate, doubles count rate and leakage multiplication, is the first case. From Rossi-alpha combined distribution and measured ratio of each area obtained by Differential Die-Away Self-Interrogation (DDSI) and conventional assay data, it is possible to estimate: the number of induced fissions per unit time by fast neutron and by thermal neutron; the number of induced fissions (< 1) by one source neutron; and individual doubles count rates. During the research, a hypothesis introduced in their report was proved to be true. Provisional calculations were done for UO_2 of 1∼10 kgU containing ∼ 0.009 wt% "2"4"4Cm. (author)

Measurement of neutron flux distribution by semiconductor detector

International Nuclear Information System (INIS)

Obradovic, D.; Bosevski, T.

1964-01-01

Application of semiconductor detectors for measuring neutron flux distribution is about 10 times faster than measurements by activation foils and demands significantly lower reactor power. Following corrections are avoided: mass of activation foils which influences the self shielding, nuclear decay during activity measurements; counter dead-time. It is possible to control the measured data during experiment and repeat measurements if needed. Precision of the measurement is higher since it is possible to choose the wanted statistics. The method described in this paper is applied for measurements at the RB reactor. It is concluded that the method is suitable for fast measurements but the activation analysis is still indispensable

Multiple scattering effects in fast neutron polarization experiments using high-pressure helium-xenon gas scintillators as analyzers

International Nuclear Information System (INIS)

Tornow, W.; Mertens, G.

1977-01-01

In order to study multiple scattering effects both in the gas and particularly in the solid materials of high-pressure gas scintillators, two asymmetry experiments have been performed by scattering of 15.6 MeV polarized neutrons from helium contained in stainless steel vessels of different wall thicknesses. A monte Carlo computer code taking into account the polarization dependence of the differential scattering cross sections has been written to simulate the experiments and to calculate corrections for multiple scattering on helium, xenon and the gas containment materials. Besides the asymmetries for the various scattering processes involved, the code yields time-of-flight spectra of the scattered neutrons and pulse height spectra of the helium recoil nuclei in the gas scintillator. The agreement between experimental results and Monte Carlo calculations is satisfactory. (Auth.)

Quantitative detection of microscopic lithium distributions with neutrons

Energy Technology Data Exchange (ETDEWEB)

Neri, Giulia; Gernhaeuser, Roman; Lichtinger, Josef; Winkler, Sonja; Seiler, Dominik; Bendel, Michael [Technische Universitaet Muenchen, Physik-Department (Germany); Kunze-Liebhaeuser, Julia; Brumbarov, Jassen; Portenkirchner, Engelbert [Institut fuer Physikalische Chemie, Leopold-Franzens-Universitaet Innsbruck (Austria); Renno, Axel; Rugel, Georg [Helmholtz Zentrum Dresden Rossendorf, Helmholtz-Institut Freiberg fuer Ressourcentechnologie (Germany)

2016-07-01

The importance of lithium in the modern industrial society is continuously increasing. Spatially resolved detection of tritium particles from {sup 6}Li(n,α){sup 3}H nuclear reactions is used to reconstruct microscopic lithium distributions. Samples are exposed to a flux of cold neutrons. Emitted charged particles are detected with a PSD. Introducing a pinhole aperture between target and detector, the experimental setup works like a ''camera obscura'', allowing to perform spatially resolved measurements. Tritium detection analysis was successfully used to reconstruct the lithium content in self-organized TiO{sub 2-x}-C and Si/TiO{sub 2-x}-C nanotubes electrochemically lithiated, for the first time. Titanium dioxide nanotubes are a candidate for a safe anode material in lithium-ion batteries. Also lithium distributions in geological samples, so called ''pathfinder-minerals'' containing lithium, like lepidolite from a pegmatite, were analyzed. With this development we present a new precision method using nuclear physics for material science.

Fast neutron distributions from Be and C thick targets bombarded with 80 and 160 MeV deuterons

International Nuclear Information System (INIS)

Pauwels, N.; Laurent, H.; Clapier, F.; Brandenburg, S.; Beijers, J. P .M.; Zegers, R. G. T.; Lebreton, H.; Saint-Laurent, M.G.; Mirea, M.

2001-01-01

Production of fast neutron studies have come to the fore in the past few years because of the great interest for the possible applications of induced fission to produce neutron rich ion beams. In this context, the main objective of the SPIRAL II (Systeme de Production d'Ions Radioactifs Acceleres en Ligne) and PARRNe (Production d'Atomes Radioactifs Riches en Neutrons) R and D projects is the investigation of the feasibility and of the optimum parameters for a neutron rich isotope source. Special attention is dedicated to the energy and angular distributions of the neutrons obtained through deuteron break--up in different types of converters and different incident energies. Analysis and modelling of such behaviors, together with the study of the yields of neutron induced fission, can be used to optimize the productivity of the fissioning target its geometry and designing it accordingly. The present report continues our previous studies realised for 17, 20, 28 and 200 MeV deuteron energies and it is focused on deuteron incident energies of 80 and 160 MeV. In the experiment, the double differential cross section for neutron production induced by 80 and 160 MeV deuterons impinging on thick C and Be targets, in which the incident deuterons were complete stopped, have been measured. The energy of the neutrons was determined from the time--of--flight (TOF) measurement. To obtain an energy resolution of about 4% for the fastest, forward--emitted neutrons, which have approximately beam velocity, the length of the flightpath for the detectors at angles up to 30 angle was chosen to be 6 m. At backward angles, where the neutron energies are lower, a shorter flightpath was chosen. A schematic drawing of the setup is shown. A 100 mm thick Be target and a 70 mm thick C target were used. Results are exemplified with the angular and energy distributions of neutron obtained for Be target at 80 MeV. (authors)

Study on uranium-water multiplicative means of the (RESUCO-Subcritical experimental reactor of uranium with oxygen) subcritical assembly by pulsed neutron technique

International Nuclear Information System (INIS)

Jesus Barbosa, S. de.

1987-01-01

The effective multiplication factor and the nuclear parameters associated with the variation of (RESUCO- Subcritical Experimental Reactor of Uranium with Oxygen) Subcritical Assembly Configuration, using pulsed neutron technique are analysed. BF3 detectors were used to detect the variation of thermal neutrons in the system, positioned parallelly to fuel elements, and a proton recoil detector was used for monitoring the neutron generation. (M.C.K.) [pt

Prompt fission neutron spectra and anti νp

International Nuclear Information System (INIS)

Madland, D.G.

Methods used to obtain the evaluated prompt fisson neutron spectrum N(E) and the average prompt neutron multiplicity anti ν/sub p/ are reviewed. The relative influence of experimental data; interpolated, extrapolated, and fitted experimental data; systematics; and nuclear theory are considered for the cases where (a) abundant experimental data exist, (b) some experimental data exist, and (c) no experimental data exist. The Maxwellian and Watt distributions, and the determination of the parameters of these distributions by data fitting, are described and compared to recent new theoretical work on the calculation of N(E). Similarly, various expressions for anti γ/sub p/ that have been obtained by data fitting and systematics are described and compared to recent new theoretical work. Complications in the evaluation of N(E) and anti γ/sub p/ due to the onset of multiple-chance fission and the interrelationships between N(E), anti γΔ/sub p/ and the multiple-chance fission cross section are discussed using the example of the fission of 235 U. Some statistics and comments are given on the evaluations of N(E) and anti γ/sub p/ contained in ENDF/B-V, and a number of concluding recommendations are made for future evaluation work

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

The interpretation of neutron noise in boiling water reactors

International Nuclear Information System (INIS)

John, T.M.; Singh, O.P.

1985-01-01

Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

Use of indicating material for direct indication of neutrons distributed photographically

International Nuclear Information System (INIS)

Weiss, A.K.; Fico, S.S.

1977-01-01

Electrons distributed photographically can be clearly indicated by a layer, which contains a crystalline polyacetylene connection. An indicating material which contained a short chain alkylester of diindiacid, a polyacetylene polyacidic alkylamide, a polyacetylene amine salt or a diindiolurethane proved particularly suitable. The materials are used in neutron radiography. The construction of the layers is described in detail. 11 extensive examples explain possible variations. (UWI) [de

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Quantification of transuranic elements by time interval correlation spectroscopy of the detected neutrons

Science.gov (United States)

Baeten; Bruggeman; Paepen; Carchon

2000-03-01

The non-destructive quantification of transuranic elements in nuclear waste management or in safeguards verifications is commonly performed by passive neutron assay techniques. To minimise the number of unknown sample-dependent parameters, Neutron Multiplicity Counting (NMC) is applied. We developed a new NMC-technique, called Time Interval Correlation Spectroscopy (TICS), which is based on the measurement of Rossi-alpha time interval distributions. Compared to other NMC-techniques, TICS offers several advantages.

A study of the 208Pb + 197Au reaction at 29 MeV/u through the associated neutron multiplicity

International Nuclear Information System (INIS)

Bresson, S.

1993-01-01

The investigation of this heavy symmetric system has been carried out through the study of the associated neutron multiplicity. The experimental techniques and data processing are first described, with emphasis on the Orion neutron detector and the hodoscope used to detect the charged reaction products at forward angles. It is shown that the neutron multiplicity is a good measure of the violence of the collision and a good way to characterize the different modes of the reaction. The fission of the quasi-projectile is then characterized and is shown to occur for peripheral collisions. Using simulations, the minimal values of the angular momentum transferred to the quasi-projectile are determined. The results of dynamical calculations using the Landau Vlasov equation are described, which show the importance of angular momentum. It is demonstrated that, at 29 MeV/u, the Pb + Au collision is still governed by deep inelastic reactions in which angular momentum in the exit channel plays an important role

New experimental research stand SVICKA neutron field analysis using neutron activation detector technique

Science.gov (United States)

Varmuza, Jan; Katovsky, Karel; Zeman, Miroslav; Stastny, Ondrej; Haysak, Ivan; Holomb, Robert

2018-04-01

Knowledge of neutron energy spectra is very important because neutrons with various energies have a different material impact or a biological tissue impact. This paper presents basic results of the neutron flux distribution inside the new experimental research stand SVICKA which is located at Brno University of Technology in Brno, Czech Republic. The experiment also focused on the investigation of the sandwich biological shielding quality that protects staff against radiation effects. The set of indium activation detectors was used to the investigation of neutron flux distribution. The results of the measurement provide basic information about the neutron flux distribution inside all irradiation channels and no damage or cracks are present in the experimental research stand biological shielding.

Combined local current distribution measurements and high resolution neutron radiography of operating direct methanol fuel cells

Energy Technology Data Exchange (ETDEWEB)

Schroeder, Alexander; Wippermann, Klaus [Forschungszentrum Juelich GmbH (Germany). Inst. of Energy Research, IEF-3: Fuel Cells; Sanders, Tilman [RWTH Aachen (DE). Inst. for Power Electronics and Electrical Drives (ISEA); Arlt, Tobias [Helmholtz Centre Berlin (Germany). Inst. for Applied Materials

2010-07-01

Neutron radiography allows the investigation of the local fluid distribution in direct methanol fuel cells (DMFCs) under operating conditions. Spatial resolutions in the order of some tens of micrometers at the full test cell area are achieved. This offers the possibility to study practice-oriented, large stack cells with an active area of several hundred cm{sup 2} as well as specially designed, small test cells with an area of some cm{sup 2}. Combined studies of high resolution neutron radiography and segmented cell measurements are especially valuable, because they enable a correlation of local fluid distribution and local performance [1, 2]. The knowledge of this interdependency is essential to optimise the water management and performance respecting a homogeneous fluid, current and temperature distribution and to achieve high performance and durability of DMFCs. (orig.)

Multiplicity distributions in QCD cascades

International Nuclear Information System (INIS)

Gustafson, G.

1992-03-01

Multiplicity distributions for hadrons and for jets are studied in QCD parton cascades. The colour dipole formalism is used and earlier results in the double log approximation are generalized to include terms which are suppressed by colour factors or factors of ln s. The result is a set of coupled differential equations, together with appropriate boundary conditions

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

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2001-01-01

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

Polycapillary neutron lenses

International Nuclear Information System (INIS)

Mildner, D.F.R.

1997-01-01

The principle of multiple mirror reflection from smooth surfaces at small grazing angles enables the transport and guiding of high intensity slow neutron beams to locations of low background for neutron scattering and absorption experiments and to provide facilities for multiple instruments. Curved guides have been widely used at cold neutron facilities to remove the unwanted radiation (fast neutrons and gamma rays) from the beam without the use of filters. A typical guide has transverse dimensions of 50 mm and, with a radius of curvature of 1 km, transmits wavelengths longer than 5 A. Much tighter curves requires narrower transverse dimensions, otherwise there is little transmission. Typical neutron benders have a number of slots with transverse dimensions of ∼5 mm. Based on the same principle but using a different technology, recent developments in glass polycapillary fibers have produced miniature versions of neutron guides. Fibers with many thousands of channels having sizes of ∼ 10 μm enable beams of long wavelength neutrons (λ > 4 A) to be transmitted efficiently in a radius of curvature as small as a fraction of 1 m. A large collection of these miniature versions of neutron guides can be used to bend the neutron trajectories such that the incident beam can be focused. (author)

Resonant neutron-induced atomic displacements

Energy Technology Data Exchange (ETDEWEB)

Elmaghraby, Elsayed K., E-mail: e.m.k.elmaghraby@gmail.com

2017-05-01

Highlights: • Neutron induced atomic displacements was investigated based on scattering of energy of neutron. • Model for cascade function (multiplication of displacements with increasing energy transfer) was proposed and justified. • Parameterizations for the dpa induced in all elements were performed. • Table containing all necessary parameters to calculate the displacement density induced by neutron is given. • Contribution of non resonance displacement and resonant-neutron induced displacements are distinguished. - Abstract: A model for displacement cascade function was modified to account for the continuous variation of displacement density in the material in response to neutron exposure. The model is based on the Gaussian distribution of displacement energies of atoms in a material. Analytical treatment for moderated epithermal neutron field was given in which the displacement density was divided into two terms, discrete-resonance term and continuum term. Calculation are done for all isotopes using ENDF/B VII.1 data files and temperature dependent cross section library. Weighted elemental values were reported a fitting was performed to obtain energy-dependent formula of displacement density and reduce the number of parameters. Results relevant the present specification of the cascade function are tabulated for each element to enable calculation of displacement density at any value of displacement energy in the between 5 eV and 55 eV.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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