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Sample records for delayed-neutron energy spectra

  1. Statistical theory for calculating energy spectra of β-delayed neutrons

    International Nuclear Information System (INIS)

    Kawano, Toshihiko; Moeller, Peter; Wilson, William B.

    2008-01-01

    Theoretical β-delayed neutron spectra are calculated based on the Quasi-particle Random Phase Approximation (QRPA) and the Hauser-Feshbach statistical model. Neutron emissions from an excited daughter nucleus after β-decay to the granddaughter residual are more accurately calculated than previous evaluations, including all the microscopic nuclear structure information, such as a Gamow-Teller strength distribution and discrete states in the granddaughter. The calculated delayed-neutron spectra reasonably agree with those evaluations in the ENDF decay library, which are based on experimental data. The model was adopted to generate the delayed-neutron spectra for all 271 precursors. (authors)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Analytical applications for delayed neutrons

    International Nuclear Information System (INIS)

    Eccleston, G.W.

    1983-01-01

    Analytical formulations that describe the time dependence of neutron populations in nuclear materials contain delayed-neutron dependent terms. These terms are important because the delayed neutrons, even though their yields in fission are small, permit control of the fission chain reaction process. Analytical applications that use delayed neutrons range from simple problems that can be solved with the point reactor kinetics equations to complex problems that can only be solved with large codes that couple fluid calculations with the neutron dynamics. Reactor safety codes, such as SIMMER, model transients of the entire reactor core using coupled space-time neutronics and comprehensive thermal-fluid dynamics. Nondestructive delayed-neutron assay instruments are designed and modeled using a three-dimensional continuous-energy Monte Carlo code. Calculations on high-burnup spent fuels and other materials that contain a mix of uranium and plutonium isotopes require accurate and complete information on the delayed-neutron periods, yields, and energy spectra. A continuing need exists for delayed-neutron parameters for all the fissioning isotopes

  4. Improved Delayed-Neutron Spectroscopy Using Trapped Ions

    Energy Technology Data Exchange (ETDEWEB)

    Norman, Eric

    2018-04-24

    The neutrons emitted following the  decay of fission fragments (known as delayed neutrons because they are emitted after fission on a timescale of the -decay half-lives) play a crucial role in reactor performance and control. Reviews of delayed-neutron properties highlight the need for high-quality data for a wide variety of delayed-neutron emitters to better understand the timedependence and energy spectrum of the neutrons as these properties are essential for a detailed understanding of reactor kinetics needed for reactor safety and to understand the behavior of these reactors under various accident and component-failure scenarios. For fast breeder reactors, criticality calculations require accurate delayed-neutron energy spectra and approximations that are acceptable for light-water reactors such as assuming the delayed-neutron and fission-neutron energy spectra are identical are not acceptable and improved -delayed neutron data is needed for safety and accident analyses for these reactors. With improved nuclear data, the delayedneutrons flux and energy spectrum could be calculated from the contributions from individual isotopes and therefore could be accurately modeled for any fuel-cycle concept, actinide mix, or irradiation history. High-quality -delayed neutron measurements are also critical to constrain modern nuclear-structure calculations and empirical models that predict the decay properties for nuclei for which no data exists and improve the accuracy and flexibility of the existing empirical descriptions of delayed neutrons from fission such as the six-group representation

  5. Delayed neutron spectra and their uncertainties in fission product summation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

    1997-03-01

    Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

  6. Delayed neutron spectra from short pulse fission of uranium-235

    International Nuclear Information System (INIS)

    Atwater, H.F.; Goulding, C.A.; Moss, C.E.; Pederson, R.A.; Robba, A.A.; Wimett, T.F.; Reeder, P.; Warner, R.

    1986-01-01

    Delayed neutron spectra from individual short pulse (∼50 μs) fission of small 235 U samples (50 mg) were measured using a small (5 cm OD x 5 cm length) NE 213 neutron spectrometer. The irradiating fast neutron flux (∼10 13 neutrons/cm 2 ) for these measurements was provided by the Godiva fast burst reactor at the Los Alamos Critical Experiment Facility (LACEF). A high speed pneumatic transfer system was used to transfer the 50 mg 235 U samples from the irradiation position near the Godiva assembly to a remote shielded counting room containing the NE 213 spectrometer and associated electronics. Data were acquired in sixty-four 0.5 s time bins and over an energy range 1 to 7 MeV. Comparisons between these measurements and a detailed model calculation performed at Los Alamos is presented

  7. Bench mark spectra for high-energy neutron dosimetry

    International Nuclear Information System (INIS)

    Dierckx, R.

    1986-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-29

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

  9. SOURCES-3A: A code for calculating (α, n), spontaneous fission, and delayed neutron sources and spectra

    International Nuclear Information System (INIS)

    Perry, R.T.; Wilson, W.B.; Charlton, W.S.

    1998-04-01

    In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay α-particles in (α,n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO 2 , ThO 2 , MOX, etc.), enrichment plant operations (UF 6 , PuF 4 , etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from (α,n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of α-emitting source material and low-Z target material) and in interface problems (i.e., a slab of α-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to (α,n) reactions induced by a monoenergetic beam of α-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The (α,n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay α-particle spectra, 24 sets of measured and/or evaluated (α,n) cross sections and product nuclide level branching fractions, and functional α-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an

  10. D-D neutron energy-spectra measurements in Alcator C

    International Nuclear Information System (INIS)

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

    1982-08-01

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

  11. Energy spectra of neutrons accompanying the emission fission of 238U

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. Delayed neutrons in liquid metal spallation targets

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Giacri-Mauborgne, M.L

    2005-11-15

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

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  15. Measurement of time-dependent fast neutron energy spectra in a depleted uranium assembly

    International Nuclear Information System (INIS)

    Whittlestone, S.

    1980-10-01

    Time-dependent neutron energy spectra in the range 0.6 to 6.4 MeV have been measured in a depleted uranium assembly. By selecting windows in the time range 0.9 to 82 ns after the beam pulse, it was possible to observe the change of the neutron energy distributions from spectra of predominantly 4 to 6 MeV neutrons to spectra composed almost entirely of fission neutrons. The measured spectra were compared to a Monte Carlo calculation of the experiment using the ENDF/B-IV data file. At times and energies at which the calculation predicted a fission spectrum, the experiment agreed with the calculation, confirming the accuracy of the neutron spectroscopy system. However, the presence of discrepancies at other times and energies suggested that there are significant inconsistencies in the inelastic cross sections in the 1 to 6 MeV range. The time response generated concurrently with the energy spectra was compared to the Monte Carlo calculation. From this comparison, and from examination of time spectra measured by other workers using 235 U and 237 Np fission detectors, it would appear that there are discrepancies in the ENDF/B-IV cross sections below 1 MeV. The predicted decay rates were too low below and too high above 0.8 MeV

  16. Energy spectra of primary knock-on atoms under neutron irradiation

    International Nuclear Information System (INIS)

    Gilbert, M.R.; Marian, J.; Sublet, J.-Ch.

    2015-01-01

    Materials subjected to neutron irradiation will suffer from a build-up of damage caused by the displacement cascades initiated by nuclear reactions. Previously, the main “measure” of this damage accumulation has been through the displacements per atom (dpa) index, which has known limitations. This paper describes a rigorous methodology to calculate the primary atomic recoil events (often called the primary knock-on atoms or PKAs) that lead to cascade damage events as a function of energy and recoiling species. A new processing code SPECTRA-PKA combines a neutron irradiation spectrum with nuclear recoil data obtained from the latest nuclear data libraries to produce PKA spectra for any material composition. Via examples of fusion relevant materials, it is shown that these PKA spectra can be complex, involving many different recoiling species, potentially differing in both proton and neutron number from the original target nuclei, including high energy recoils of light emitted particles such as α-particles and protons. The variations in PKA spectra as a function of time, neutron field, and material are explored. The application of PKA spectra to the quantification of radiation damage is exemplified using two approaches: the binary collision approximation and stochastic cluster dynamics, and the results from these different models are discussed and compared. - Highlights: • Recoil cross-section matrices under neutron irradiation are generated. • Primary knock-on atoms (PKA) spectra are calculated for fusion relevant materials. • Variation in PKA spectra due to changes in geometry are considered. • Inventory simulations to consider time-evolution in PKA spectra. • Damage quantification using damage functions from different approximations.

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

    International Nuclear Information System (INIS)

    Jacobs, G.J.H.

    1982-01-01

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

  18. MONSTER: a TOF Spectrometer for beta-delayed Neutron Spetroscopy

    CERN Document Server

    Martinez, T; Castilla, J; Garcia, A R; Marin, J; Martinez, G; Mendoza, E; Santos, C; Tera, F; Jordan, M D; Rubio, B; Tain, J L; Bhattacharya, C; Banerjee, K; Bhattacharya, S; Roy, P; Meena, J K; Kundu, S; Mukherjee, G; Ghosh, T K; Rana, T K; Pandey, R; Saxena, A; Behera, B; Penttila, H; Jokinen, A; Rinta-Antila, S; Guerrero, C; Ovejero, M C; Villamarin, D; Agramunt, J; Algora, A

    2014-01-01

    Beta-delayed neutron (DN) data, including emission probabilities, P-n, and energy spectrum, play an important role in our understanding of nuclear structure, nuclear astrophysics and nuclear technologies. A MOdular Neutron time-of-flight SpectromeTER (MONSTER) is being built for the measurement of the neutron energy spectra and branching ratios. The TOF spectrometer will consist of one hundred liquid scintillator cells covering a significant solid angle. The MONSTER design has been optimized by using Monte Carlo (MC) techniques. The response function of the MONSTER cell has been characterized with mono-energetic neutron beams and compared to dedicated MC simulations.

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

    International Nuclear Information System (INIS)

    Kroc, T.K.

    2009-01-01

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

  20. Neutron spectra produced by moderating an isotopic neutron source

    International Nuclear Information System (INIS)

    Carrillo Nunnez, Aureliano; Vega Carrillo, Hector Rene

    2001-01-01

    A Monte Carlo study has been carried out to determine the neutron spectra produced by an isotopic neutron source inserted in moderating media. Most devices used for radiation protection have a response strongly dependent on neutron energy. ISO recommends several neutron sources and monoenergetic neutron radiations, but actual working situations have broad spectral neutron distributions extending from thermal to MeV energies, for instance, near nuclear power plants, medical applications accelerators and cosmic neutrons. To improve the evaluation of the dosimetric quantities, is recommended to calibrate the radiation protection devices in neutron spectra which are nearly like those met in practice. In order to complete the range of neutron calibrating sources, it seems useful to develop several wide spectral distributions representative of typical spectra down to thermal energies. The aim of this investigation was to use an isotopic neutron source in different moderating media to reproduce some of the neutron fields found in practice. MCNP code has been used during calculations, in these a 239PuBe neutron source was inserted in H2O, D2O and polyethylene moderators. Moderators were modeled as spheres and cylinders of different sizes. In the case of cylindrical geometry the anisotropy of resulting neutron spectra was calculated from 0 to 2 . From neutron spectra dosimetric features were calculated. MCNP calculations were validated by measuring the neutron spectra of a 239PuBe neutron source inserted in a H2O cylindrical moderator. The measurements were carried out with a multisphere neutron spectrometer with a 6LiI(Eu) scintillator. From the measurements the neutron spectrum was unfolded using the BUNKIUT code and the UTA4 response matrix. Some of the moderators with the source produce a neutron spectrum close to spectra found in actual applications, then can be used during the calibration of radiation protection devices

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

    Directory of Open Access Journals (Sweden)

    Valencia E.

    2014-03-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    CERN Document Server

    Ogawa, Y; Sagawa, H; Tsujimoto, T

    2002-01-01

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

  4. The implication of sensitivity analysis on the safety and delayed-neutron parameters for fast breeder reactors

    International Nuclear Information System (INIS)

    Onega, R.J.; Florian, R.J.

    1983-01-01

    The delayed-neutron energy spectra for LMFBRs are not as well known as those for LWRs. These spectra are necessary for kinetics calculations which play an important role in safety and accident analyses. A sensitivity analysis was performed to study the response of the reactor power and power density to uncertainties in the delayed-neutron spectra during a rod-ejection accident. The accidents studied were central control-rod-ejections with ejection times of 2,10 and 30s. A two-energy group and two-precursor group model was formulated for the International Nuclear Fuel Cycle Evaluation (INFCE) reference design MOX-fueled LMFBR. The sensitivity analysis is based on the use of adjoints so that it is not necessary to repeatedly solve the governing (kinetics) equations to obtain the sensitivity derivatives. This is of particular importance when large systems of equations are used. The power and power-density responses were found to be most sensitive to uncertainties in the spectrum of the second delayed-neutron precursor group, resulting from the fission of 238 U, producing neutrons in the first energy group. It was found, for example, that for a rod-ejection time of 30s, and uncertainty of 7.2% in the fast components of the spectra resulted in a 24% uncertainty in the predicted power and power density. These responses were recalculated by repeatedly solving the kinetics equations. The maximum discrepancy between the recalculated and the sensitivity analysis response was only 1.6%. The results of the sensitivity analysis indicate the need for improved delayed-neutron spectral data in order to reduce the uncertainties in accident analyses. (author)

  5. Determination of fast neutrons energy spectra by Monte-Carlo Method

    International Nuclear Information System (INIS)

    Chetaine, A.

    1986-01-01

    Two computation codes based on the Monte-Carlo method are established for studying the spectrometry of neutrons with 14 Mev as initial energy. The spectra are determined, on one hand, around a neutron generator Ti-T target and, on the other hand, in a big paraffin cylinder. One code allows to determine the spectrum of neutrons irradiating the sample at various distances from the Ti-T target versus accelerator parameters: high voltage, atomic or molecular nature of deuterons beam, target thickness and materials surrounding the target. The other code determines neutron spectra at various positions inside and outside the 30 x 30 cm paraffin cylinder. The validity of the procedure used in these codes is verified by determining the spectrum of neutrons crossing a big surface, using the procedure in question and using direct simulation method. The biasing procedure used in the two codes permits to have results with good statistics from a reduced number of drawings. 70 figs.; 62 refs.; 1 tab. (author)

  6. Neutron energy spectra from the thick target 9Be(d,n)10B reaction

    International Nuclear Information System (INIS)

    Whittlestone, S.

    1976-12-01

    The energy spectrum of neutrons emitted when deuterons impinge on a thick beryllium target has been measured using an NE213 scintillation detector and the time-of-flight technique. Spectra were measured at angles of 0, 30, 45, 60, 90, 120 and 150 0 for deuteron energies of 1.4, 1.8, 2.3 and 2.8 MeV. Tables are presented of these angle-dependent energy spectra, the angle-integrated energy dependent yeidls, and the total neutron yield as a function of deuteron energy. (author)

  7. Systematic of delayed neutron parameters

    International Nuclear Information System (INIS)

    Isaev, S.G.; Piksaikin, V.M.

    2000-01-01

    The experimental studies of the energy dependence of the delayed neutron (DN) parameters for various fission systems has shown that the behaviour of a some combination of delayed neutron parameters has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. It was presented the preliminary results which were obtained during study the physics interpretation of the results [ru

  8. Neutron and photon spectra in LINACs

    International Nuclear Information System (INIS)

    Vega-Carrillo, H.R.; Martínez-Ovalle, S.A.; Lallena, A.M.; Mercado, G.A.; Benites-Rengifo, J.L.

    2012-01-01

    A Monte Carlo calculation, using the MCNPX code, was carried out in order to estimate the photon and neutron spectra in two locations of two linacs operating at 15 and 18 MV. Detailed models of both linac heads were used in the calculations. Spectra were estimated below the flattening filter and at the isocenter. Neutron spectra show two components due to evaporation and knock-on neutrons. Lethargy spectra under the filter were compared to the spectra calculated from the function quoted by Tosi et al. that describes reasonably well neutron spectra beyond 1 MeV, though tends to underestimate the energy region between 10 –6 and 1 MeV. Neutron and the Bremsstrahlung spectra show the same features regardless of the linac voltage. - Highlights: ► With MCNPX code realistic models of two LINACs were built. ► Photon and neutron spectra below the flattening filter and at the isocenter were calculated. ► Neutron spectrum at the flattening filter was compared against the Tosi et al. source-term model. ► Tosi et al. model underestimates the neutron contribution below 1 MeV. ► Photon spectra look alike to those published in literature.

  9. Beta-delayed neutron decay of $^{33}$Na

    CERN Document Server

    Radivojevic, Z; Caurier, E; Cederkäll, J; Courtin, S; Dessagne, P; Jokinen, A; Knipper, A; Le Scornet, G; Lyapin, V G; Miehé, C; Nowacki, F; Nummela, S; Oinonen, M; Poirier, E; Ramdhane, M; Trzaska, W H; Walter, G; Äystö, J

    2002-01-01

    Beta-delayed neutron decay of /sup 33/Na has been studied using the on-line mass separator ISOLDE. The delayed neutron spectra were measured by time-of-flight technique using fast scintillators. Two main neutron groups at 800(60) and 1020(80) keV were assigned to the /sup 33/Na decay, showing evidence for strong feeding of states at about 4 MeV in /sup 33/Mg. By simultaneous beta - gamma -n counting the delayed neutron emission probabilities P/sub 1n/ = 47(6)% and P /sub 2n/ = 13(3)% were determined. The half-life value for /sup 33 /Na, T/sub 1/2/ = 8.0(3) ms, was measured by three different techniques, one employing identifying gamma transitions and two employing beta and neutron counting. (21 refs).

  10. Monte Carlo calculations of neutron and gamm-ray energy spectra for fusion-reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.

    1983-08-01

    Neutron and gamma-ray spectra resulting from the interactions of approx. 14-MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree within 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra is also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE

  11. Study of $\\beta$-delayed neutron decay of $^{8}$He

    CERN Multimedia

    The goal of the present proposal is to study $\\beta$-delayed neutron decay branch of $^{8}$He. The energy spectra of the emitted neutrons will be measured in the energy range of 0.1 – 6 MeV using the VANDLE spectrometer. Using coincident $\\gamma$-ray measurement, components of the spectrum corresponding to transitions to the ground- and first- excited states of $^{7}$Li will be disentangled. The new data will allow us to get a more complete picture of the $\\beta$-decay of $^{8}$He and to clarify the discrepancy between the B(GT) distributions derived from the $\\beta$-decay and $^{8}$He(p, n)$^{8}$Li reaction studies.

  12. Energy dependence of relative abundances and periods of delayed neutron separate groups from neutron induced fission of 239Pu in the virgin neutron energy range 0.37-4.97 MeV

    International Nuclear Information System (INIS)

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

    2002-01-01

    Relative yield and group period of delayed neutrons induced by the 239 Pu fission in the 0.37-4.97 MeV range were measured. Comparative analysis of experimental data was conducted in terms of middle period of half-life of delayed neutron nuclei-precursors. Character and scale of changing values of delayed neutron group parameters as changing excitation energy of fission compound-nucleus have been demonstrated for the first time. Considerable energy dependence of group parameters under the neutron induced 239 Pu fission that was expressed by the decreasing middle period of half-life of nuclei-precursors by 10 % in the 2.85 eV - 5 MeV range of virgin neutrons was detected [ru

  13. The determination of neutron energy spectra of radioisotope sources

    International Nuclear Information System (INIS)

    Lutkin, J.E.

    1975-08-01

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

  14. Energy dependence of relative abundances and periods of separate groups of delayed neutrons at neutron induced fission of 239Pu in a range of neutrons energies 0.37 - 5 MeV

    International Nuclear Information System (INIS)

    Roschenko, V.A.; Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Tarasko, M.Z.; Tertychnyi, R.G.

    2001-01-01

    The fundamental role of delayed neutrons in behavior, control and safety of reactors is well known today. Delayed neutron data are of great interest not only for reactor physics but also for nuclear fission physics and astrophysics. The purpose of the present work was the measurement of energy dependence of delayed neutrons (DN) group parameters at fission of nuclei 239 Pu in a range of energies of primary neutrons from 0.37 up to 5 MeV. The measurements were executed on installation designed on the basis of the electrostatic accelerator of KG - 2.5 SSC RF IPPE. The data are obtained in 6-group representation. It is shown, that there is a significant energy dependence of DN group parameters in a range of primary neutrons energies from thermal meanings up to 5 MeV, which is expressed in reduction of the average half-life of nuclei of the DN precursors on 10 %. The data, received in the present work, can be used at creation of a set of group constants for reactors with an intermediate spectrum of neutrons. (authors)

  15. Delayed neutron yield from fast neutron induced fission of 238U

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  16. Neutron energy spectra calculations in the low power research reactor

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. Neutron dose and energy spectra measurements at Savannah River Plant

    International Nuclear Information System (INIS)

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

    1987-08-01

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

  18. Catalogue of neutron spectra

    International Nuclear Information System (INIS)

    Buxerolle, M.; Massoutie, M.; Kurdjian, J.

    1987-09-01

    Neutron dosimetry problems have arisen as a result of developments in the applications of nuclear energy. The largest number of possible irradiation situations has been collected: they are presented in the form of a compilation of 44 neutron spectra. Diagrams show the variations of energy fluence and energy fluence weighted by the dose equivalent/fluence conversion factor, with the logarithm of the corresponding energy. The equivalent dose distributions are presented as percentages for the following energy bins: 0.01 eV/0.5 eV/50 keV/1 MeV/5 MeV/15 MeV. The dose equivalent, the mean energy and the effective energy for the dose equivalent for 1 neutron cm -2 are also given [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-11

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-11

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

  1. A parametric model to describe neutron spectra around high-energy electron accelerators and its application in neutron spectrometry with Bonner Spheres

    Science.gov (United States)

    Bedogni, Roberto; Pelliccioni, Maurizio; Esposito, Adolfo

    2010-03-01

    Due to the increased interest of the scientific community in the applications of synchrotron light, there is an increasing demand of high-energy electron facilities, testified by the construction of several new facilities worldwide. The radiation protection around such facilities requires accurate experimental methods to determine the dose due to prompt radiation fields. Neutron fields, in particular, are the most complex to measure, because they extend in energy from thermal (10 -8 MeV) up to hundreds MeV and because the responses of dosemeters and survey meters usually have large energy dependence. The Bonner Spheres Spectrometer (BSS) is in practice the only instrument able to respond over the whole energy range of interest, and for this reason it is frequently used to derive neutron spectra and dosimetric quantities in accelerator workplaces. Nevertheless, complex unfolding algorithms are needed to derive the neutron spectra from the experimental BSS data. This paper presents a parametric model specially developed for the unfolding of the experimental data measured with BSS around high-energy electron accelerators. The work consists of the following stages: (1) Generation with the FLUKA code, of a set of neutron spectra representing the radiation environment around accelerators with different electron energies; (2) formulation of a parametric model able to describe these spectra, with particular attention to the high-energy component (>10 MeV), which may be responsible for a large part of the dose in workplaces; and (3) implementation of this model in an existing unfolding code.

  2. A parametric model to describe neutron spectra around high-energy electron accelerators and its application in neutron spectrometry with Bonner Spheres

    International Nuclear Information System (INIS)

    Bedogni, Roberto; Pelliccioni, Maurizio; Esposito, Adolfo

    2010-01-01

    Due to the increased interest of the scientific community in the applications of synchrotron light, there is an increasing demand of high-energy electron facilities, testified by the construction of several new facilities worldwide. The radiation protection around such facilities requires accurate experimental methods to determine the dose due to prompt radiation fields. Neutron fields, in particular, are the most complex to measure, because they extend in energy from thermal (10 -8 MeV) up to hundreds MeV and because the responses of dosemeters and survey meters usually have large energy dependence. The Bonner Spheres Spectrometer (BSS) is in practice the only instrument able to respond over the whole energy range of interest, and for this reason it is frequently used to derive neutron spectra and dosimetric quantities in accelerator workplaces. Nevertheless, complex unfolding algorithms are needed to derive the neutron spectra from the experimental BSS data. This paper presents a parametric model specially developed for the unfolding of the experimental data measured with BSS around high-energy electron accelerators. The work consists of the following stages: (1) Generation with the FLUKA code, of a set of neutron spectra representing the radiation environment around accelerators with different electron energies; (2) formulation of a parametric model able to describe these spectra, with particular attention to the high-energy component (>10 MeV), which may be responsible for a large part of the dose in workplaces; and (3) implementation of this model in an existing unfolding code.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  4. Correlation properties of delayed neutrons from fast neutron induced fission

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Isaev, S.G.

    1998-01-01

    The experimental studies of the energy dependence of the delayed neutron parameters for various fissioning systems has shown that the behavior of a some combination of delayed neutron parameters (group relative abundances a i and half lives T i ) has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. Below we will present the preliminary results which were obtained during this study omitting the physics interpretation of the results. (author)

  5. Neutron Energy Spectra from Neutron Induced Fission of 235U at 0.95 MeV and of 238U at 1.35 and 2.02 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Almen, E; Holmqvist, B; Wiedling, T

    1971-09-15

    The shapes of fission neutron spectra are of interest for power reactor calculations. Recently it has been suggested that the neutron induced fission spectrum of 235U may be harder than was earlier assumed. For this reason measurements of the neutron spectra of some fissile isotopes are in progress at our laboratory. This report will present results from studies of the energy spectra of the neutrons emitted in the neutron induced fission of 235U and 238U. The measurements were performed at an incident neutron energy of 0.95 MeV for 235U and at energies of 1.35 and 2.02 MeV for 238U using time-of-flight techniques. The time-of-flight spectra were only analysed at energies higher than those of the incident neutrons and up to about 10 MeV. Corrections for neutron attenuation in the uranium samples were calculated using a Monte Carlo program. The corrected fission neutron spectra were fitted to Maxwellian temperature distributions. For 235U a temperature of 1.27 +- 0.01 MeV gives the best fit to the experimental data and for 238U the corresponding values are 1.29 +- 0.03 MeV at 1.35 MeV and 1.29 +- 0.02 MeV at 2.02 MeV

  6. Statistical effects in beta-delayed neutron emission from fission product nuclides

    International Nuclear Information System (INIS)

    McElroy, R.D. Jr.

    1986-01-01

    The delayed neutron spectra for the precursors Rb-93, 94, 95, 96, 97 and Cs-145 were measured by use of the on-line isotope separator facility TRISTAN and a time-of-flight (TOF) spectrometer. Flight paths were used that provided, for energies below 70 keV, a FWHM energy resolution between 2 and 4 percent. Each spectrum showed discrete neutron peaks below 156 keV, with as many as 26 in the Rb-95 spectra. Level densities near the neutron binding energy in the neutron-emitting nuclide were deduced using a missing-level indicator based on a Porter-Thomas distribution of neutron peak intensities. The resulting level density data were compared to the predictions of the Gilbert and Cameron formulism and to those of Dilg, Schantl, Vonach and Uhl. Comparisons were made between the empirically-based level parameter a and the values predicted by each model for Sr-93, 94, 95, 97 and Ba-145. The two models appear, within the uncertainties, to be equally capable of describing these neutron-rich nuclides and equally as capable for them as they are for nuclides in the valley of beta stability. Measurements of the neutron strength function are sometimes possible with the present TOF system for neutron decays with competing neutron branches to levels in the grandchild nucleus. A value for the d-wave strength function of Sr-96 is found to be (4.2 +- 1.1)/10 4 . Improvements in the TOF system, allowing the measurement of the neutron strength function for the more general case, are discussed. 72 refs., 56 figs., 16 tabs

  7. Evaluation of double differential yield as used for representation of neutron spectra

    International Nuclear Information System (INIS)

    Solieman, A.H.M.; Comsan, M.N.H.

    2002-01-01

    The neutron intensity for TOF spectra representation has, until now, only been expressed in terms of double differential yield; number of neutrons per unit charge per unit solid angle per unit neutron energy interval (i.e. neutron intensity at a given resolving power). For accelerator-based neutron sources, the double differential yield - in terms of neutron energy interval - is found to be affected by the kinematics of the neutron producing reaction, to produce intensity irrelevant spectra. The results affect not only the applications that depend on relative neutron intensities, but also the applications that depend on the neutron intensity-weighted integration of the neutron spectra (e.g. neutron average energy calculation, and dose calculation using kerma factors). Other definition of the double differential yield - in terms of projectile energy loss - is suggested to avoid the drawbacks of the old definition. The neutron spectra that are driven using the two definitions are discussed

  8. Prompt fission neutron spectra and average prompt neutron multiplicities

    International Nuclear Information System (INIS)

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

    1983-01-01

    We present a new method for calculating the prompt fission neutron spectrum N(E) and average prompt neutron multiplicity anti nu/sub p/ as functions of the fissioning nucleus and its excitation energy. The method is based on standard nuclear evaporation theory and takes into account (1) the motion of the fission fragments, (2) the distribution of fission-fragment residual nuclear temperature, (3) the energy dependence of the cross section sigma/sub c/ for the inverse process of compound-nucleus formation, and (4) the possibility of multiple-chance fission. We use a triangular distribution in residual nuclear temperature based on the Fermi-gas model. This leads to closed expressions for N(E) and anti nu/sub p/ when sigma/sub c/ is assumed constant and readily computed quadratures when the energy dependence of sigma/sub c/ is determined from an optical model. Neutron spectra and average multiplicities calculated with an energy-dependent cross section agree well with experimental data for the neutron-induced fission of 235 U and the spontaneous fission of 252 Cf. For the latter case, there are some significant inconsistencies between the experimental spectra that need to be resolved. 29 references

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  10. Neutron spectra unfolding in Bonner spheres spectrometry using neural networks

    International Nuclear Information System (INIS)

    Kardan, M.R.; Setayeshi, S.; Koohi-Fayegh, R.; Ghiassi-Nejad, M.

    2003-01-01

    The neural network method has been used for the unfolding of neutron spectra in neutron spectrometry by Bonner spheres. A back propagation algorithm was used for training of neural networks 4mm x 4 mm bare LiI(Eu) and in a polyethylene sphere set: 2, 3, 4, 5, 6, 7, 8, 10, 12, 18 inch diameter have been used for unfolding of neutron spectra. Neural networks were trained by 199 sets of neutron spectra, which were subdivided into 6, 8, 10, 12, 15 and 20 energy bins and for each of them an appropriate neural network was designed and trained. The validation was performed by the 21 sets of neutron spectra. A neural network with 10 energy bins which had a mean value of error of 6% for dose equivalent estimation of spectra in the validation set showed the best results. The obtained results show that neural networks can be applied as an effective method for unfolding neutron spectra especially when the main target is neutron dosimetry. (author)

  11. Study of neutron rich nuclei by delayed neutron decay using the Tonnerre multidetector; Etude de la decroissance par neutrons retardes de noyaux legers riches en neutrons avec le multidetecteur tonnerre

    Energy Technology Data Exchange (ETDEWEB)

    Timis, C.N

    2001-07-01

    A new detection array for beta delayed neutrons was built. It includes up to 32 plastic scintillation counters 180 cm long located at 120 cm from the target. Neutron energy spectra are measured by time-of-flight in the 300 keV-15 MeV range with good energy resolution. The device was tested with several known nuclei. Its performances are discussed in comparison with Monte Carlo simulations. They very high overall detection efficiency on the TONNERRE array made it possible to study one and two neutron emission of {sup 11}Li. A complete decay scheme was obtained. The {sup 33}Mg and {sup 35}Al beta decays were investigated for the first time by neutron and gamma spectroscopy. Complete decay schemes were established and compared to large scale shell-model calculations. (authors)

  12. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  13. Evaluation method for uncertainty of effective delayed neutron fraction βeff

    International Nuclear Information System (INIS)

    Zukeran, Atsushi

    1999-01-01

    Uncertainty of effective delayed neutron fraction β eff is evaluated in terms of three quantities; uncertainties of the basic delayed neutron constants, energy dependence of delayed neutron yield ν d m , and the uncertainties of the fission cross sections of fuel elements. The uncertainty of β eff due to the delayed neutron yield is expressed by a linearized formula assuming that the delayed neutron yield does not depend on the incident energy, and the energy dependence is supplemented by using the detailed energy dependence proposed by D'Angelo and Filip. The third quantity, uncertainties of fission cross section, is evaluated on the basis of the generalized perturbation theory in relation to reaction rate rations such as central spectral indexes or average reaction rate ratios. Resultant uncertainty of β eff is about 4 to 5%s, in which primary factor is the delayed neutron yield, and the secondary one is the fission cross section uncertainty, especially for 238 U. The energy dependence of ν d m systematically reduces the magnitude of β eff about 1.4% to 1.7%, depending on the model of the energy vs. ν d m correlation curve. (author)

  14. JENDL-4.0 benchmarking for effective delayed neutron fraction with a continuous-energy Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu

    2013-01-01

    Benchmark calculations with a continuous-energy Monte Carlo code have been performed for delayed neutron data of JENDL-4.0. JENDL-4.0 gives good prediction for the effective delayed neutron fraction in the present benchmarks but further detailed analysis is required for some cores. (author)

  15. Measurement of neutron spectra for photonuclear reaction with linearly polarized photons

    Directory of Open Access Journals (Sweden)

    Kirihara Yoichi

    2017-01-01

    Full Text Available Spectra of neutrons produced by a photonuclear reaction from a 197Au target were measured using 16.95 MeV linearly and circularly polarized photon beams at NewSUBARU-BL01 using a time-of-flight method. The difference in the neutron spectra between the cases of a linearly and circularly polarized photon was measured. The difference in the neutron yield increased with the neutron energy and was approximately threefold at the maximum neutron energy. In a direction perpendicular to that of the linear polarization, the neutron yields decreased as the neutron energy increased.

  16. Time-of-flight spectrometer for the measurement of gamma correlated neutron spectra

    International Nuclear Information System (INIS)

    Andriashin, A.V.; Devkin, B.V.; Lychagin, A.A.; Minko, J.V.; Mironov, A.N.; Nesterenko, V.S.; Sztaricskai, T.; Petoe, G.; Vasvary, L.

    1986-01-01

    A time-of-flight spectrometer for the measurement of gamma correlated neutron spectra from (n,xnγ) reactions is described. The operation and the main parameters are discussed. The resolution in the neutron channel is 2.2 ns/m at the 150 keV neutron energy threshold. A simultaneous measurement of the time-of-flight and amplitude distributions makes it possible to study gamma correlated neutron spectra as well as the prompt gamma spectra in coincidence with selected energy neutrons. In order to test the spectrometer, measurements of the neutron spectrum in coincidence with the 846 keV gamma line of 56 Fe were carried out at an incident neutron energy of 14.1 MeV. (Auth.)

  17. Time-of-flight spectrometer for the measurement of gamma correlated neutron spectra

    International Nuclear Information System (INIS)

    Andryashin, A.V.; Devlein, B.V.; Lychagin, A.A.; Minko, Y.V.; Mironov, A.N.; Nesterenko, V.S.

    1986-01-01

    A time-of-flight spectrometer for the measurement of gamma correlated neutron spectra form (n,xnγ) reactions is described. The operation and the main parameters are discussed. The resolution in the neutron channel is 2.2 ns/m at the 150 keV neutron energy threshold. A simultaneous measurement of the time-of-flight and amplitude distributions makes it possible to study gamma correlated neutron spectra as well as the prompt gamma spectra in coincidence with selected energy neutrons. In order to test the spectrometer, measurements of the neutron spectrum in coincidence with the 846 keV gamma line of 56 Fe were carried out at an incident neutron energy of 14.1 MeV. (author). 3 figs., 6 refs

  18. Neutron flux density and secondary-particle energy spectra at the 184-inch synchrocyclotron medical facility

    International Nuclear Information System (INIS)

    Smith, A.R.; Schimmerling, W.; Henson, A.M.; Kanstein, L.L.; McCaslin, J.B.; Stephens, L.D.; Thomas, R.H.; Ozawa, J.; Yeater, F.W.

    1978-07-01

    Helium ions, with an energy of 920 MeV, produced by the 184-inch synchrocyclotron of the Lawrence Berkeley Laboratory are now being used in a pilot series to determine their efficacy in the treatment of tumors of large volume. The techniques for production of the large uniform radiation fields required for these treatments involve the use of beam-limiting collimators and energy degraders. Interaction of the primary beam with these beam components produces secondary charged particles and neutrons. The sources of neutron production in the beam transport system of the alpha-particle beam have been identified and their magnitudes have been determined. Measurements with activation detectors and pulse counters of differing energy responses have been used to determine secondary particle spectra at various locations on the patient table. These spectra are compared to a calculation of neutron production based on best estimates derived from published cross sections. Agreement between the calculated spectra and those derived from experimental measurements is obtained (at the 10 to 20% level) when the presence of charged particles is taken into account. The adsorbed dose in soft tissue is not very sensitive to the shape of the incident neutron energy spectrum, and the values obtained from unfolding the experimental measurements agree with the values obtained from the calculated spectra within the estimated uncertainty of +-25%. These values are about 3 x 10 -3 rad on the beam axis and about 1 x 10 -3 rad at 20 cm or more from the beam axis, per rad deposited by the incident alpha-particle beam. Estimates of upper limit dose to the lens of the eye and red bone marrow are approximately 10 rad and approximately 1 rad, respectively, for a typical treatment plan. The absorbed dose to the lens of the eye is thus well below the threshold value for cataractogenesis estimated for fission neutrons. An upper limit for the risk of leukemia is estimated to be approximately 0.04%

  19. Inclusive sum rules and spectra of neutrons at the ISR

    International Nuclear Information System (INIS)

    Grigoryan, A.A.

    1975-01-01

    Neutron spectra in pp collisions at ISR energies are studied in the framework of sum rules for inclusive processes. The contributions of protons, π- and E- mesons to the energy sum rule are calculated at √5 = 53 GeV. It is shown by means of this sum rule that the spectra of neutrons at the ISR are in contradiction with the spectra of other particles also measured at the ISR

  20. 8-group relative delayed neutron yields for monoenergetic neutron induced fission of 239Pu

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  1. Mechanical approach to the neutrons spectra collimation and detection

    Energy Technology Data Exchange (ETDEWEB)

    Sadeghi, H.; Roshan, M. V. [Energy Engineering and Physics Department, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of)

    2014-11-15

    Neutrons spectra from most of known sources require being collimated for numerous applications; among them one is the Neutron Activation Analysis. High energy neutrons are collimated through a mechanical procedure as one of the most promising methods. The output energy of the neutron beam depends on the velocity of the rotating Polyethylene disks. The collimated neutrons are then measured by an innovative detection technique with high accuracy.

  2. 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)

  3. Neutron stochastic transport theory with delayed neutrons

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  4. CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident

    International Nuclear Information System (INIS)

    Bessis, J.

    1976-01-01

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

  5. Fission neutron spectra measurements at LANSCE - Status and plans

    International Nuclear Information System (INIS)

    Haight, R. C.; Noda, S.; Nelson, R. O.; O'Donnell, J. M.; Devlin, M.; Chatillon, A.; Granier, T.; Taiebb, J.; Laurent, B.; Belier, G.; Becker, J. A.; Wu, C. Y.

    2010-01-01

    A program to measure fission neutron spectra from neutron-induced fission of actinides is underway at the Los Alamos Neutron Science Center (LANSCE) in a collaboration among the CEA laboratory at Bruyeres-le-Chatel, Lawrence Livermore National Laboratory and Los Alamos National Laboratory. The spallation source of fast neutrons at LANSCE is used to provide incident neutron energies from less than 1 MeV to 100 MeV or higher. The fission events take place in a gas-ionization fission chamber, and the time of flight from the neutron source to that chamber gives the energy of the incident neutron. Outgoing neutrons are detected by an array of organic liquid scintillator neutron detectors, and their energies are deduced from the time of flight from the fission chamber to the neutron detector. Measurements have been made of the fission neutrons from fission of 235 U, 238 U, 237 Np and 239 Pu. The range of outgoing energies measured so far is from 0.7 MeV to approximately 8 MeV. These partial spectra and average fission neutron energies are compared with evaluated data and with models of fission neutron emission. Results to date are summarized in this presentation. Future plans are to make significant improvements in the fission chambers, neutron detectors, signal processing, data acquisition and the experimental environment to provide high fidelity data including measurements of fission neutrons below 0.7 MeV and improvements in the data above 8 MeV. (authors)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-10-01

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

  8. The neutron long counter NERO for studies of β-delayed neutron emission in the r-process

    International Nuclear Information System (INIS)

    Pereira, J.; Hosmer, P.; Lorusso, G.; Santi, P.; Couture, A.; Daly, J.; Del Santo, M.; Elliot, T.

    2010-01-01

    The neutron long counter NERO was built at the National Superconducting Cyclotron Laboratory (NSCL), Michigan State University, for measuring β-delayed neutron-emission probabilities. The detector was designed to work in conjunction with a β-delay implantation station, so that β decays and β-delayed neutrons emitted from implanted nuclei can be measured simultaneously. The high efficiency of about 40%, for the range of energies of interest, along with the small background, are crucial for measuring β-delayed neutron emission branchings for neutron-rich r-process nuclei produced as low intensity fragmentation beams in in-flight separator facilities.

  9. A high-resolution neutron spectra unfolding method using the Genetic Algorithm technique

    CERN Document Server

    Mukherjee, B

    2002-01-01

    The Bonner sphere spectrometers (BSS) are commonly used to determine the neutron spectra within various nuclear facilities. Sophisticated mathematical tools are used to unfold the neutron energy distribution from the output data of the BSS. This paper highlights a novel high-resolution neutron spectra-unfolding method using the Genetic Algorithm (GA) technique. The GA imitates the biological evolution process prevailing in the nature to solve complex optimisation problems. The GA method was utilised to evaluate the neutron energy distribution, average energy, fluence and equivalent dose rates at important work places of a DIDO class research reactor and a high-energy superconducting heavy ion cyclotron. The spectrometer was calibrated with a sup 2 sup 4 sup 1 Am/Be (alpha,n) neutron standard source. The results of the GA method agreed satisfactorily with the results obtained by using the well-known BUNKI neutron spectra unfolding code.

  10. Proceedings of the specialists' meeting on delayed neutron nuclear data

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    1999-07-01

    This report is the Proceedings of the Specialists' Meeting on Delayed Neutron Nuclear Data. The meeting was held on January 28-29, 1999, at the Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of thirty specialists, who are evaluators, theorist, experimentalists. Although the fraction of the delayed neutron is no more than 1% in the total neutrons emitted in the fission process, it plays an important roll in the control of fission reactor. In the meeting, the following topics were reported: the present status of delayed neutron data in the major evaluated data libraries, measurements of effective delayed neutron fraction using FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly) and their analyses, sensitivity analysis for fast reactor, measurements of delayed neutron emission from actinides and so on. As another topics, delayed neutron in transmutation system and fission yield data were also presented. Free discussion was held on the future activity of delayed neutron data evaluation. The discussion was helpful for the future activity of the delayed neutron working group of JNDC aiming to the evaluation of delayed neutron data for JENDL-3.3. The 15 of the presented papers are indexed individually. (J.P.N.)

  11. Fission neutron spectra measurements at LANSCE - status and plans

    International Nuclear Information System (INIS)

    Haight, Robert C.; Noda, Shusaku; Nelson, Ronald O.; O' Donnell, John M.; Devlin, Matt; Chatillon, Audrey; Granier, Thierry; Taieb, Julien; Laurent, Benoit; Belier, Gilbert; Becker, John A.; Wu, Ching-Yen

    2009-01-01

    A program to measure fission neutron spectra from neutron-induced fission of actinides is underway at the Los Alamos Neutron Science Center (LANSCE) in a collaboration among the CEA laboratory at Bruyeres-le-Chatel, Lawrence Livermore National Laboratory and Los Alamos National Laboratory. The spallation source of fast neutrons at LANSCE is used to provide incident neutron energies from less than 1 MeV to 100 MeV or higher. The fission events take place in a gas-ionization fission chamber, and the time of flight from the neutron source to that chamber gives the energy of the incident neutron. Outgoing neutrons are detected by an array of organic liquid scintillator neutron detectors, and their energies are deduced from the time of flight from the fission chamber to the neutron detector. Measurements have been made of the fission neutrons from fission of 235 U, 238 U, 237 Np and 239 Pu. The range of outgoing energies measured so far is from 1 MeV to approximately 8 MeV. These partial spectra and average fission neutron energies are compared with evaluated data and with models of fission neutron emission. Results to date will be presented and a discussion of uncertainties will be given in this presentation. Future plans are to make significant improvements in the fission chambers, neutron detectors, signal processing, data acquisition and the experimental environment to provide high fidelity data including mea urements of fission neutrons below 1 MeV and improvements in the data above 8 MeV.

  12. Peculiarities of approximation for reactor neutron energy spectra during computerized simulation of radiation defects

    International Nuclear Information System (INIS)

    Kupchishin, A.A.; Kupchishin, A.I.; Stusik, G.; Omarbekova, Zh.

    2001-01-01

    Peculiarities of approximation for reactor neutron energy spectra during radiation defects computerized simulation were discussed. Approximation of neutron spectra N(E) was carried out by N(E)=α·exp(-β·E)·sh(γ·E) formula (1), where α, β, γ - approximation coefficients. In the capacity of operating reactor data experimental data on 235 U and 239 Pu were applied. The algorithm was designed, and acting soft ware for spectra parameters calculation was developed. The following values of approximation parameters were obtained: α=80.8; β=0.935;γ=2.04 (for uranium and plutonium these coefficients are less distinguishing). Then with use of formula 1 and α, β, γ coefficients the approximation curves were constructed. These curves satisfactorily describe existing experimental data and allowing to use its for radiation defects simulation in the reactor materials

  13. Measurement of thermal neutron spectra using LINAC in Japan Atomic Energy Research Institute (JAERI)

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1982-01-01

    The exact grasp of thermal neutron spectra in a core region is very important for obtaining accurate thermal neutron group constants in the calculation for the nuclear design of a reactor core. For the accurate grasp of thermal neutron spectra, the capability of thermal neutron spectra to describe the moderator cross-sections for thermal neutron scattering is a key factor. Accordingly, 0 deg angular thermal neutron spectra were measured by the time of flight (TOF) method using the JAERI LINAC as a pulsed neutron source, for light water system added with Cd and In, high temperature graphite system added with boron, and light water-natural uranium heterogeneous multiplication system among the reactor moderators of light water or graphite systems. First, the equations to give the time of flight and neutron flux by TOF method were analyzed, and several corrections were investigated, such as those for detector efficiency, background, the transmission coefficient of air and the Al window of a flight tube, mean emission time of neutrons, and the distortion effect of re-entrant hole on thermal neutron spectra. Then, the experimental system, results and calculation were reported for the experiments on the above three moderator systems. Finally, the measurement of fast neutron spectra in natural uranium system and that of the efficiency of a 6 Li glass scintillator detector are described. (Wakatsuki, Y.)

  14. A survey of neutron energy spectra and angular distributions of the 9Be(p,n)9B reaction for fast neutron radiotherapy

    International Nuclear Information System (INIS)

    Allab, M.

    1984-03-01

    Encouraging findings in radiobiology have stimulated a renewed use of fast neutrons in radiotherapy. The physical characteristics required for neutron beams to be suitable for radiotherapy are well established. As a result, the tendency is to replace the previous machines which generated the neutron beams from deuteron bombardment of thick targets (T, Li, Be) by hospital based cyclotrons which accelerate protons on thick Beryllium targets. This report surveys the available experimental data of the 9 Be(p,n) reaction (cross sections, neutron spectra, yields, mean neutron energies) from the threshold to the proton energy Esub(p)=120 MeV and the works using this reaction in dosimetry measurements, with an emphasis on the data since 1977

  15. Eigenvalue-dependent neutron energy spectra: Definitions, analyses, and applications

    International Nuclear Information System (INIS)

    Cacuci, D.G.; Ronen, Y.; Shayer, Z.; Wagschal, J.J.; Yeivin, Y.

    1982-01-01

    A general qualitative analysis of spectral effects that arise from solving the kappa-, α-, γ-, and sigma-eigenvalue formulations of the neutron transport equation for nuclear systems that deviate (to first order) from criticality is presented. Hierarchies of neutron spectra softness are established and expressed concisely in terms of the newly introduced spatialdependent local spectral indices for the core and for the reflector. It is shown that each hierarchy is preserved, regardless of the nature of the specific physical mechanism that cause the system to deviate from criticality. Qualitative conclusions regarding the general behavior of the spectrum-dependent integral spectral indices and ICRs corresponding to the kappa-, α-, γ-, and sigma-eigenvalue formalisms are also presented. By defining spectral indices separately for the core and for the reflector, it is possible to account for the characteristics of neutron spectra in both the core and the reflector. The distinctions between the spectra in the core and in the reflector could not have been accounted for by using a single type of spectral index (e.g., a spectral index for the entire system or a spectral index solely for the core)

  16. Unfolding of neutron spectra from Godiva type critical assemblies

    International Nuclear Information System (INIS)

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

    1976-01-01

    The results from three experiments conducted at the White Sands Missile Range Fast Burst Reactor Facility are discussed. The experiments were designed to measure the ''free-field'' neutron leakage spectrum and the neutron spectra from mildly perturbed environments. SAND-II was used to calculate the neutron spectrum utilizing several different trial input spectra for each experiment. Comparisons are made between the unfolded neutron spectrum for each trial input on the basis of the following parameters: average neutron energy (above 10 KeV), integral fluence (above 10 KeV), spectral index and the hardness parameter, phi/sub eq//phi

  17. Determination of the fast neutrons spectra by the Elastic scattering method (n, p)

    International Nuclear Information System (INIS)

    Elizalde D, J.

    1973-01-01

    This work consists in determining the fast neutron spectra emitted by a Pu-Be isotopic source. The implemented technique is based in the spectrometry (n, p). This consists in making to fall on a fast neutrons beams (polyenergetic) over a thin film of hydrogenated material, detecting the spectra of emitted protons at a fix angle. The polyethylene film and the used solid state detector are inside of a vacuum chamber. The detector is placed at 30 degree with respect to direction of the incident neutrons beam. The protons spectra is stored in a multichannel. the energy is obtained with the prior calibration of the system. The data processing involves the transformation of the protons spectra observed at the falling on neutrons spectra over the film. The energy of the neutrons is related with that of the protons, according to the collision kinematical equations. The cross section of elastic collision of the neutrons with the hydrogen atoms is obtained from literature. Applying these relations to the observed spectra it is obtained the falling on neutron spectra over the film. (Author)

  18. Calculated microdose spectra for intermediate energy neutrons (1 to 100 keV)

    International Nuclear Information System (INIS)

    Al-Affan, I.A.M.; Watt, D.E.

    1983-01-01

    Basic formulae for calculation of energy deposition events due to insiders, starters, stoppers and crossers, using the continuous slowing down approximation have been modified to allow for the enhanced energy deposition in spherical volumes due to elastic scattering interactions which reduce the penetration depth of the charged particle recoils. Energy deposition spectra have been obtained for energies of 1, 10, 50, 100 keV in 0.2 μm and 1 μm tissue-equivalent spheres. From these, frequency and dose distributions in lineal energy and in specific energy density have been calculated. Also calculated for different neutron energies are values of zeta, the energy average of event size, as a function of the diameter of the sensitive site. The structure of the energy event distributions can be interpreted in terms of the basic physics. The effect of the modifications to the basic formulae is to increase the number of energy deposition events due to insiders and to decrease the number of starters, stoppers and crossers. The degree of the effect increases with decreasing neutron energy, increasing sphere size, and the change is most significant for low energy deposition events. (author)

  19. Calculated microdose spectra for intermediate energy neutrons (1 to 100 keV)

    Energy Technology Data Exchange (ETDEWEB)

    Al-Affan, I.A.M.; Watt, D.E. (Dundee Univ. (UK). Dept. of Medical Biophysics); Colautti, P.; Talpo, G. (Laboratori Nazionali dell' Infn, 35020, Legnaro (Padova) (Italy))

    1983-01-01

    Basic formulae for calculation of energy deposition events due to insiders, starters, stoppers and crossers, using the continuous slowing down approximation have been modified to allow for the enhanced energy deposition in spherical volumes due to elastic scattering interactions which reduce the penetration depth of the charged particle recoils. Energy deposition spectra have been obtained for energies of 1, 10, 50, 100 keV in 0.2 ..mu..m and 1 ..mu..m tissue-equivalent spheres. From these, frequency and dose distributions in lineal energy and in specific energy density have been calculated. Also calculated for different neutron energies are values of zeta, the energy average of event size, as a function of the diameter of the sensitive site. The structure of the energy event distributions can be interpreted in terms of the basic physics. The effect of the modifications to the basic formulae is to increase the number of energy deposition events due to insiders and to decrease the number of starters, stoppers and crossers. The degree of the effect increases with decreasing neutron energy, increasing sphere size, and the change is most significant for low energy deposition events.

  20. Prompt fission neutron spectra of n + 235U above the (n, nf) fission threshold

    International Nuclear Information System (INIS)

    Shu Nengchuan; Chen Yongjing; Liu Tingjin; Jia Min

    2015-01-01

    Calculations of prompt fission neutron spectra (PFNS) from the 235 U(n, f) reaction were performed with a semi-empirical method for En = 7.0 and 14.7 MeV neutron energies. The total PFNS were obtained as a superposition of (n, xnf) pre-fission neutron spectra and post-fission spectra of neutrons which were evaporated from fission fragments, and these two kinds of spectra were taken as an expression of the evaporation spectrum. The contributions of (n, xnf) fission neutron spectra on the calculated PFNS were discussed. The results show that emission of one or two neutrons in the (n, nf) or (n, 2nf) reactions influences the PFNS shape, and the neutron spectra of the (n, xnf) fission-channel are soft compared with the neutron spectra of the (n, f) fission channel. In addition, analysis of the multiple-chance fission component showed that second-chance fission dominates the PFNS with an incident neutron energy of 14.7 MeV whereas first-chance fission dominates the 7 MeV case. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. Proceedings of the specialists' meeting on delayed neutron nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Katakura, Jun-ichi [ed.] [Japanese Nuclear Data Committee, Tokai, Ibaraki (Japan)

    1999-07-01

    This report is the Proceedings of the Specialists' Meeting on Delayed Neutron Nuclear Data. The meeting was held on January 28-29, 1999, at the Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of thirty specialists, who are evaluators, theorist, experimentalists. Although the fraction of the delayed neutron is no more than 1% in the total neutrons emitted in the fission process, it plays an important roll in the control of fission reactor. In the meeting, the following topics were reported: the present status of delayed neutron data in the major evaluated data libraries, measurements of effective delayed neutron fraction using FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly) and their analyses, sensitivity analysis for fast reactor, measurements of delayed neutron emission from actinides and so on. As another topics, delayed neutron in transmutation system and fission yield data were also presented. Free discussion was held on the future activity of delayed neutron data evaluation. The discussion was helpful for the future activity of the delayed neutron working group of JNDC aiming to the evaluation of delayed neutron data for JENDL-3.3. The 15 of the presented papers are indexed individually. (J.P.N.)

  3. Fast neutron spectra determination by threshold activation detectors using neural networks

    International Nuclear Information System (INIS)

    Kardan, M.R.; Koohi-Fayegh, R.; Setayeshi, S.; Ghiassi-Nejad, M.

    2004-01-01

    Neural network method was used for fast neutron spectra unfolding in spectrometry by threshold activation detectors. The input layer of the neural networks consisted of 11 neurons for the specific activities of neutron-induced nuclear reaction products, while the output layers were fast neutron spectra which had been subdivided into 6, 8, 10, 12, 15 and 20 energy bins. Neural network training was performed by 437 fast neutron spectra and corresponding threshold activation detector readings. The trained neural network have been applied for unfolding 50 spectra, which were not in training sets and the results were compared with real spectra and unfolded spectra by SANDII. The best results belong to 10 energy bin spectra. The neural network was also trained by detector readings with 5% uncertainty and the response of the trained neural network to detector readings with 5%, 10%, 15%, 20%, 25% and 50% uncertainty was compared with real spectra. Neural network algorithm, in comparison with other unfolding methods, is very fast and needless to detector response matrix and any prior information about spectra and also the outputs have low sensitivity to uncertainty in the activity measurements. The results show that the neural network algorithm is useful when a fast response is required with reasonable accuracy

  4. Quasi-monoenergetic neutron energy spectra for 246 and 389 MeV (7)Li(p,n) reactions at angles from 0 degrees to 300 degrees

    CERN Document Server

    Iwamoto, Y; Nakamura, T; Nakashima, H; Mares, V; Itoga, T; Matsumoto, T; Nakane, Y; Feldbaumer, E; Jaegerhofer, L; Pioch, C; Tamii, A; Satoh, D; Masuda, A; Sato, T; Iwase, H; Yashima, H; Nishiyama, J; Hagiwara, M; Hatanaka, K; Sakamoto, Y

    2011-01-01

    The authors measured the neutron energy spectra of a quasi-monoenergetic (7)Li(p,n) neutron source with 246 and 389 MeV protons set at seven angles (0 degrees, 2.5 degrees, 5 degrees, 10 degrees, 15 degrees, 20 degrees and 30 degrees), using a time-of-flight (TOF) method employing organic scintillators NE213 at the Research Center for Nuclear Physics (RCNP) of Osaka University. The energy spectra of the source neutrons were precisely deduced down to 2 MeV at 0 degrees and 10 MeV at other angles. The cross-sections of the peak neutron production reaction at 0 degrees were on the 35-40 mb line of other experimental data, and the peak neutron angular distribution agreed well with the Taddeucci formula. Neutron energy spectra below 100 MeV at all angles were comparable, but the shapes of the continuum above 150 MeV changed considerably with the angle. In order to consider the correction required to derive the response in the peak region from the measured total response for high-energy neutron monitors such as DAR...

  5. Calculation of Spectra of Neutrons and Charged Particles Produced in a Target of a Neutron Generator

    Science.gov (United States)

    Gaganov, V. V.

    2017-12-01

    An algorithm for calculating the spectra of neutrons and associated charged particles produced in the target of a neutron generator is detailed. The products of four nuclear reactions 3H( d, n)4He, 2H( d, n)3He, 2H( d, p)3H, and 3He( d, p)4He are analyzed. The results of calculations are presented in the form of neutron spectra for several emission angles and spectra of associated charged particles emitted at an angle of 180° for a deuteron initial energy of 0.13 MeV.

  6. Fast neutron spectra unfolding with SAND-11 and maximum likelihoed methods

    International Nuclear Information System (INIS)

    Bondars, Kh.Ya.; Kamnev, V.A.; Lapenas, A.A.; Troshin, V.S.

    1980-01-01

    Mutual comparison of the methods SAND-II and maximal likeness for neutron spectra determination are represented. Spectra were restored according to the measures reaction rate of ten activation detectors using the device B-2 of the reactor BR-5 behind two thicknesses of steel-graphite shielding: Z=6.5 cm and Z=42.5 cm. The influence of earlier information on the results of neutron spectra determination was studied. Differential and integral energy dependences of neutron flux density for three initial spectra and two cross section libraries (BGS-1 and ZACRSS) are presented. The both methods yield close differential spectra (discrepancies < 10 %) when identical cross section libraries and reference spectra are used

  7. Different spectra with the same neutron source

    International Nuclear Information System (INIS)

    Vega C, H. R.; Ortiz R, J. M.; Hernandez D, V. M.; Martinez B, M. R.; Hernandez A, B.; Ortiz H, A. A.; Mercado, G. A.

    2010-01-01

    Using as source term the spectrum of a 239 Pu-Be source several neutron spectra have been calculated using Monte Carlo methods. The source term was located in the centre of spherical moderators made of light water, heavy water and polyethylene of different diameters. Also a 239 Pu-Be source was used to measure its neutron spectrum, bare and moderated by water. The neutron spectra were measured at 100 cm with a Bonner spheres spectrometer. Monte Carlo calculations were used to calculate the neutron spectra of bare and water-moderated spectra that were compared with those measured with the spectrometer. Resulting spectra are similar to those found in power plants with PWR, BWR and Candu nuclear reactors. Beside the spectra the dosimetric features were determined. Using moderators and a single neutron source can be produced neutron spectra alike those found in workplaces, this neutron fields can be utilized to calibrate neutron dosimeters and area monitors. (Author)

  8. Importance of delayed neutron data in transmutation system

    International Nuclear Information System (INIS)

    Tsujimoto, Kazufumi

    1999-01-01

    The accelerator-driven transmutation system has been studied at the Japan Atomic Energy Research Institute. This system is a hybrid system which consists of a high intensity accelerator, a spallation target and a subcritical core region. The subcritical core is driven by neutrons generated by spallation reaction in the target region. There is no control rod in this system, so the power is controlled only by proton beam current. The beam current to keep constant power change with effective multiplication factor of subcritical core. So, the evaluation of delayed neutron fraction which is strongly connected to the measurement of subcritical level is important factor in operation of accelerator-driven system. In this paper, important nuclides for the delayed neutron fraction of ADS will be discussed, moreover, present state of delayed neutron data in evaluated nuclear data library is presented. (author)

  9. Calculations of neutron spectra after neutron-neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-09-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  11. 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)

  12. Measurements of the prompt neutron spectra in 233U, 235U, 239Pu thermal neutron fission in the energy range of 0.01-5 MeV and in 252Cf spontaneous fission in the energy range of 0.01-10 MeV

    International Nuclear Information System (INIS)

    Starostov, B.I.; Semenov, A.F.; Nefedov, V.N.

    1978-01-01

    The measurement results on the prompt neutron spectra in 233 U, 235 U, 239 Pu thermal neutron fission in the energy range of 0.01-5 MeV and in 252 Cf spontaneous fission in the energy range of 0.01-10 MeV are presented. The time-of-flight method was used. The exceeding of the spectra over the Maxwell distributions is observed at E 252 Cf neutron fission spectra. The spectra analysis was performed after normalization of the spectra and corresponding Maxwell distributions for one and the same area. In the range of 0.05-0.22 MeV the yield of 235 U + nsub(t) fission neutrons is approximately 8 and approximately 15 % greater than the yield of 252 Cf and 239 Pu + nsub(t) fission neutrons, respectively. In the range of 0.3-1.2 MeV the yield of 235 U + nsub(t) fission neutrons is 8 % greater than the fission neutron yield in case of 239 Pu + nsub(t) fission. The 235 U + nsub(t) and 233 U + nsub(t) fission neutron spectra do not differ from one another in the 0.05-0.6 MeV range

  13. Systematic evaluation of prompt neutron spectra in fission

    International Nuclear Information System (INIS)

    Osawa, Takaaki

    1995-01-01

    To create the nuclear data fail JEND-32, the prompt fission neutron spectra X(E) of 233 U, 235 U, 238 U and 239 Pu were reevaluated and some improvement were added to the calculation models. We tried to extend the calculation method of fission spectra of nuclides with poor measurement data in consideration of increasing the importance of nuclear data of minor actinoids. We improved and extended the following five points. (1) On JENDL-3.1, the fission spectra of principal fissible materials had been calculated by the Modland-Nix model which the neutron emissions of fragments were calculated under the approximation of the constant inverse process cross section. In the paper, the spectra were calculated by the use of the inverse process cross section depend on the energy obtained by the calculation of the optical model. The result showed the increase of low energy components and the softening effect of spectra (2) On JENDL-3.1, the all fission processes were assumed to undergo (n,f) reaction. In the paper, they were calculated by the multi-chance fission such as (n, n'f), (n, 2nf) and (n, 3nf) etc. Softening of the spectra (En > 6 MeV) was obtained by this method. (3) The level density parameter (LDP) has been assumed as a = A/C in either case of light fragment (LF) and heavy fragment (HF) in the original Madland-Nix model. But we used LDP based on the Ignatyuk model under consideration of the shell effects of nuclear fragments, hence the neutron spectra of heavy fragments were hardening. (4) Nuclear temperature of both fragments had been assumed to be the same at original model, but now R T = Tm/TmH was derived to calculate them. The ratio of middle/both side components of spectra was changed. (5) Unknown neutron fission spectra of minor actinide were able to the assumed on the basis of Moriyama-Ohnishi model. (S.Y.)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Simakov, S.P.; Kaeppeler, F.

    2011-10-01

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

  16. Pulse-shape discrimination in radioanalytical methods. Part I. Delayed fission neutron counting

    International Nuclear Information System (INIS)

    Posta, S.; Vacik, J.; Hnatowicz, V.; Cervena, J.

    1999-01-01

    In this study the principle of pulse shape discrimination (PSD) has been employed in delayed fission neutron counting (DNC) method. Effective elimination of unwanted gamma background signals in measured radiation spectra has been proved. (author)

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

    CERN Document Server

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

    2001-01-01

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

  18. Proceedings of the symposium on measurements of neutron energy spectra using recoil proton proportional counters

    International Nuclear Information System (INIS)

    Urabe, Itsumasa

    1986-01-01

    This is a report of the symposium on measurements of neutron energy spectra using recoil proton proportional counters held at the Research Reactor Institute of Kyoto University on January 27 in 1986. An energy resolution, wall effects of response functions, n · γ discrimination methods and other fundamental properties of recoil proton counters are discussed for a new development of an application of this counter. (author)

  19. Measurement of fast neutron spectra. 1-2

    International Nuclear Information System (INIS)

    Kimura, Itsuro

    1976-01-01

    The present status of the techniques for the measurement of fast neutron spectra is reviewed with particular attention to the recent activities in Japan. The first section of this report defines the energy range of fast neutrons, and various techniques are classified into four groups. In the following sections, recent development in each group is reviewed. The first part is the integral method represented mainly by the activation method. The variation of this method is shortly reviewed, and some results of the spectrum measurement for JRR-4 (a thermal research reactor) and YAYOI (a fast neutron source reactor) are presented together with the results of computed spectra. The second part is the method of proton recoil. The improvement of a proportional counter by Ichimori is shortly reviewed. The use of liquid scintillator is also discussed together with the experimental and computational results of YAYOI benchmark spectra of fast neutrons transmitted through the layers of iron. The utilization of n-α or n-p reaction as a sandwitch counter is discussed in the third part. Measured and calculated spectra in the FCA (a fast critical assembly) core are presented as examples. The method of time-of-flight is discussed in the fourth part. Recent developments in Japan such as the method with a double-scintillation counter are shortly presented together with its block diagram. (Aoki, K.)

  20. Gamma-ray emission spectra from spheres with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Yamamoto, Junji; Kanaoka, Takeshi; Murata, Isao; Takahashi, Akito; Sumita, Kenji

    1989-01-01

    Energy spectra of neutron-induced gamma-rays emitted from spherical samples were measured using a 14 MeV neutron source. The samples in use were LiF, Teflon:(CF 2 ) n , Si, Cr, Mn, Co, Cu, Nb, Mo, W and Pb. A diameter of the sphere was either 40 or 60 cm. The gamma-ray energy in the emission spectra covered the range from 500 keV to 10 MeV. Measured spectra were compared with transport calculations using the nuclear data files of JENDL-3T and ENDF/B-IV. The agreements between the measurements and the JENDL-3T calculations were good in the emission spectra for the low energy gamma-rays from inelastic scattering. (author)

  1. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Aguilar, F.; Paredes, L.; Rivera M, T.

    2013-10-01

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  2. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

    2013-10-15

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  3. Status of measurements of fission neutron spectra of Minor Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Drapchinsky, L.; Shiryaev, B. [V.G. Khlopin Radium Inst., Saint Petersburg (Russian Federation)

    1997-03-01

    The report considers experimental and theoretical works on studying the energy spectra of prompt neutrons emitted in spontaneous fission and neutron induced fission of Minor Actinides. It is noted that neutron spectra investigations were done for only a small number of such nuclei, most measurements, except those of Cf-252, having been carried out long ago by obsolete methods and imperfectapparatus. The works have no detailed description of experiments, analysis of errors, detailed numerical information about results of experiments. A conclusion is made that the available data do not come up to modern requirements. It is necessary to make new measurements of fission prompt neutron spectra of transuranium nuclides important for the objectives of working out a conception of minor actinides transmutation by means of special reactors. (author)

  4. RDANN a new methodology to solve the neutron spectra unfolding problem

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

    The optimization processes known as Taguchi method and DOE methodology are applied to the design, training and testing of Artificial Neural Networks in the neutron spectrometry field, which offer potential benefits in the evaluation of the behavior of the net as well as the ability to examine the interaction of the weights and neurons inside the same one. In this work, the Robust Design of Artificial Neural Networks methodology is used to solve the neutron spectra unfolding problem, designing, training and testing an ANN using a set of 187 neutron spectra compiled by the International Atomic Energy Agency, to obtain the better neutron spectra unfolded from the Bonner spheres spectrometer's count rates. (Author)

  5. RDANN a new methodology to solve the neutron spectra unfolding problem

    International Nuclear Information System (INIS)

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

    2006-01-01

    The optimization processes known as Taguchi method and DOE methodology are applied to the design, training and testing of Artificial Neural Networks in the neutron spectrometry field, which offer potential benefits in the evaluation of the behavior of the net as well as the ability to examine the interaction of the weights and neurons inside the same one. In this work, the Robust Design of Artificial Neural Networks methodology is used to solve the neutron spectra unfolding problem, designing, training and testing an ANN using a set of 187 neutron spectra compiled by the International Atomic Energy Agency, to obtain the better neutron spectra unfolded from the Bonner spheres spectrometer's count rates. (Author)

  6. Measurement and analysis of neutron flux spectra in a neutronics mock-up of the HCLL test blanket module

    International Nuclear Information System (INIS)

    Klix, A.; Batistoni, P.; Boettger, R.; Lebrun-Grandie, D.; Fischer, U.; Henniger, J.; Leichtle, D.; Villari, R.

    2010-01-01

    Fast neutron and gamma-ray flux spectra and time-of-arrival spectra of slow neutrons have been measured in a neutronics mock-up of the European Helium-Cooled Lithium-Lead Test Blanket Module with the aim to validate nuclear cross-section data. The mock-up was irradiated with fusion peak neutrons from the DT neutron generator of the Technical University of Dresden. A well characterized cylindrical NE-213 scintillator was inserted into two positions in the LiPb/EUROFER assembly. Pulse height spectra from neutrons and gamma-rays were recorded from the NE-213 output. The spectra were then unfolded with experimentally obtained response matrices of the NE-213 detector. Time-of-arrival spectra of slow neutrons were measured with a 3 He counter placed in the mock-up, and the neutron generator was operated in pulsed mode. Monte Carlo calculations using the MCNP code and nuclear cross-section data from the JEFF-3.1.1 and FENDL-2.1 libraries were performed and the results are compared with the experimental results. A good agreement of measurement and calculation was found with some deviations in certain energy intervals.

  7. Neutron Energy Spectra and Yields from the 7Li(p,n) Reaction for Nuclear Astrophysics

    Science.gov (United States)

    Tessler, M.; Friedman, M.; Schmidt, S.; Shor, A.; Berkovits, D.; Cohen, D.; Feinberg, G.; Fiebiger, S.; Krása, A.; Paul, M.; Plag, R.; Plompen, A.; Reifarth, R.

    2016-01-01

    Neutrons produced by the 7Li(p, n)7Be reaction close to threshold are widely used to measure the cross section of s-process nucleosynthesis reactions. While experiments have been performed so far with Van de Graaff accelerators, the use of RF accelerators with higher intensities is planned to enable investigations on radioactive isotopes. In parallel, high-power Li targets for the production of high-intensity neutrons at stellar energies are developed at Goethe University (Frankfurt, Germany) and SARAF (Soreq NRC, Israel). However, such setups pose severe challenges for the measurement of the proton beam intensity or the neutron fluence. In order to develop appropriate methods, we studied in detail the neutron energy distribution and intensity produced by the thick-target 7Li(p,n)7Be reaction and compared them to state-of- the-art simulation codes. Measurements were performed with the bunched and chopped proton beam at the Van de Graaff facility of the Institute for Reference Materials and Measurements (IRMM) using the time-of-flight (TOF) technique with thin (1/8") and thick (1") detectors. The importance of detailed simulations of the detector structure and geometry for the conversion of TOF to a neutron energy is stressed. The measured neutron spectra are consistent with those previously reported and agree well with Monte Carlo simulations that include experimentally determined 7Li(p,n) cross sections, two-body kinematics and proton energy loss in the Li-target.

  8. Reconstruction of Time-Resolved Neutron Energy Spectra in Z-Pinch Experiments Using Time-of-flight Method

    International Nuclear Information System (INIS)

    Rezac, K.; Klir, D.; Kubes, P.; Kravarik, J.

    2009-01-01

    We present the reconstruction of neutron energy spectra from time-of-flight signals. This technique is useful in experiments with the time of neutron production in the range of about tens or hundreds of nanoseconds. The neutron signals were obtained by a common hard X-ray and neutron fast plastic scintillation detectors. The reconstruction is based on the Monte Carlo method which has been improved by simultaneous usage of neutron detectors placed on two opposite sides from the neutron source. Although the reconstruction from detectors placed on two opposite sides is more difficult and a little bit inaccurate (it followed from several presumptions during the inclusion of both sides of detection), there are some advantages. The most important advantage is smaller influence of scattered neutrons on the reconstruction. Finally, we describe the estimation of the error of this reconstruction.

  9. Consultants' meeting on prompt fission neutron spectra of major actinides. Summary report

    International Nuclear Information System (INIS)

    Capote Noy, R.; Maslov, V.; Bauge, E.; Ohsawa, T.; Vorobyev, A.; Chadwick, M.B.; Oberstedt, S.

    2009-01-01

    A Consultants' Meeting on 'Prompt Fission Neutron Spectra of Major Actinides' was held at IAEA Headquarters, Vienna, Austria, to discuss the adequacy and quality of the recommended prompt fission neutron spectra to be found in existing nuclear data applications libraries. These prompt fission neutron spectra were judged to be inadequate, and this problem has proved difficult to resolve by means of theoretical modelling. Major adjustments may be required to ensure the validity of such important data. There is a strong requirement for an international effort to explore and resolve these difficulties and recommend prompt fission neutron spectra and uncertainty covariance matrices for the actinides over the neutron energy range from thermal to 20 MeV. Participants also stressed that there would be a strong need for validation of the resulting data against integral critical assembly and dosimetry data. (author)

  10. Application of a Bonner sphere spectrometer for determination of the energy spectra of neutrons generated by ≈1 MJ plasma focus

    Czech Academy of Sciences Publication Activity Database

    Králík, M.; Krása, Josef; Velyhan, Andriy; Scholz, M.; Ivanova-Stanik, I.M.; Bienkowska, B.; Miklaszewski, R.; Schmidt, H.; Řezáč, K.; Klír, D.; Kravárik, J.; Kubeš, P.

    2010-01-01

    Roč. 81, č. 11 (2010), 113503/1-113503/5 ISSN 0034-6748 R&D Projects: GA MŠk LA08024 Grant - others:FP-6 EU(XE) RITA-CT2006-26095 Institutional research plan: CEZ:AV0Z10100523 Keywords : plasma focus * fusion DD neutrons * Bonner sphere spectrometer * energy spectra of scattered neutrons * unfolded and calculated spectra Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.598, year: 2010

  11. Survey of neutron spectra generated by the fission of heavy nuclei induced by fast neutrons

    International Nuclear Information System (INIS)

    Lovchikova, G.N.; Trufanov, A.M.

    1997-01-01

    A review of neutron fission spectra measurements is presented. This review and the results of this analysis was performed with the participation of the authors. It is shown that there is a need for additional measurements of the energy and angular distributions of secondary neutrons in order to improve the understanding of the neutron emission mechanism in fission. (author). 21 refs, 6 figs

  12. Test of statistical models of the ν-delayed neutron emission by application of the Monte Carlo method

    International Nuclear Information System (INIS)

    Ohm, H.

    1982-01-01

    Using the example of the delayed neutron spectrum of 24 s- 137 I the statistical model is tested in view of its applicability. A computer code was developed which simulates delayed neutron spectra by the Monte Carlo method under the assumption that the transition probabilities of the ν and the neutron decays obey the Porter-Thomas distribution while the distances of the neutron emitting levels are Wigner distribution. Gramow-Teller ν-transitions and simply forbidden ν-transitions from the preceding nucleus to the emitting nucleus were regarded. (orig./HSI) [de

  13. Neutron energy spectra of sup 2 sup 5 sup 2 Cf, Am-Be source and of the D(d,n) sup 3 He reaction

    CERN Document Server

    Sang Tae Park

    2003-01-01

    The neutron energy spectrum of the following sources were measured using a fast neutron spectrometer with the NE-213 liquid scintillator: sup 2 sup 5 sup 2 Cf, Am-Be and D(d,n) sup 3 He reaction from a 3 MeV Pelletron accelerator in Tokyo Institute of Technology. The measured proton recoil pulse height data of sup 2 sup 5 sup 2 Cf, Am-Be and D(d,n) sup 3 He were unfolded using the mathematical program to obtain the neutron energy spectrum. The sup 2 sup 5 sup 2 Cf and Am-Be neutron energy spectra were measured and the results obtained showed a good agreement with the spectra usually published in the literature. The neutron energy spectrum from D(d,n) sup 3 He was measured and the results obtained also showed a good agreement with the calculation by time of flight (TOF) methods. (author)

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  15. Measurement of very forward neutron energy spectra for 7 TeV proton--proton collisions at the Large Hadron Collider

    CERN Document Server

    Adriani, O.; Bonechi, L.; Bongi, M.; Castellini, G.; D'Alessandro, R.; Del Prete, M.; Haguenauer, M.; Itow, Y.; Kasahara, K.; Kawade, K.; Makino, Y.; Masuda, K.; Matsubayashi, E.; Menjo, H.; Mitsuka, G.; Muraki, Y.; Okuno, Y.; Papini, P.; Perrot, A-L.; Ricciarini, S.; Sako, T.; Sakurai, N.; Sugiura, Y.; Suzuki, T.; Tamura, T.; Tiberio, A.; Torii, S.; Tricomi, A.; Turner, W.C.; Zhou, Q.D.

    2015-01-01

    The Large Hadron Collider forward (LHCf) experiment is designed to use the LHC to verify the hadronic-interaction models used in cosmic-ray physics. Forward baryon production is one of the crucial points to understand the development of cosmic-ray showers. We report the neutron-energy spectra for LHC $\\sqrt{s}$ = 7 TeV proton--proton collisions with the pseudo-rapidity $\\eta$ ranging from 8.81 to 8.99, from 8.99 to 9.22, and from 10.76 to infinity. The measured energy spectra obtained from the two independent calorimeters of Arm1 and Arm2 show the same characteristic feature before unfolding the difference in the detector responses. We unfolded the measured spectra by using the multidimensional unfolding method based on Bayesian theory, and the unfolded spectra were compared with current hadronic-interaction models. The QGSJET II-03 model predicts a high neutron production rate at the highest pseudo-rapidity range similar to our results and the DPMJET 3.04 model describes our results well at the lower pseudo-...

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. Use of the response function in the analysis of complex neutron spectra

    International Nuclear Information System (INIS)

    Kegel, G.H.R.; Ciarcia, C.; Couchell, G.P.; Shao, J.

    1974-01-01

    Neutron time-of-flight spectra with overlapping peaks must be unfolded to yield contributions of individual neutron groups. This requires an accurate knowledge of the resolution profile of each group. It is also desirable to know the shape of the spectra of neutrons which were scattered more than once in the scatterer, so that corrections for multiple interactions can be made. These resolution profiles and spectra shapes are not readily available. We have developed a series of measures to account for these effects in our work. We monitor the neutron target thickness during target preparation with a separate time-of-flight spectrometer; we measure detector and accelerator time resolutions for different neutron energies using a thin target and we use computer codes to simulate those factors not amenable to direct measurement

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  19. Microdosimetric spectra measurements of JANUS neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, I.R.; Williamson, F.S.

    1985-01-01

    Neutron radiation from the JANUS reactor at Argonne National Laboratory is being used with increasing frequency for major biological experiments. The fast neutron spectrum has a Kerma-weighted mean energy of 0.8 MeV and low gamma-ray contamination. In 1984 the JANUS fission converter plate of highly enriched uranium was replaced by one made of low-enriched uranium. We recorded microdosimetric spectra at several different positions in the high-flux irradiation room of JANUS before the change of the converter plate. Each set of measurements consisted of spectra taken at three different site diameters (0.5, 1.0, and 5.0 ..mu..m) and in both ''attenuator up'' and ''attenuator down'' configurations. At two conventional dosimetry reference positions, two sets of measurements were recorded. At three biological reference positions, measurements simulating several biological irradiation conditions, were taken. The dose rate at each position was estimated and compared with dose rates obtained previously by conventional dosimetry. Comparison of the different measurements showed no major change in spectra as a function of position or irradiation condition. First results from similar sets of measurements recorded after the installment of the new converter plate indicate no major change in the spectra. 11 refs., 4 figs., 5 tabs.

  20. Microdosimetric spectra measurements of JANUS neutrons

    International Nuclear Information System (INIS)

    Marshall, I.R.; Williamson, F.S.

    1985-01-01

    Neutron radiation from the JANUS reactor at Argonne National Laboratory is being used with increasing frequency for major biological experiments. The fast neutron spectrum has a Kerma-weighted mean energy of 0.8 MeV and low gamma-ray contamination. In 1984 the JANUS fission converter plate of highly enriched uranium was replaced by one made of low-enriched uranium. We recorded microdosimetric spectra at several different positions in the high-flux irradiation room of JANUS before the change of the converter plate. Each set of measurements consisted of spectra taken at three different site diameters (0.5, 1.0, and 5.0 μm) and in both ''attenuator up'' and ''attenuator down'' configurations. At two conventional dosimetry reference positions, two sets of measurements were recorded. At three biological reference positions, measurements simulating several biological irradiation conditions, were taken. The dose rate at each position was estimated and compared with dose rates obtained previously by conventional dosimetry. Comparison of the different measurements showed no major change in spectra as a function of position or irradiation condition. First results from similar sets of measurements recorded after the installment of the new converter plate indicate no major change in the spectra. 11 refs., 4 figs., 5 tabs

  1. Cross sections and differential spectra for reactions of 2-20 MeV neutrons on /sup nat/Cr

    International Nuclear Information System (INIS)

    Blann, M.; Komoto, T.T.

    1988-01-01

    This report summarizes product yields, secondary n,p and α spectra, and γ-ray spectra calculated for incident neutrons of 2 to 20 MeV on /sup nat/Cr targets. Results are all from the code ALICE, using the version ALISO which does weighting of results for targets which are a mix of isotopes. Where natural isotopic targets are involved, yields and n,p,α spectra will be reported weighted over isotopic yields. Gamma-ray spectra, however, will be reported for the most abundant isotope. We present product yields versus incident neutron energy, n,p,α spectra versus incident neutron energy, and calculated γ-ray spectra

  2. Measurement of spectra and neutron fluxes on artificial earth satellites from the Cosmos series

    Science.gov (United States)

    Dudkin, V. Y.; Kovalev, Y. Y.; Novikova, M. R.; Potapov, Y. V.; Skvortsov, S. S.; Smirennyy, L. N.

    1975-01-01

    In 1966-1967 measurements were carried out at the altitudes of 200 to 400 km to determine the spectra and fluxes of fast neutrons inside the hermetically sealed artificial earth satellites of the Cosmos series. The detectors used were nuclear emulsions of the B9 and BR types and an emulsion of the P9 type, filled with Li and P. Spectra and fluxes of neutrons in the range of energies from thermal energies to 10 MeV are presented. Neutron doses are also estimated.

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

    International Nuclear Information System (INIS)

    Holden, N.E.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Pearlstein, S.

    1999-01-01

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

  5. Features of the neutron spectra accompanying the fission of actinide nuclei

    International Nuclear Information System (INIS)

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

    2000-01-01

    The spectra of fission neutrons from 238 U are measured by the time-of-flight technique at incident-neutron energies E n = 5.0 and 13.2 MeV. The data are compared with those obtained in the previous studies for 232 Th, 235,238 U, 237 Np at E n = 2.9 and 14.7 MeV; for 232 Th at E n = 14.6 and 17.7 MeV; for 238 U at 16.0 and 17.7 MeV. An excess of soft neutrons, which is observed in comparing experimental spectra for E n 13.2, 14.7, 16.0 and 17.7 MeV with the results of traditional theoretical calculations, is reproduced fairly well under the assumption that, at high excitation energies of a compound system, some part of post-fission neutrons can be emitted by nonaccelerated fragments [ru

  6. Calculated /alpha/-induced thick target neutron yields and spectra, with comparison to measured data

    International Nuclear Information System (INIS)

    Wilson, W.B.; Bozoian, M.; Perry, R.T.

    1988-01-01

    One component of the neutron source associated with the decay of actinide nuclides in many environments is due to the interaction of decay /alpha/ particles in (/alpha/,n) reactions on low Z nuclides. Measurements of (/alpha/,n) thick target neutron yields and associated neutron spectra have been made for only a few combinations of /alpha/ energy and target nuclide or mixtures of actinide and target nuclides. Calculations of thick target neutron yields and spectra with the SOURCES code require /alpha/-energy-dependent cross sections for (/alpha/,n) reactions, as well as branching fractions leading to the energetically possible levels of the product nuclides. A library of these data has been accumulated for target nuclides of Z /le/ 15 using that available from measurements and from recent GNASH code calculations. SOURCES, assuming neutrons to be emitted isotopically in the center-of-mass system, uses libraries of /alpha/ stopping cross sections, (/alpha/,n) reaction cross reactions, product nuclide level branching fractions, and actinide decay /alpha/ spectra to calculate thick target (/alpha/,n) yields and neutron spectra for homogeneous combinations of nuclides. The code also calculates the thick target yield and angle intergrated neutron spectrum produced by /alpha/-particle beams on targets of homogeneous mixtures of nuclides. Illustrative calculated results are given and comparisons are made with measured thick target yields and spectra. 50 refs., 1 fig., 2 tabs

  7. Neutron spectra and H*(10) around and 18 MV Linac by Ann's

    International Nuclear Information System (INIS)

    Banuelos F, A.; Valero L, C.; Borja H, C. G.; Hernandez D, V. M.; Vega C, H. R.

    2011-10-01

    Neutron spectra and ambient dose equivalent H*(10) were calculated for a radiotherapy room in 16 point-like detectors, 15 located inside the vault room and 1 located outside the bunker. The calculation was carried out using Monte Carlo Methods with the MCNP5 code for a generic radiotherapy room model operating with a 18 MV Linac, obtaining 16 neutron spectra with 47 energy bins, the H*(10) values were calculated from the neutron spectra by the use of the fluence-dose conversion factors. An artificial neural network were designed and trained to determine the neutron H*(10) in 15 different locations inside the vault room from the H*(10) dose calculated for the detector located outside the room, using the calculated dose values as training set, using the scaled conjugated gradient training algorithm. The mean squared error set for the network training was 1E(-14), adjusting the data in 99.992 %. In the treatment hall, as the distance respect to the isocenter is increased, the amount of neutrons and the H*(10) are reduced, neutrons in the high-energy region are shifted to lower region peaking around 0.1 MeV, however the epithermal and thermal neutrons remain constant due to the room-return effect. In the maze the spectra are dominated by epithermal and thermal neutrons that contributes to produce activation and the production of prompt gamma-rays. The results shows the using this artificial intelligence technic as a useful tool for the neutron spectrometry and dosimetry by the simplification on the neutronic fields characterization inside radiotherapy rooms avoiding the use of traditional spectrometric systems. And once the H*(10) doses have been calculated, to take the appropriated actions to reduce or prevent the patient and working staff exposure to this undesirable neutron radiation. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  9. The thick-target 9Be(d,n) neutron spectra for deuteron energies between 2.6 and 7.0-MeV

    International Nuclear Information System (INIS)

    Meadows, J.W.

    1991-11-01

    The measurement of the zero deg. neutron spectra and yields from deuterons incident on thick beryllium metal targets is described. 235 U and 238 U fission ion chambers were used as neutron detectors to span the neutron energy range above 0.05-MeV with a time resolution of ≤ 3 nanosec. Measurements were made for incident deuteron energies from 2.6 to 7.0-MeV, at 0.4-MeV intervals, using time-of-flight techniques with flight paths of 2.7 and 6.8 meters. The results are presented in graphical form and in tables

  10. Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

    International Nuclear Information System (INIS)

    Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

    1974-01-01

    In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

  11. Neutron delayed choice experiments

    International Nuclear Information System (INIS)

    Bernstein, H.J.

    1986-01-01

    Delayed choice experiments for neutrons can help extend the interpretation of quantum mechanical phenomena. They may also rule out alternative explanations which static interference experiments allow. A simple example of a feasible neutron test is presented and discussed. (orig.)

  12. COLLI-PTB, Neutron Fluence Spectra for 3-D Collimator System by Monte-Carlo

    International Nuclear Information System (INIS)

    Schlegel-Bickmann, Dietrich

    1995-01-01

    1 - Description of program or function: For optimizing collimator systems (shieldings) for fast neutrons with energies between 10 KeV and 20 MeV. Only elastic and inelastic neutron scattering processes are involved. Isotropic angular distribution for inelastic scattering in the center of mass system is assumed. 2 - Method of solution: The Monte Carlo method with importance sampling technique, splitting and Russian Roulette is used. The neutron attenuation and scattering kinematics is taken into account. 3 - Restrictions on the complexity of the problem: Energy range from 10 KeV to 20 MeV. For the output spectra any bin width is possible. The output spectra are confined to 40 equidistant channels

  13. Recent activities for β-decay half-lives and β-delayed neutron emission of very neutron-rich isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Dillmann, Iris [TRIUMF, Vancouver BC, V6T 2A3, Canada and GSI Helmholtzzentrum für Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Abriola, Daniel [Laboratorio Tandar, Comisión Nacional de Energía Atómica, B1650KINA, San Martín, Buenos Aires (Argentina); Singh, Balraj [Department of Physics and Astronomy, McMaster University, Hamilton ON, L8S 4M1 (Canada)

    2014-05-02

    Beta-delayed neutron (βn) emitters play an important, two-fold role in the stellar nucleosynthesis of heavy elements in the 'rapid neutron-capture process' (r process). On one hand they lead to a detour of the material β-decaying back to stability. On the other hand, the released neutrons increase the neutron-to-seed ratio, and are re-captured during the freeze-out phase and thus influence the final solar r-abundance curve. A large fraction of the isotopes inside the r-process reaction path are not yet experimentally accessible and are located in the (experimental) 'Terra Incognita'. With the next generation of fragmentation and ISOL facilities presently being built or already in operation, one of the main motivation of all projects is the investigation of these very neutron-rich isotopes. A short overview of one of the planned programs to measure βn-emitters at the limits of the presently know isotopes, the BRIKEN campaign (Beta delayed neutron emission measurements at RIKEN) will be given. Presently, about 600 β-delayed one-neutron emitters are accessible, but only for a third of them experimental data are available. Reaching more neutron-rich isotopes means also that multiple neutron-emission becomes the dominant decay mechanism. About 460 β-delayed two-, three-or four-neutron emitters are identified up to now but for only 30 of them experimental data about the neutron branching ratios are available, most of them in the light mass region below A=30. The International Atomic and Energy Agency (IAEA) has identified the urgency and picked up this topic recently in a 'Coordinated Research Project' on a 'Reference Database for Beta-Delayed Neutron Emission Data'. This project will review, compile, and evaluate the existing data for neutron-branching ratios and half-lives of β-delayed neutron emitters and help to ensure a reliable database for the future discoveries of new isotopes and help to constrain astrophysical and

  14. Neutron spectra of /sup 239/Pu-Be neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A; Nagarajan, P S [Bhabha Atomic Research Centre, Bombay (India). Div. of Radiation Protection

    1977-01-01

    Neutron spectra of /sup 239/Pu-Be(..cap alpha..,n) sources have been calculated by using the most recent data on the differential cross sections and angular distributions. The contribution from the multibody break-up reaction /sup 9/Be(..cap alpha..,..cap alpha..n)/sup 8/Be has also been incorporated. Modifications to the primary spectrum due to the secondary interactions in the source such as elastic scattering with beryllium, oxygen and plutonium and the /sup 9/Be(n,2n) and /sup 239/Pu(n,f) reaction have been calculated for different strengths and geometries. The present calculation has shown that the spectrum changes considerably because of these events within the source by way of smearing of peaks and filling up of valleys and raising the low energy part of the spectrum. Increase in H/D value leads to channeling of extra neutrons into the equatorial plane at the cost of the neutrons along the axial direction. The present calculations show that inclusion of secondary interactions to the extent considered in this work does not account completely for the increased intensity in the lower energy end of the measured spectrum.

  15. Determination of the fast neutrons spectra by the Elastic scattering method (n, p); Determinacion del espectro de neutrones rapidos por el metodo de la dispersion elastica (n, p)

    Energy Technology Data Exchange (ETDEWEB)

    Elizalde D, J

    1973-07-01

    This work consists in determining the fast neutron spectra emitted by a Pu-Be isotopic source. The implemented technique is based in the spectrometry (n, p). This consists in making to fall on a fast neutrons beams (polyenergetic) over a thin film of hydrogenated material, detecting the spectra of emitted protons at a fix angle. The polyethylene film and the used solid state detector are inside of a vacuum chamber. The detector is placed at 30 degree with respect to direction of the incident neutrons beam. The protons spectra is stored in a multichannel. the energy is obtained with the prior calibration of the system. The data processing involves the transformation of the protons spectra observed at the falling on neutrons spectra over the film. The energy of the neutrons is related with that of the protons, according to the collision kinematical equations. The cross section of elastic collision of the neutrons with the hydrogen atoms is obtained from literature. Applying these relations to the observed spectra it is obtained the falling on neutron spectra over the film. (Author)

  16. Measurement of neutron energy spectra for Eg=23.1 and 26.6 MeV mono-energetic photon induced reaction on natC using laser electron photon beam at NewSUBARU

    Science.gov (United States)

    Itoga, Toshiro; Nakashima, Hiroshi; Sanami, Toshiya; Namito, Yoshihito; Kirihara, Yoichi; Miyamoto, Shuji; Takemoto, Akinori; Yamaguchi, Masashi; Asano, Yoshihiro

    2017-09-01

    Photo-neutron energy spectra for Eg=23.1 and 26.6 MeV mono-energetic photons on natC were measured using laser Compton scattering facility at NewSUBARU BL01. The photon energy spectra were evaluated through measurements and simulations with collimator sizes and arrangements for the laser electron photon. The neutron energy spectra for the natC(g,xn) reaction were measured at 60 degrees in horizontal and 90 degrees in horizontal and vertical with respect to incident photon. The spectra show almost isotropic angular distribution and flat energy distribution from detection threshold to upper limit defined by reaction Q-value.

  17. Microdosimetry of intermediate energy neutrons in fast neutron fields

    International Nuclear Information System (INIS)

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

    1988-01-01

    A coaxial double cylindrical proportional counter has been constructed for microdosimetry of intermediate energy neutrons in mixed fields. Details are given of the measured gas gain and resolution characteristics of the counter for a wide range of anode voltages. Event spectra due to intermediate neutrons in any desired energy band is achieved by an appropriate choice of thickness of the common dividing wall in the counter and by appropriate use of the coincidence, anticoincidence pulse counting arrangements. Calculated estimates indicate that the dose contribution by fast neutrons to the energy deposition events in the intermediate neutron range may be as large as 25%. Empirical procedures being investigated aim to determine the necessary corrections to be applied to the microdose distributions, with a precision of 10%. (author)

  18. Neutron spectra from radionuclide sources for cardiac pacemakers

    International Nuclear Information System (INIS)

    Kluge, H.

    1975-01-01

    Neutron spectra from Plutonium 238 radioisotope batteries powering cardiac pacemakers are measured in the energy range above 0.7 MeV. The results are used to calculate radiation doses within a cylindrical phantom. There are only minor differences between the different types of plutonium 238-batteries and californium 252-batteries

  19. Energy spectra of fast neutrons by nuclear emulsion method

    International Nuclear Information System (INIS)

    Quaresma, A.A.

    1977-01-01

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

  20. Systematics in delayed neutron yields

    Energy Technology Data Exchange (ETDEWEB)

    Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.

    1998-03-01

    An attempt was made to reproduce the systematic trend observed in the delayed neutron yields for actinides on the basis of the five-Gaussian representation of the fission yield together with available data sets for delayed neutron emission probability. It was found that systematic decrease in DNY for heavier actinides is mainly due to decrease of fission yields of precursors in the lighter side of the light fragment region. (author)

  1. Measurement of neutron energy spectra for Eg=23.1 and 26.6 MeV mono-energetic photon induced reaction on natC using laser electron photon beam at NewSUBARU

    Directory of Open Access Journals (Sweden)

    Itoga Toshiro

    2017-01-01

    Full Text Available Photo-neutron energy spectra for Eg=23.1 and 26.6 MeV mono-energetic photons on natC were measured using laser Compton scattering facility at NewSUBARU BL01. The photon energy spectra were evaluated through measurements and simulations with collimator sizes and arrangements for the laser electron photon. The neutron energy spectra for the natC(g,xn reaction were measured at 60 degrees in horizontal and 90 degrees in horizontal and vertical with respect to incident photon. The spectra show almost isotropic angular distribution and flat energy distribution from detection threshold to upper limit defined by reaction Q-value.

  2. Unfolding neutron spectra from simulated response of thermoluminescence dosimeters inside a polyethylene sphere using GRNN neural network

    Science.gov (United States)

    Lotfalizadeh, F.; Faghihi, R.; Bahadorzadeh, B.; Sina, S.

    2017-07-01

    Neutron spectrometry using a single-sphere containing dosimeters has been developed recently, as an effective replacement for Bonner sphere spectrometry. The aim of this study is unfolding the neutron energy spectra using GRNN artificial neural network, from the response of thermoluminescence dosimeters, TLDs, located inside a polyethylene sphere. The spectrometer was simulated using MCNP5. TLD-600 and TLD-700 dosimeters were simulated at different positions in all directions. Then the GRNN was used for neutron spectra prediction, using the TLDs' readings. Comparison of spectra predicted by the network with the real spectra, show that the single-sphere dosimeter is an effective instrument in unfolding neutron spectra.

  3. Unfolding neutron spectra with BS-TLD system using genetic algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Joelan A.L., E-mail: jasantos@cnen.gov.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Silva, Everton R. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Informatica; Ferreira, Tiago A.E. [Universidade Federal Rural de Pernambuco (UFRPE), Recife, PE (Brazil). Dept. de Estatistica e Informatica; Fonseca, Evaldo S. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Vilela, Eudice C., E-mail: ecvilela@cnen.gov.br [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2011-07-01

    Due to the variability of neutron spectrum within the same environment, it is essential that the spectral distribution as function of energy to be characterized. To perform this task, the neutron spectrometer has a primary role in determining the neutron flux ({Phi}{sub E}(E)). Precise information allows radiological quantities establishment related to that spectrum but it is necessary, however, a series of steps with a spectrometric system that can cover a large interval of energy and whose answer is isotropic. The most widely used for accomplishing this task is the spectrometric Bonner spheres system. One of the biggest problems related to neutron spectrometry is the process of data analysis, known as unfolding. Most of the work undertaken to implement new techniques of this process, using data obtained with the scintillator {sup 6}LiI(I). However, characteristics related to the dead time make it not be so effective when used in high flow neutron fields. An alternative to this problem is the use of thermoluminescent detectors (TLD), but the codes used do not provide a more specific response matrix to unfolding the information obtained through these materials, which makes the development of a specific response matrix important to adequately characterize the response obtained by them. This paper proposes using a technique of artificial intelligence called genetic algorithm, which uses bio-inspired mathematical models and through the implementation of a specific matrix to unfolding data obtained from a combination of TLDs embedded in a system of Bonner spheres, such as thermal neutron detectors, to characterize the neutron spectrum as a function of energy. The results obtained with this method were in accordance with reference spectra, thus enables of this technique to unfolding neutrons spectra with BS-TLD system. (author)

  4. Unfolding neutron spectra with BS-TLD system using genetic algorithms

    International Nuclear Information System (INIS)

    Santos, Joelan A.L.; Silva, Everton R.; Vilela, Eudice C.

    2011-01-01

    Due to the variability of neutron spectrum within the same environment, it is essential that the spectral distribution as function of energy to be characterized. To perform this task, the neutron spectrometer has a primary role in determining the neutron flux (Φ E (E)). Precise information allows radiological quantities establishment related to that spectrum but it is necessary, however, a series of steps with a spectrometric system that can cover a large interval of energy and whose answer is isotropic. The most widely used for accomplishing this task is the spectrometric Bonner spheres system. One of the biggest problems related to neutron spectrometry is the process of data analysis, known as unfolding. Most of the work undertaken to implement new techniques of this process, using data obtained with the scintillator 6 LiI(I). However, characteristics related to the dead time make it not be so effective when used in high flow neutron fields. An alternative to this problem is the use of thermoluminescent detectors (TLD), but the codes used do not provide a more specific response matrix to unfolding the information obtained through these materials, which makes the development of a specific response matrix important to adequately characterize the response obtained by them. This paper proposes using a technique of artificial intelligence called genetic algorithm, which uses bio-inspired mathematical models and through the implementation of a specific matrix to unfolding data obtained from a combination of TLDs embedded in a system of Bonner spheres, such as thermal neutron detectors, to characterize the neutron spectrum as a function of energy. The results obtained with this method were in accordance with reference spectra, thus enables of this technique to unfolding neutrons spectra with BS-TLD system. (author)

  5. {sup 235}U(n,F) prompt fission neutron spectra

    Energy Technology Data Exchange (ETDEWEB)

    Maslov, M.V.; Tetereva, N.A. [Joint Institute of Nuclear and Energy Research, Minsk-Sosny (Belarus); Pronyaev, V.P.; Kagalenko, A.B. [Institute of Physics and Power Engineering, Obninsk (Russian Federation); Capote, R. [International Atomic Energy Agency, Vienna (Austria); Granier, T.; Morillon, B. [CEA, Centre DAM-IIe de France, 91 - Arpajon (France); Hambsch, F.J. [EC-JRC Institute for Reference Materials and Measurements, Geel (Belgium); Sublet, J.C. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2009-07-01

    The longstanding problem of inconsistency of integral thermal data testing and differential prompt fission neutron spectra data (PFNS) is mostly due to rather poor fits of differential PFNS data in major data libraries. The measured database is updated by using modern standards including Manhart's evaluation of the spontaneous fission neutron spectra of {sup 252}Cf(sf). That largely removes the inconsistency of older thermal neutron-induced PFNS measurements with newest data of JRC IRMM by Hambsch et al. (2009). A phenomenological approach, developed by Kornilov et al. (1999), for the first-chance fission and extended for the emissive fission domain by Maslov et al. (2005) is calibrated at E{sub th} to predict both the PFNS average energy and PFNS shape up to 20 MeV. The latter is extremely important, since rather close values in fact correspond to quite discrepant spectra shapes, which influences reactor neutronics strongly. The proposed phenomenological representation of the PFNS reproduces both soft and hard energy tails of {sup 235}U(n{sub th},F) PFNS at thermal incident neutron energy E{sub th}. In the first-chance and emissive fission domain evaluated PFNS are consistent with the data by Ethvignot et al. (2005). A compiled MF=5 Endf/B-formatted file of the {sup 235}U(n,F) PFNS largely removes the inconsistencies of the evaluated differential PFNS with integral data benchmarks. Almost perfect fits are attained for available differential PFNS data from E{sub th} up to E{sub n}=14.7 MeV, with few exceptions at E{sub n}=2.9 and E{sub n}=5 MeV. Fast integral critical experiment like GODIVA or Flattop benchmarks might be reproduced almost with the same accuracy as with the PFNS of the major data libraries. That reveals a rather delicate compensation effect, since present and previous PFNS shapes are drastically different from each other. Thermal assemblies benchmarking reveals positive biases in k(eff), which might be attributed to the influence of

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

    CERN Document Server

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

    2010-01-01

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

  7. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for PKA energy spectra and heating number under neutron irradiation

    International Nuclear Information System (INIS)

    Iwamoto, Y.; Ogawa, T.

    2016-01-01

    The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials

  8. A new method to evaluate neutron spectra for bnct

    International Nuclear Information System (INIS)

    Martin Hernandez, Guido

    2001-01-01

    This paper deals with the development of a method to evaluate neutron spectra for BNCT. Physical dose deposition calculations for different neutron energies, ranging from thermal to fast, were performed. A matrix, containing dose for each energy and position in the beam center line was obtained. MCNP 4B and Snyder's head model were used. A simple computer code containing the matrix calculates the dose for each point in the beam center line depending on the input energy spectrum to be evaluated. The output of this program is the dose distribution in the brain and the dose gain, that is the ratio between dose to tumor and maximum dose to healthy tissue maximum

  9. Neutron Thermalization and Reactor Spectra. Vol. II. Proceedings of the Symposium on Neutron Thermalization and Reactor Spectra

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held at Ann Arbor, Michigan, USA, 17 - 21 July 1967. The meeting was attended by 143 participants from 24 Member States and one international organization. Contents: (Vol.I) Theory of neutron thermalization (15 papers); Scattering law (20 papers); Angular, space, temperature and time dependence of neutron spectra (9 papers). (Vol.II) Measurement of thermal neutron spectra and spectral indices, and comparison with theory (17 papers); Time-dependent problems in neutron thermalization (12 papers). Each paper is in its original language (61 English, 1 French and 11 Russian) and is preceded by an abstract in English with one in the original language if this is not English. Discussions are in English.

  10. Quality factor calculations for neutron spectra below 4 MeV

    International Nuclear Information System (INIS)

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

    1979-01-01

    A method is described for computing the distribution of absorbed dose, D(L), as a function of linear energy transfer, L, for any neutron spectrum with energies below 4 MeV. The results are used to determine the average quality factor for two distinctly different neutron spectra using the ICRP recommended values of the quality factor, Q(L). A comparison is made between the calculations and measurements of D(L) using a spherical tissue equivalent proportional counter. Heavy ion recoil contributions to the average quality factor are examined in detail. (author)

  11. Computed secondary-particle energy spectra following nonelastic neutron interactions with 12C for En between 15 and 60 MeV: Comparisons of results from two calculational methods

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1991-04-01

    The organic scintillation detector response code SCINFUL has been used to compute secondary-particle energy spectra, dσ/dE, following nonelastic neutron interactions with 12 C for incident neutron energies between 15 and 60 MeV. The resulting spectra are compared with published similar spectra computed by Brenner and Prael who used an intranuclear cascade code, including alpha clustering, a particle pickup mechanism, and a theoretical approach to sequential decay via intermediate particle-unstable states. The similarities of and the differences between the results of the two approaches are discussed. 16 refs., 44 figs., 2 tabs

  12. Measurement of neutron energy spectra of PuO[sub 2]-UO[sub 2] mixed oxide fuel and penetrated through surrounding lead-acryl shield

    Energy Technology Data Exchange (ETDEWEB)

    Nakao, Noriaki; Tsujimura, Norio; Nakamura, Takashi (Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center); Momose, Takumaro; Ninomiya, Kazushige; Ishiguro; Hideharu

    1993-12-01

    The energy spectra of neutrons emitted from an aluminum can containing PuO[sub 2]-UO[sub 2] mixed oxide fuel and penetrated through a 35mm thick lead-acryl shield surrounding the can, were measured with the NE-213 organic liquid scintillator, the proton recoil proportional counter and the multi-moderator [sup 3]He spectrometer (Bonner Ball). The measured results were compared with the results calculated by the MORSE-CG Monte Carlo code on the basis of source neutron yields obtained by the ORIGEN-2 code and the source energy spectrum cited from the reference data. The agreement between these two was pretty good. The dose equivalents were then calculated from thus-obtained energy spectra and the flux-to-dose conversion factor and showed good agreement with the data measured with the neutron dose-equivalent counters (rem counters). Since the published data on energy spectrum of mixed oxide fuel are very scarce, these results can be useful as basic data for shielding design study and radiation control of nuclear fuel facilities. (author).

  13. The application of n-γ discrimination in 252Cf spontaneous neutron TOF spectra measurement

    International Nuclear Information System (INIS)

    Zhou Haojun; Zhang Yi; Li Jiansheng; Jin Yu; Wang Jie; Li Chunyuan

    2004-01-01

    The BC501 scintillator is used as a fast neutron detector. The effect that the pulse rise time method was used to discriminate γ from 252 Cf spontaneous neutron TOF spectra is studied in the experiment. A pulse rise time separation spectra of γ and 252 Cf spontaneous neuron upon 1 MeV is obtained, the n-γ separation function reaches to 4.6. When the result of pulse rise time separation coincides with the time-of-flight spectra in which the neutron energy is upon 0.5 MeV, 0.8 MeV and 1.0 MeV, comparing with the anticoincidence, γ was eliminated 99.90% at least. (authors)

  14. Measurement and Analysis of the Neutron and Gamma-Ray Flux Spectra in a Neutronics Mock-Up of the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Seidel, K.; Freiesleben, H.; Poenitz, E.; Klix, A.; Unholzer, S.; Batistoni, P.; Fischer, U.; Leichtle, D.

    2006-01-01

    The nuclear parameters of a breeding blanket, such as tritium production rate, nuclear heating, activation and dose rate, are calculated by integral folding of an energy dependent cross section (or coefficient) with the neutron (or gamma-ray) flux energy spectra. The uncertainties of the designed parameters are determined by the uncertainties of both the cross section data and the flux spectra obtained by transport calculations. Also the analysis of possible discrepancies between measured and calculated integral nuclear parameter represents a two-step procedure. First, the energy region and the amount of flux discrepancies has to be found out and second, the cross section data have to be checked. To this end, neutron and gamma-ray flux spectra in a mock-up of the EU Helium-Cooled Pebble Bed (HCPB) breeder Test Blanket Module (TBM), irradiated with 14 MeV neutrons, were measured and analysed by means of Monte Carlo transport calculations. The flux spectra were determined for the energy ranges that are relevant for the most important nuclear parameters of the TBM, which are the tritium production rate and the shielding capability. The fast neutron flux which determines the tritium production on 7 Li and dominates the shield design was measured by the pulse-height distribution obtained from an organic liquid scintillation detector. Simultaneously, the gamma-ray flux spectra were measured. The neutron flux at lower energies, down to thermal, which determines the tritium production on 6 Li, was measured with time-of-arrival spectroscopy. For this purpose, the TUD neutron generator was operated in pulsed mode (pulse width 10 μs, frequency 1 kHz) and the neutrons arriving at a 3 He proportional counter in the mock-up were recorded as a function of time after the source neutron pulse. The spectral distributions for the two positions in the mock-up, where measurements were carried out, were calculated with the Monte Carlo code MCNP, version 5, and nuclear data from the

  15. Neutron spectra and H*(10) around and 18 MV Linac by Ann's

    Energy Technology Data Exchange (ETDEWEB)

    Banuelos F, A.; Valero L, C.; Borja H, C. G.; Hernandez D, V. M.; Vega C, H. R., E-mail: alanb535@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2011-10-15

    Neutron spectra and ambient dose equivalent H*(10) were calculated for a radiotherapy room in 16 point-like detectors, 15 located inside the vault room and 1 located outside the bunker. The calculation was carried out using Monte Carlo Methods with the MCNP5 code for a generic radiotherapy room model operating with a 18 MV Linac, obtaining 16 neutron spectra with 47 energy bins, the H*(10) values were calculated from the neutron spectra by the use of the fluence-dose conversion factors. An artificial neural network were designed and trained to determine the neutron H*(10) in 15 different locations inside the vault room from the H*(10) dose calculated for the detector located outside the room, using the calculated dose values as training set, using the scaled conjugated gradient training algorithm. The mean squared error set for the network training was 1E(-14), adjusting the data in 99.992 %. In the treatment hall, as the distance respect to the isocenter is increased, the amount of neutrons and the H*(10) are reduced, neutrons in the high-energy region are shifted to lower region peaking around 0.1 MeV, however the epithermal and thermal neutrons remain constant due to the room-return effect. In the maze the spectra are dominated by epithermal and thermal neutrons that contributes to produce activation and the production of prompt gamma-rays. The results shows the using this artificial intelligence technic as a useful tool for the neutron spectrometry and dosimetry by the simplification on the neutronic fields characterization inside radiotherapy rooms avoiding the use of traditional spectrometric systems. And once the H*(10) doses have been calculated, to take the appropriated actions to reduce or prevent the patient and working staff exposure to this undesirable neutron radiation. (Author)

  16. Neutron and gamma-ray spectra measurement on the model of the KS-150 reactor radial shielding

    International Nuclear Information System (INIS)

    Holman, M.; Hogel, J.; Marik, J.; Kovarik, K.; Franc, L.; Vespalec, R.

    1977-01-01

    A shortened model of the peripheral region of the KS-150 reactor core consisting of two rows of fuel elements and a reflector was constructed from the peripheral fuel elements of the KS-150 reactor core in an experiment on the TR-0 reactor. The mockup of the thermal shield (10 cm of steel), the pressure vessel (15 cm of steel) and the inner wall of the water biological shielding (2 cm of steel) of the KS-150 reactor were erected outside the TR-0 vessel. Fast neutron and gamma spectra were measured with a stilbene crystal scintillation spectrometer. The resonance neutron spectra were measured with 197 Au, 63 Cu and 23 Na resonance activation detectors. Fast neutron spectra inside the reactor were measured with a 10 mm diameter by 10 mm thick stilbene crystal spectrometer, outside the reactor with a 10 mm diameter by 10 mm thick and a 20 mm diameter by 20 mm thick stilbene crystal spectrometer. Neutron spectra in the energy regions of 1 eV to 3 keV and 0.6 MeV to 0.8 MeV were obtained on the core periphery, on the reflector half-thickness and in front of and behind the reactor thermal shield. Gamma spectra were obtained in front of and behind the thermal shield. It was found that the attenuation of neutron fluxes by the reflector and the thermal shield increased with increasing energy while gamma radiation attenuation decreased with increasing energy. It was not possible to obtain the neutron spectrum in the 10 to 600 keV energy range because suitable detection instrumentation was not available. (J.P.)

  17. Time-of-flight neutron spectra measurements in Zenith

    Energy Technology Data Exchange (ETDEWEB)

    Barclay, F R; Coates, M S; Diment, K M; Durrani, S A; Gayther, D B; Poole, M J; Reed, D L

    1962-01-15

    Neutron spectra in the second core loading of ZENITH have been measured using a neutron chopper. Spectra at two positions in the reactore core were obtained over a range of temperatures extending to 650 deg C.

  18. An experimental facility for studying delayed neutron emission

    International Nuclear Information System (INIS)

    Dermendzhiev, E.; Nazarov, V.M.; Pavlov, S.S.; Ruskov, Iv.; Zamyatin, Yu.S.

    1993-01-01

    A new experimental facility for studying delayed neutron emission has been designed and tested. A method based on utilization of the Dubna IBR-2 pulsed reactor, has been proposed and realized for periodical irradiation of targets composed of fissionable isotopes. Such a powerful pulsed neutron source in combination with a slow neutron chopper synchronized with the reactor bursts makes possible variation of the exposure duration and effective suppression of the fast neutron background due to delay neutrons emitted from the reactor core. Detection of delayed neutrons from the target is carried out by a high-efficiency multicounter neutron detector with a near-4π geometry. Some test measurements and results are briefly described. Possible use of the facility for other tasks is also discussed. 14 refs.; 14 figs

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

    International Nuclear Information System (INIS)

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

    1987-02-01

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

  20. Theoretical and Experimental Analysis of Fast Neutron Spectra

    Energy Technology Data Exchange (ETDEWEB)

    Van Dam, H.; Kleijn, H. R. [Reactor Instituut, Delft (Netherlands)

    1968-04-15

    The reactor physics division of the Inter-Academic Reactor Institute at Delft is concentrating its efforts in the field of fast reactor physics on problems of a more fundamental nature. The object of the programme is to determine experimentally a number of microscopic reactor physics parameters such as conversion potentials, fission ratios and Doppler coefficients for simple geometries and material compositions. Because of the extreme importance of knowledge of the neutron spectrum for the interpretation of the results, attention has initially been concentrated on both the measurement and the calculation of fast neutron spectra. The transport of neutrons in absorbing and non-absorbing heavy atom materials is studied by solving the Boltzmann equation. Both isotropic and anisotropic scattering are considered. Anisotropic scattering is treated by the P{sub n}-approximation, while flux-anisotropy is handled with the S{sub N}-method. In the code FAST-DELFT, scattering is treated up to the P{sub 4} component, a further extension of which is useless because of the lack of available cross-section data. By using this method, the effect of scattering anisotropy on the spectrum formation has been studied. In addition the influence of group cross-section inaccuracies was determined. The experimental work has been concentrated on methods to determine in-core spectra. Using home-made proportional counters with gamma-ray discrimination provisions fast neutron spectra have been measured in simple geometries. These experiments were complemented by foil measurements in the lower energy region. The results of this work are presented in this paper. (author)

  1. The LANL/LLNL Program to Measure Prompt Fission Neutron Spectra at LANSCE

    Science.gov (United States)

    Haight, Robert; Wu, Ching Yen; Lee, Hye Young; Taddeucci, Terry; Mosby, Shea; O'Donnell, John; Fotiades, Nikolaos; Devlin, Mattew; Ullmann, John; Nelson, Ronald; Wender, Stephen; White, Morgan; Solomon, Clell; Neudecker, Denise; Talou, Patrick; Rising, Michael; Bucher, Brian; Buckner, Matthew; Henderson, Roger

    2015-10-01

    Accurate data on the spectrum of neutrons emitted in neutron-induced fission are needed for applications and for a better understanding of the fission process. At LANSCE we have made important progress in understanding systematic uncertainties and in obtaining data for 235U on the low-energy part of the prompt fission neutron spectra (PFNS), a particularly difficult region because down-scattered neutrons go in this direction. We use a double time-of-flight technique to determine energies of incoming and outgoing neutrons. With data acquisition via waveform digitizers, accidental coincidences between fission chamber and neutron detector are measured to high statistical accuracy and then subtracted from measured events. Monte Carlo simulations with high performance computers have proven to be essential in the design to minimize neutron scattering and in calculating detector response. Results from one of three approaches to analyzing the data will be presented. This work is funded by the US Department of Energy, National Nuclear Security Administration and Office of Nuclear Physics.

  2. Radiochemical Means of Investigating Delayed Neutron Precursors

    International Nuclear Information System (INIS)

    Marmol, P. del

    1968-01-01

    Fast radiochemical methods used now for the determination of delayed neutron precursors are classified and reviewed: precipitations, solvent extractions, range experiments, milking, gas sweeping, isotopic and ion exchange, hot atom reactions and diffusion loss. Advantages and limitations of irradiation systems with respect to fast separations are discussed: external beams which allow faster separations only have low neutron fluxes, internal beams which are mostly fit for gaseous reactions; and rabbits for solution irradiations. Future prospects of radiochemical procedures are presented; among these, studies should be mostly oriented towards gaseous reactions which offer possibilities of isolating very short-lived delayed neutron precursors. Chemical procedures for delayed neutron precursor detection are compared with mass spectrometric and isotope separator techniques; it is concluded that the methods are complementary. (author)

  3. Radiochemical Means of Investigating Delayed Neutron Precursors

    International Nuclear Information System (INIS)

    Marmol, P. del

    1968-01-01

    Fast radiochemical methods used now for the determination of delayed neutron precursors are classified and reviewed: precipitations, solvent extractions, range experiments, milking, gas sweeping, isotopic and ion exchange, hot-atom reactions and diffusion loss. Advantages and limitations of irradiation systems with respect to fast separations are discussed: external beams which allow faster separations only have low neutron fluxes, internal beams which are mostly fit for gaseous reactions; and rabbits for solution irradiations. Future prospects of radiochemical procedures are presented; among these, studies should be mostly oriented towards gaseous reactions which offer possibilities of isolating very short-lived delayed neutron precursors. Chemical procedures for delayed neutron precursor detection are compared with mass spectrometric and isotope-separator techniques; it is concluded that the methods are complementary. (author)

  4. Study of calculated and measured time dependent delayed neutron yields

    International Nuclear Information System (INIS)

    Waldo, R.W.

    1980-05-01

    Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of 232 U, 237 Np, 238 Pu, 241 Am, /sup 242m/Am, 245 Cm, and 249 Cf were studied for the first time. The delayed neutron emission from 232 Th, 233 U, 235 U, 238 U, 239 Pu, 241 Pu, and 242 Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from 232 Th to 252 Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables

  5. Measurement and model description of differential neutron spectra of the californium 252 spontaneous fission depending on THETA, Msub(T), Esub(kin sum)

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.A.; Sidorov, L.V.; Vasil'eva, N.K.; Barashkov, Yu.A.; Golovanov, O.A.; Kopalkin, N.V.; Nemudrov, N.I.; Surin, V.M.; Khachaturov, Yu.F.

    1984-01-01

    The results of the 4π-spectrometer mesurement of the neutron spectra in the 26-154 deg angle range for seven groups of fragments with different masses and total kinetic energies are given. Experimental spectra have been analyzed for consistency with the evaporation model of neutrons from moving fragments. The results of an analysis of differential neutron spectra shows that the main reason of the ''yearly'' neutron emission is a neutron evaporation from fragments with large excitation energy and from fragments with neutron number N>82 during the time as compared with the time of fragment acceleration

  6. Neutron leakage spectra from Be, Pb and U spheres at 14 MeV energy

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Devkin, B.V.

    1989-01-01

    Experimental data on neutron leakage spectra from beryllium, lead and uranium spheres with a central 14 MeV neutron source using a time-of-flight spectrometer have been measured. The data were compared with those calculated with the BLANK code using different nuclear data files. 15 refs, 1 fig., 2 tabs

  7. Measurement and Analysis of Neutron Leakage Spectra from Pb and LBE Cylinders with D-T Neutrons

    Science.gov (United States)

    Chen, Size; Gan, Leting; Li, Taosheng; Han, Yuncheng; Liu, Chao; Jiang, Jieqiong; Wu, Yican

    2017-09-01

    For validating the current evaluated neutron data libraries, neutron leakage spectra from lead and lead bismuth eutectic (LBE) cylinders have been measured using an intense D-T pulsed neutron source with time-of-flight (TOF) method by Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS). The measured leakage spectra have been compared with the calculated ones using Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC) with the evaluated pointwise data of lead and bismuth processed from ENDF/B-VII.1, JEFF-3.1 and JENDL-4.0 libraries. This work shows that calculations of the three libraries are all generally consistent with the lead experimental result. For LBE experiment, the JEFF-3.1 and JENDL-4.0 calculations both agree well with the measurement. However, the result of ENDF/B-VII.1 fails to fit with the measured data, especially in the energy range of 5.5 and 7 MeV with difference more than 80%. Through sensitivity analysis with partial cross sections of 209Bi in ENDF/B-VII.1 and JEFF, the difference between the measurement and the ENDF/B-VII.1 calculation in LBE experiment is found due to the neutron data of 209Bi.

  8. Unfolding neutron spectra obtained from BS–TLD system using genetic algorithm

    International Nuclear Information System (INIS)

    Santos, J.A.L.; Silva, E.R.; Ferreira, T.A.E; Vilela, E.C.

    2012-01-01

    Due to the variability of neutron spectrum within the same environment, it is essential that the spectral distribution as a function of energy should be characterized. The precise information allows radiological quantities establishment related to that spectrum, but it is necessary that a spectrometric system covers a large interval of energy and an unfolding process is appropriate. This paper proposes use of a technique of Artificial Intelligence (AI) called genetic algorithm (GA), which uses bio-inspired mathematical models with the implementation of a specific matrix to unfolding data obtained from a combination of TLDs embedded in a BS system to characterize the neutron spectrum as a function of energy. The results obtained with this method were in accordance with reference spectra, thus enabling this technique to unfold neutron spectra with the BS–TLD system. - Highlights: ► The unfolding code used the artificial intelligence technique called genetic algorithms. ► A response matrix specific to the unfolding data obtained with the BS–TLD system is used by the AGLN. ► The observed results demonstrate the potential use of genetic algorithms in solving complex nuclear problems.

  9. Study and building of a detection array for delayed neutrons: TONNERRE

    International Nuclear Information System (INIS)

    Martin, Thierry

    1998-01-01

    This work has been undertaken within a French-Romanian collaboration in order to build a high efficiency detector array for delayed neutrons: barrel-shaped TONNERRE. Some neutron-rich nuclei decay through 1, 2 or 3 neutron emission after β - decay. More exotic nuclei will be produced by SPIRAL at GANIL. An array with high efficiency and good resolution is then required. Thirty two BC400 plastic scintillators (160 x 20 x 4 cm 3 ) allow us to get the time of flight neutron spectra. They are bent for uniform flight path and viewed by a photomultiplier tube at both ends. Simulations have allowed to establish scintillator size and to minimize light attenuation. Intrinsic efficiency and crosstalk have been measured with 252 Cf and compared to GEANT. 1 to 5 MeV neutrons are detected with good timing and position properties. Other counters will be built for neutrons from 300 keV to 1 MeV. Planned to run at several particle accelerators (GANIL, CERN, and others), TONNERRE is modular and many geometries are possible. (author)

  10. Reconstruction of neutron spectra through neural networks

    International Nuclear Information System (INIS)

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

    2003-01-01

    A neural network has been used to reconstruct the neutron spectra starting from the counting rates of the detectors of the Bonner sphere spectrophotometric system. A group of 56 neutron spectra was selected to calculate the counting rates that would produce in a Bonner sphere system, with these data and the spectra it was trained the neural network. To prove the performance of the net, 12 spectra were used, 6 were taken of the group used for the training, 3 were obtained of mathematical functions and those other 3 correspond to real spectra. When comparing the original spectra of those reconstructed by the net we find that our net has a poor performance when reconstructing monoenergetic spectra, this attributes it to those characteristic of the spectra used for the training of the neural network, however for the other groups of spectra the results of the net are appropriate with the prospective ones. (Author)

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

    International Nuclear Information System (INIS)

    Shaughnessy, Dawn A.

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  13. Targeted Modification of Neutron Energy Spectra for National Security Applications

    Science.gov (United States)

    Bevins, James Edward

    with the current sample doping approach and applied neutron spectral shaping to design an ETA that can create realistic synthetic fission and activation products and improve technical nuclear forensics outcomes. However, the ETA presented in this research represents more than a stand alone point design with a limited scope and application. It is proof of a concept and the product of a unique capability that has a wide range of potential applications. This research demonstrates that the concept of neutron spectral shaping can be used to engineer complex neutron spectra within the confines of physics. There are many possible applications that could benefit from the ability to generate custom energy neutron spectra that fall outside of current sources and methods. The ETA is the product of a general-purpose optimization algorithm, Gnowee, and design framework, Coeus, which enables the use of Gnowee for complex nuclear design problems. Through Gnowee and Coeus, new ETA neutronics designs can be generated in days, not months or years, with a drastic reduction in the research effort required to do so. (Abstract shortened by ProQuest.).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-11-01

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

  15. Measurement of leakage neutron spectra for Tungsten with D-T neutrons and validation of evaluated nuclear data

    International Nuclear Information System (INIS)

    Zhang, S.; Chen, Z.; Nie, Y.; Wada, R.; Ruan, X.; Han, R.; Liu, X.; Lin, W.; Liu, J.; Shi, F.; Ren, P.; Tian, G.; Luo, F.; Ren, J.; Bao, J.

    2015-01-01

    Highlights: • Evaluated data for Tungsten are validated by integral experiment. • Leakage neutron spectra from the irradiation of D-T neutrons on Tungsten are measured at 60° and 120° by using a time-of-flight method. • The measured results are compared to the MCNP-4C calculated ones with evaluated data of the different libraries. - Abstract: Integral neutronics experiments have been investigated at Institute of Modern Physics, Chinese Academy of Sciences (IMP, CAS) in order to validate evaluated nuclear data related to the design of Chinese Initiative Accelerator Driven Systems (CIADS). In the present paper, the accuracy of evaluated nuclear data for Tungsten has been examined by comparing measured leakage neutron spectra with calculated ones. Leakage neutron spectra from the irradiation of D-T neutrons on Tungsten slab sample were experimentally measured at 60° and 120° by using a time-of-flight method. Theoretical calculations are carried out by Monte Carlo neutron transport code MCNP-4C with evaluated nuclear data of the ADS-2.0, ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0 and CENDL-3.1 libraries. From the comparisons, it is found that the calculations with ADS-2.0 and ENDF/B-VII.1 give good agreements with the experiments in the whole energy regions at 60°, while a large discrepancy is observed at 120° in the elastic scattering peak, caused by a slight difference in the oscillation pattern of the elastic angular distribution at angles larger than 20°. However, the calculated spectra using data from ENDF/B-VII.0, JENDL-4.0 and CENDL-3.1 libraries showed larger discrepancies with the measured ones, especially around 8.5–13.5 MeV. Further studies are presented for these disagreements

  16. Testing of the IRDF-90 cross-section library in benchmark neutron spectra

    International Nuclear Information System (INIS)

    Nolthenius, H.J.; Zsolnay, E.M.; Szondi, E.J.

    1993-09-01

    The new version of the International Reactor Dosimetry File IRDF-90 (called ''Version April 1993'') has been tested by calculation of average cross-sections and their uncertainties in a coarse three energy group structure and by neutron spectrum adjustments in reference neutron spectra. This paper presents the results obtained and compares them with the corresponding ones of the old IRDF-85 and with the data of the Nuclear Data Guide for Reactor Neutron Metrology. The applicability of the new library in the field of neutron metrology is discussed. (orig.)

  17. The impact of ICRP 60 recommendations on the dose equivalent in low- and high energy neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Jakes, J; Schraube, H [GSF-Forschungszentrum Neuberg, D-85758 Oberschleissheim (Germany). Inst. fuer Strahlenschutz

    1996-12-31

    The objectives of this study was to determine the impact of the increased risk factors for neutrons after ICRP 60 on the operational dose equivalent quantities at a few neutron fields selected with the respect to cover the broad variety of neutron spectra: (1) Cadarache calibration assembly, with average neutron energy around 0.6 MeV, designed to simulate realistic neutron spectra at workplaces. This assembly is basically composed of an almost spherical {sup 238}U converter irradiated by 14.6 MeV neutrons from an accelerator target, placed at its center, and a scattering chamber consisting of a cylindrical polyethylene duct and a series of additional shieldings; (2) Neutron spectra at exposed workplaces in nuclear power plants; (3) Moderated spectra of {sup 252}Cf fission source; (4) Neutron spectra behind a shielding made of the iron (the average energy 5.,89 MeV) and concrete (the average energy 46.51 MeV), respectively; (5) Cosmic rays induced neutron spectra measured on the top of the Zugspitze (2968 m) where there is the average neutron energy around 40 MeV. From the derived neutron spectra, the mean quality factors and conversion factors h after ICRP 21 and ICRP 60, respectively, were calculated. The dose equivalent conversion factors were taken for the region below 20 MeV, and the energy region above 20 MeV. The results show that the operational quantities were affected predominately in the low energy fields, where the changes are given by a factor of 1,3 for the neutron fields given above. As has been expected, the impact of the new recommendations depends on the shape of the neutron spectra. Therefore, this factor can be much higher in the fields where the intermediate energy region is dominant, which is the case of moderated and scattered spectra at some places in the nuclear power plant and around containers with the spent fuel elements. (J.K.) 9 refs.

  18. International intercomparison of neutron spectra evaluating methods using activation detectors

    International Nuclear Information System (INIS)

    Fischer, A.

    1975-06-01

    The international intercomparison of neutron spectrum evaluation methods using activation detectors was organized by the IAEA in 1971 - 1972. All of the contributions and the results of a critical evaluation are presented here. The spectra of different contributors are compared to a reference spectrum by means of different integrals and weighting functions. Different cross section sets, foil numbers, energy point systems, guess spectra used by the contributors cause differences in the resulting spectra. The possible ways of separating these effects are also investigated. Suggestions are made for the organization of a new intercomparison on the basis of more uniform input data. (orig.) [de

  19. 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

  20. Recent improvements in the calculation of prompt fission neutron spectra: Preliminary results

    International Nuclear Information System (INIS)

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

    1989-01-01

    We consider three topics in the refinement and improvement of our original calculations of prompt fission neutron spectra. These are an improved calculation of the prompt fission neutron spectrum N(E) from the spontaneous fission of 252 Cf, a complete calculation of the prompt fission neutron spectrum matrix N(E,E n ) from the neutron-induced fission of 235 U, at incident neutron energies ranging from 0 to 15 MeV, and an assessment of the scission neutron component of the prompt fission neutron spectrum. Preliminary results will be presented and compared with experimental measurements and an evaluation. A suggestion is made for new integral cross section measurements. (author). 45 refs, 12 figs, 1 tab

  1. Measurement of leakage neutron spectra from silicon carbide cylinders with D–T neutrons and validation of evaluated nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Luo, F. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Han, R. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Key Laboratory of Nuclear Data, China Institute of Atomic Energy, Beijing 102413 (China); Nie, Y. [Key Laboratory of Nuclear Data, China Institute of Atomic Energy, Beijing 102413 (China); Chen, Z., E-mail: zqchen@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Zhang, S. [College of Physics Electronic Information, Inner Mongolia University for the Nationalities, Tongliao 028000 (China); Shi, F.; Lin, W.; Ren, P.; Tian, G.; Sun, Q.; Gou, B. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Ruan, X.; Ren, J. [Key Laboratory of Nuclear Data, China Institute of Atomic Energy, Beijing 102413 (China); Ye, M. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China)

    2016-11-15

    Highlights: • Evaluated data for SiC are validated by a high precision benchmark experiment. • Leakage neutron spectra from SiC cylinders are measured at 60° and 120° using time-of-flight method. • The experimental results are compared with the MCNP-4C calculations with ENDF-BVII.1, JENDL-4.0 and CENDL-3.1 libraries. • The SiC evaluated nuclear data from CENDL-3.1 library was checked for the first time and proved to be reliable. - Abstract: Benchmarking of evaluated nuclear data libraries was performed for 14 MeV neutrons on silicon carbide samples. The experiments were carried out by using the benchmark experimental facility at China Institute of Atomic Energy (CIAE). The leakage neutron spectra from SiC (Φ13 cm × 20 cm) at 60° and 120° and SiC (Φ13 cm × 2 cm) at 60° were measured by the TOF method. The measured spectra are well reproduced by MCNP-4C calculations with the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 evaluated nuclear data libraries, except 5–8 MeV range for 20 cm thickness. The discrepancies are mostly considered as caused by the improper evaluation of the angular distribution and secondary neutron energy distribution of the elastic scattering and inelastic scattering in evaluated nuclear data libraries.

  2. Measurement of leakage neutron spectra from silicon carbide cylinders with D–T neutrons and validation of evaluated nuclear data

    International Nuclear Information System (INIS)

    Luo, F.; Han, R.; Nie, Y.; Chen, Z.; Zhang, S.; Shi, F.; Lin, W.; Ren, P.; Tian, G.; Sun, Q.; Gou, B.; Ruan, X.; Ren, J.; Ye, M.

    2016-01-01

    Highlights: • Evaluated data for SiC are validated by a high precision benchmark experiment. • Leakage neutron spectra from SiC cylinders are measured at 60° and 120° using time-of-flight method. • The experimental results are compared with the MCNP-4C calculations with ENDF-BVII.1, JENDL-4.0 and CENDL-3.1 libraries. • The SiC evaluated nuclear data from CENDL-3.1 library was checked for the first time and proved to be reliable. - Abstract: Benchmarking of evaluated nuclear data libraries was performed for 14 MeV neutrons on silicon carbide samples. The experiments were carried out by using the benchmark experimental facility at China Institute of Atomic Energy (CIAE). The leakage neutron spectra from SiC (Φ13 cm × 20 cm) at 60° and 120° and SiC (Φ13 cm × 2 cm) at 60° were measured by the TOF method. The measured spectra are well reproduced by MCNP-4C calculations with the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 evaluated nuclear data libraries, except 5–8 MeV range for 20 cm thickness. The discrepancies are mostly considered as caused by the improper evaluation of the angular distribution and secondary neutron energy distribution of the elastic scattering and inelastic scattering in evaluated nuclear data libraries.

  3. Neutron spectra due 13N production in a PET cyclotron

    International Nuclear Information System (INIS)

    Benavente, J.A.; Vega-Carrillo, H.R.; Lacerda, M.A.S.; Fonseca, T.C.F.; Faria, F.P.; Silva, T.A. da

    2015-01-01

    Monte Carlo and experimental methods have been used to characterize the neutron radiation field around PET (Positron Emission Tomography) cyclotrons. In this work, the Monte Carlo code MCNPX was used to estimate the neutron spectra, the neutron fluence rates and the ambient dose equivalent (H*(10)) in seven locations around a PET cyclotron during 13 N production. In order to validate these calculations, H*(10) was measured in three sites and were compared with the calculated doses. All the spectra have two peaks, one above 0.1 MeV due to the evaporation neutrons and another in the thermal region due to the room-return effects. Despite the relatively large difference between the measured and calculated H*(10) for one point, the agreement was considered good, compared with that obtained for 18 F production in a previous work. - Highlights: • MCNPX code was used to estimate the neutron spectra in a PET cyclotron. • Neutrons were estimated when 13 N is produced. • Neutron spectra show evaporation and room-return neutrons. • Calculated H*(10) were compared with measured H*(10)

  4. Neutron spectra measurement and calculations using data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 in iron benchmark assemblies

    Science.gov (United States)

    Jansky, Bohumil; Rejchrt, Jiri; Novak, Evzen; Losa, Evzen; Blokhin, Anatoly I.; Mitenkova, Elena

    2017-09-01

    The leakage neutron spectra measurements have been done on benchmark spherical assemblies - iron spheres with diameter of 20, 30, 50 and 100 cm. The Cf-252 neutron source was placed into the centre of iron sphere. The proton recoil method was used for neutron spectra measurement using spherical hydrogen proportional counters with diameter of 4 cm and with pressure of 400 and 1000 kPa. The neutron energy range of spectrometer is from 0.1 to 1.3 MeV. This energy interval represents about 85 % of all leakage neutrons from Fe sphere of diameter 50 cm and about of 74% for Fe sphere of diameter 100 cm. The adequate MCNP neutron spectra calculations based on data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 were done. Two calculations were done with CIELO library. The first one used data for all Fe-isotopes from CIELO and the second one (CIELO-56) used only Fe-56 data from CIELO and data for other Fe isotopes were from ENDF/B-VII.1. The energy structure used for calculations and measurements was 40 gpd (groups per decade) and 200 gpd. Structure 200 gpd represents lethargy step about of 1%. This relatively fine energy structure enables to analyze the Fe resonance neutron energy structure. The evaluated cross section data of Fe were validated on comparisons between the calculated and experimental spectra.

  5. Calculation of the effective delayed neutron fraction by TRIPOLI-4 code for IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Lee, Y.K.; Hugot, F.X.

    2011-01-01

    The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)

  6. Measurement of the neutron and gamma-ray spectra originating from a 14-MeV neutron source in liquid nitrogen and liquid air

    International Nuclear Information System (INIS)

    Broecker, B.; Clausen, K.; Schneider-Kuehnle, P.; Weinert, M.

    1975-01-01

    An experiment to measure the radiation transport originating from a 14-MeV neutron source in liquid nitrogen and liquid air is presented. Neutron and gamma-ray spectra were measured with a proton-recoil NE 213 scintillator and with four spherical proportional counters in a tank filled with liquid nitrogen or liquid air. The neutron spectra cover the energy range of 20 keV to 18 MeV. The source-detector separation varies in the liquid medium between 60 and 240 cm. The experimental setup is briefly described and the errors are estimated. (2 tables, 9 figures) (auth)

  7. A general formula considering one group delayed neutron under nonequilibrium condition

    International Nuclear Information System (INIS)

    Li Haofeng; Chen Wenzhen; Zhu Qian; Luo Lei

    2008-01-01

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

  8. Neutron quality parameters versus energy below 4 MeV from microdosimetric calculations

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  9. Ionizing Energy Depositions After Fast Neutron Interactions in Silicon

    CERN Document Server

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

    2016-01-01

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

  10. Determination of neutron spectra formed by 40-MeV deuteron bombardment of a lithium target with multi-foil activation technique

    CERN Document Server

    Maekawa, F; Wada, M; Wilson, P P H; Ikeda, Y

    2000-01-01

    Neutron flux spectra at an irradiation field produced by a 40-MeV deuteron bombardment on a thick lithium-target at Forschungszentrum Karlsruhe, Germany, have been determined by the multi-foil activation technique. Twenty-seven dosimetry reactions having a wide energy range of threshold energies up to 38 MeV were employed as detectors for the neutron flux spectra extending to 55 MeV. The spectra were adjusted with the SAND-II code with the experimental reaction rates based on an iterative method. The adjusted spectra validated quantitatively the Monte Carlo deuteron-lithium (d-Li) neutron source model code (M sup C DeLi) which was used to calculate initial guess spectra and also has been used for IFMIF nuclear designs. Accuracy of the adjusted spectra was approx 10% that was suitable for successive integral tests of activation cross section data.

  11. Detection of special nuclear material from delayed neutron emission induced by a dual-particle monoenergetic source

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, M. [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, Pennsylvania 16802 (United States); Nattress, J.; Jovanovic, I., E-mail: ijov@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan 48109 (United States)

    2016-06-27

    Detection of unique signatures of special nuclear materials is critical for their interdiction in a variety of nuclear security and nonproliferation scenarios. We report on the observation of delayed neutrons from fission of uranium induced in dual-particle active interrogation based on the {sup 11}B(d,n γ){sup 12}C nuclear reaction. Majority of the fissions are attributed to fast fission induced by the incident quasi-monoenergetic neutrons. A Li-doped glass–polymer composite scintillation neutron detector, which displays excellent neutron/γ discrimination at low energies, was used in the measurements, along with a recoil-based liquid scintillation detector. Time-dependent buildup and decay of delayed neutron emission from {sup 238}U were measured between the interrogating beam pulses and after the interrogating beam was turned off, respectively. Characteristic buildup and decay time profiles were compared to the common parametrization into six delayed neutron groups, finding a good agreement between the measurement and nuclear data. This method is promising for detecting fissile and fissionable materials in cargo scanning applications and can be readily integrated with transmission radiography using low-energy nuclear reaction sources.

  12. Compendium on neutron spectra in criticality accident dosimetry

    International Nuclear Information System (INIS)

    Ing, H.

    1978-01-01

    Graphical and tabulated neutron spectra are presented: from selected critical assemblies; from critical solutions; of fission neutrons through shielding; of H 2 O-moderated fission neutrons through shielding; of D 2 O-moderated fission neutrons through shielding; of fission neutrons reflected from various materials; from the D(T, 4 He)n reaction (''14 MeV'' neutrons) through shielding and of ''14 MeV'' neutrons reflected from various materials

  13. Study of beta-delayed neutron with proton-neutron QRPA plus statistical model

    International Nuclear Information System (INIS)

    Minato, Futoshi; Iwamoto, Osamu

    2015-01-01

    β-delayed neutron is known to be important for safety operation of nuclear reactor and prediction of elemental abundance after freeze-out of r-process. A lot of researches on it have been performed. However, the experimental data are far from complete since the lifetime of most of the relevant nuclei is so short that one cannot measure in a high efficiency. In order to estimate half-lives and delayed neutron emission probabilities of unexplored nuclei, we developed a new theoretical method which combines a proton-neutron quasi-particle random-phase-approximation and the Hauser-Feshbach statistical model. The present method reproduces experimentally known β-decay half-lives within a factor of 10 and about 40% of within a factor of 2. However it fails to reproduce delayed neutron emission probabilities. We discuss the problems and remedy for them to be made in future. (author)

  14. Review of experimental methods for evaluating effective delayed neutron fraction

    Energy Technology Data Exchange (ETDEWEB)

    Yamane, Yoshihiro [Nagoya Univ. (Japan). School of Engineering

    1997-03-01

    The International Effective Delayed Neutron Fraction ({beta}{sub eff}) Benchmark Experiments have been carried out at the Fast Critical Assembly of Japan Atomic Energy Research Institute since 1995. Researchers from six countries, namely France, Italy, Russia, U.S.A., Korea, and Japan, participate in this FCA project. Each team makes use of each experimental method, such as Frequency Method, Rossi-{alpha} Method, Nelson Number Method, Cf Neutron Source Method, and Covariance Method. In this report these experimental methods are reviewed. (author)

  15. Delayed neutron emission near the shell-closures

    Directory of Open Access Journals (Sweden)

    Borzov Ivan

    2016-01-01

    Full Text Available The self-consistent Density Functional + Continuum QRPA approach (DF+CQRPA provides a good description of the recent experimental beta-decay half-lives and delayed neutron emission branchings for the nuclei approaching to (and beyond the neutron closed shells N = 28; 50; 82. Predictions of beta-decay properties are more reliable than the ones of standard global approaches traditionally used for the r-process modelling. An impact of the quasi-particle phonon coupling on the delayed multi-neutron emission rates P2n, P3n,… near the closed shells is also discussed.

  16. Neutron spectra measurements and neutron flux monitoring for radiation damage purposes

    International Nuclear Information System (INIS)

    Osmera, B.; Petr, J.; Racek, J.; Rumler, C.; Turzik, Z.; Franc, L.; Holman, M.; Hogel, J.; Kovarik, K.; Marik, P.; Vespalec, R.; Albert, D.; Hansen, V.; Vogel, W.

    1979-09-01

    Neutron spectra were measured for the TR-0, WWR-S and SR-0 experimental reactors using the recoil proton method, 6 Li spectrometry, scintillation spectrometry and activation detectors in a variety of conditions. Neutron fluence was also measured and calculated. (M.S.)

  17. Kalman filter analysis of delayed neutron nondestructive assay measurements

    International Nuclear Information System (INIS)

    Aumeier, S. E.

    1998-01-01

    The ability to nondestructively determine the presence and quantity of fissile and fertile nuclei in various matrices is important in several nuclear applications including international and domestics safeguards, radioactive waste characterization and nuclear facility operations. Material irradiation followed by delayed neutron counting is a well known and useful nondestructive assay technique used to determine the fissile-effective content of assay samples. Previous studies have demonstrated the feasibility of using Kalman filters to unfold individual isotopic contributions to delayed neutron measurements resulting from the assay of mixes of uranium and plutonium isotopes. However, the studies in question used simulated measurement data and idealized parameters. We present the results of the Kalman filter analysis of several measurements of U/Pu mixes taken using Argonne National Laboratory's delayed neutron nondestructive assay device. The results demonstrate the use of Kalman filters as a signal processing tool to determine the fissile and fertile isotopic content of an assay sample from the aggregate delayed neutron response following neutron irradiation

  18. Beta-delayed proton emission in neutron-deficient lanthanide isotopes

    International Nuclear Information System (INIS)

    Wilmarth, P.A.

    1988-01-01

    Forty-two β-delayed proton precursors with 56≤Z≤71 and 63≤N≤83 were produced in heavy-ion reactions at the Lawrence Berkeley Laboratory SuperHILAC and their radioactive decay properties studied at the on-line mass separation facility OASIS. Twenty-five isotopes and eight delayed proton branches were identified for the first time. Delayed proton energy spectra and proton coincident γ-ray and x-ray spectra were measured for all precursors. In a few cases, proton branching ratios were also determined. The precursor mass numbers were determined by the separator, while the proton coincident x-ray energies provided unambiguous Z identifications. The proton coincident γ-ray intensities were used to extract final state branching ratios. Proton emission from ground and isomeric states was observed in many cases. The majority of the delayed proton spectra exhibited the smooth bell-shaped distribution expected for heavy mass precursors. The experimental results were compared to statistical model calculations using standard parameter sets. Calculations using Nilsson model/RPA β-strength functions were found to reproduce the spectral shapes and branching ratios better than calculations using either constant or gross theory β-strength functions. Precursor half-life predictions from the Nilsson model/RPA β-strength functions were also in better agreement with the measured half-lives than were gross theory predictions. The ratios of positron coincident proton intensities to total proton intensities were used to determine Q/sub EC/-B/sub p/ values for several precursors near N=82. The statistical model calculations were not able to reproduce the experimental results for N=81 precursors. 154 refs., 82 figs., 19 tabs

  19. Possibilities of delayed neutron fraction (βeff) calculation and measurement

    International Nuclear Information System (INIS)

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

    2008-01-01

    The influence of the delayed neutrons on the reactor dynamics can be understood through their impact on the reactor power change rate. In spite of the fact that delayed neutrons constitute only a very small fraction of the total number of neutrons generated from fission, they play a dominant role in the fission chain reaction control. If only the prompt neutrons existed, the reactor operation would become impossible due to the fast reactor power changes. The exact determination of delayed neutrons main parameter, the delayed neutron fraction (β eff ), is very important in the field of reactor physics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of calculations and experiments. In consequence of difficulties in β eff experimental measurement, this value in exact state use to be determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. Determination of β eff requires criticality calculations. In the past, k eff used to be traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum- weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. In this work, a summary of possible β eff calculation methods can be found and a calculation of β eff for VR-1 training reactor in one operation state is made using the prompt method, by MCNP5 code. Also a method of β eff kinetic measurement on VR-1 training reactor at Czech Technical University in Prague using in-pile kinetic technique is outlined (authors)

  20. Measurement of neutron spectra through composed material block bombarded with D-T neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, T.H. [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, P.O. BOX 919-213, Mian yang 621900 (China)], E-mail: zhutonghua@yahoo.com.cn; Liu, R.; Lu, X.X.; Jiang, L.; Wen, Z.W.; Wang, M.; Lin, J.F. [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, P.O. BOX 919-213, Mian yang 621900 (China)

    2009-12-15

    A 2-dimensional composed material assembly made of the iron and hydric block has been established. The neutron spectra from the assembly bombarded with 14-MeV neutrons at neutron generator have been obtained using the proton recoil technique with a stillbene detector. The detector positions were selected at the 60 deg., 120 deg., 180 deg. on the surface of the iron spherical shell. The background neutron spectra consisted of background and room return radiation were subtracted with combination of methods of experimental shielding and MCNP calculation. The uncertainty of results was 6.3-7.4%. The experiment results were analyzed and simulated by MCNP code and two data library. The difference is integral neutron flux (background neutron subtracted) of measured results greater than calculations with maximum of 21.2% in the range of 1-16 MeV.

  1. Statistical precision of delayed-neutron nondestructive assay techniques

    International Nuclear Information System (INIS)

    Bayne, C.K.; McNeany, S.R.

    1979-02-01

    A theoretical analysis of the statistical precision of delayed-neutron nondestructive assay instruments is presented. Such instruments measure the fissile content of nuclear fuel samples by neutron irradiation and delayed-neutron detection. The precision of these techniques is limited by the statistical nature of the nuclear decay process, but the precision can be optimized by proper selection of system operating parameters. Our method is a three-part analysis. We first present differential--difference equations describing the fundamental physics of the measurements. We then derive and present complete analytical solutions to these equations. Final equations governing the expected number and variance of delayed-neutron counts were computer programmed to calculate the relative statistical precision of specific system operating parameters. Our results show that Poisson statistics do not govern the number of counts accumulated in multiple irradiation-count cycles and that, in general, maximum count precision does not correspond with maximum count as first expected. Covariance between the counts of individual cycles must be considered in determining the optimum number of irradiation-count cycles and the optimum irradiation-to-count time ratio. For the assay system in use at ORNL, covariance effects are small, but for systems with short irradiation-to-count transition times, covariance effects force the optimum number of irradiation-count cycles to be half those giving maximum count. We conclude that the equations governing the expected value and variance of delayed-neutron counts have been derived in closed form. These have been computerized and can be used to select optimum operating parameters for delayed-neutron assay devices

  2. High energy neutron dosimetry for the fusion program

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. Neutron spectra determination methods using the measured reaction rates in SAPIS

    International Nuclear Information System (INIS)

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

    1980-01-01

    Mathematical basis of algorithms is given for methods of neutron spectra restoration in accordance with the measured reaction rates of the activation detectors included into the information-determination system SAIPS aimed at generalization of the most popular home and foreign neutron spectra determination methods as well as the establishment of their mutual relations. The following neutron spectra determination methods are described: SAND-II, CRYSTAL BALL, WINDOWS, SPECTRA, RESP, JUL; polynominal and directed divergence methods. The algorithms have been realized on the ES computer

  4. Subcritical Neutron Multiplication Measurements of HEU Using Delayed Neutrons as the Driving Source

    International Nuclear Information System (INIS)

    Hollas, C.L.; Goulding, C.A.; Myers, W.L.

    1999-01-01

    A new method for the determination of the multiplication of highly enriched uranium systems is presented. The method uses delayed neutrons to drive the HEU system. These delayed neutrons are from fission events induced by a pulsed 14-MeV neutron source. Between pulses, neutrons are detected within a medium efficiency neutron detector using 3 He ionization tubes within polyethylene enclosures. The neutron detection times are recorded relative to the initiation of the 14-MeV neutron pulse, and subsequently analyzed with the Feynman reduced variance method to extract singles, doubles and triples neutron counting rates. Measurements have been made on a set of nested hollow spheres of 93% enriched uranium, with mass values from 3.86 kg to 21.48 kg. The singles, doubles and triples counting rates for each uranium system are compared to calculations from point kinetics models of neutron multiplicity to assign multiplication values. These multiplication values are compared to those from MC NP K-Code calculations

  5. Experiment and analysis of neutron spectra in a concrete assembly bombarded by 14 MeV neutrons

    International Nuclear Information System (INIS)

    Oishi, Koji; Tomioka, Kazuyuki; Ikeda, Yujiro; Nakamura, Tomoo.

    1988-01-01

    Neutron spectrum in concrete bombarded by 14 MeV neutrons was measured using a miniature NE213 spectrometer and multi-foil activation method. A good agreement between those two experimental methods was obtained within experimental errors. The measured spectrum was compared with calculated ones using two-dimensional transport code DOT3.5 with 125 group structure cross section libraries based on ENDF/B-IV, JENDL-2, and JENDL-3T (the testing version of JENDL-3.) In the D-T neutron peak region, measured and calculated neutron spectra agreed well with each other for those libraries. However, disagreements of about -10 % to +50 % and -30 % to +40 % were obtained in the MeV region and still lower neutron energy range, respectively. As a result, it was concluded that those discrepancies were caused by the overestimation of secondary neutrons emitted by inelastic scattering from O, Si, and/or Ca which were the main components of concrete. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Shaughnessy, Dawn A. [Univ. of California, Berkeley, CA (United States)

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  8. Beta-decay rate and beta-delayed neutron emission probability of improved gross theory

    Science.gov (United States)

    Koura, Hiroyuki

    2014-09-01

    A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for unmeasured nuclei are adopted from the KTUY nuclear mass formula, which is based on the spherical-basis method. Considering the properties of the integrated Fermi function, we can roughly categorized energy region of excited-state of a daughter nucleus into three regions: a highly-excited energy region, which fully affect a delayed neutron probability, a middle energy region, which is estimated to contribute the decay heat, and a region neighboring the ground-state, which determines the beta-decay rate. Some results will be given in the presentation. A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for

  9. Neutron spectra and level density parameters from 16O + 12C fusion reaction

    International Nuclear Information System (INIS)

    Kasagi, J.; Remington, B.; Galonsky, A.; Haas, F.; Racca, R.; Prosser, F.W.

    1985-01-01

    Residues following 16 O + 12 C fusion were identified by their characteristic γ-rays. For several transitions in 23 Mg, 25 Mg, and 26 Al coincident neutron spectra were measured at six angles. Through use of the evaporation code CASCADE, comparisons were made of these spectra with predictions of the statistical model at five 16 O projectile energies between 43.2 and 56.0 MeV. The results require an excitation energy dependence for the effective radius parameter r 0 which determines the spin cutoff factor

  10. Neutron reference spectra measurements with the Bonner multi-spheres spectrometer

    International Nuclear Information System (INIS)

    Lemos Junior, Roberto Mendonca de

    2004-01-01

    This paper aims to define a procedure to use the Bonner Multisphere Spectrometer with a 6 LiI(Eu) detector in order to determine of neutron spectra. It was measured 238 PuBe spectra and same of reference ( 241 AmBe, 252 Cf e 252 Cf+D 2 O) published in ISO 8529-1 (2001) Norm. The data were processed by a computer program (BUNKI), which presents the results in neutrons energy fluency. Each input parameter of the program was studied in order to establish their influence in the adjustment result. The environment dose equivalent rate obtained placing the detector 1 m from the 241 AmBe source was 122 ± 4 μSv/h with 7% of uncertainty and 95% of confidence level. The procedure established in this work was tested with the 238 PuBe spectrum, obtaining an environment dose equivalent rate of 286 ± 9 μSv/h, 8% lower than the value measured experimentally used as reference. Through this procedure will be possible to measure neutron spectra in different work places where neutrons sources are used. Knowing these spectra, it will be possible to evaluate which area monitors, are more suitable, as well as, to study better the response of individual neutron monitors, as for instance, to obtain a conversion coefficient more appropriate to the albedo dosimeter used in different work places. As the measurements need a long time to be accomplished, the work optimization is fundamental to reduce the exposing time of the Bonner spectrometer operator. For this reason, an important parameter examined in this paper was the possibility of reducing the number of spheres used during the measurement without changing the final result. Considering the radiation protection standards, this parameter has a huge importance when the measurements are performed in work places where the neutron fluency and gamma rate offer risks to the operator's health, as for instance, in nuclear centrals. Studying this parameter, it was possible to conclude that removing the 20,32 cm diameter sphere it will be

  11. 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

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

    International Nuclear Information System (INIS)

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

    1993-02-01

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

  13. First Measurement of Several β-Delayed Neutron Emitting Isotopes Beyond N=126.

    Science.gov (United States)

    Caballero-Folch, R; Domingo-Pardo, C; Agramunt, J; Algora, A; Ameil, F; Arcones, A; Ayyad, Y; Benlliure, J; Borzov, I N; Bowry, M; Calviño, F; Cano-Ott, D; Cortés, G; Davinson, T; Dillmann, I; Estrade, A; Evdokimov, A; Faestermann, T; Farinon, F; Galaviz, D; García, A R; Geissel, H; Gelletly, W; Gernhäuser, R; Gómez-Hornillos, M B; Guerrero, C; Heil, M; Hinke, C; Knöbel, R; Kojouharov, I; Kurcewicz, J; Kurz, N; Litvinov, Yu A; Maier, L; Marganiec, J; Marketin, T; Marta, M; Martínez, T; Martínez-Pinedo, G; Montes, F; Mukha, I; Napoli, D R; Nociforo, C; Paradela, C; Pietri, S; Podolyák, Zs; Prochazka, A; Rice, S; Riego, A; Rubio, B; Schaffner, H; Scheidenberger, Ch; Smith, K; Sokol, E; Steiger, K; Sun, B; Taín, J L; Takechi, M; Testov, D; Weick, H; Wilson, E; Winfield, J S; Wood, R; Woods, P; Yeremin, A

    2016-07-01

    The β-delayed neutron emission probabilities of neutron rich Hg and Tl nuclei have been measured together with β-decay half-lives for 20 isotopes of Au, Hg, Tl, Pb, and Bi in the mass region N≳126. These are the heaviest species where neutron emission has been observed so far. These measurements provide key information to evaluate the performance of nuclear microscopic and phenomenological models in reproducing the high-energy part of the β-decay strength distribution. This provides important constraints on global theoretical models currently used in r-process nucleosynthesis.

  14. Calculation of neutron spectra produced in neutron generator target: Code testing.

    Science.gov (United States)

    Gaganov, V V

    2018-03-01

    DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  15. Evaluation of secondary and prompt fission neutron spectra

    Energy Technology Data Exchange (ETDEWEB)

    Porodzinskij, Yu.V.; Sukhovitskij, E.Sh. [Radiation Physics and Chemistry Problems Inst., Minsk-Sosny (Belarus)

    1997-03-01

    A simple model allowing to split neutron emission spectra into reaction partials is suggested. Predicted spectra of (n,n`{gamma}), (n,n`f), etc appear to be much harder than usually evaluated. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-11

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

  17. The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

    The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

  18. Modeling delayed neutron monitoring systems for fast breeder reactors

    International Nuclear Information System (INIS)

    Bunch, W.L.; Tang, E.L.

    1983-10-01

    The purpose of the present work was to develop a general expression relating the count rate of a delayed neutron monitoring system to the introduction rate of fission fragments into the sodium coolant of a fast breeder reactor. Most fast breeder reactors include a system for detecting the presence of breached fuel that permits contact between the sodium coolant and the mixed oxide fuel. These systems monitor for the presence of fission fragments in the sodium that emit delayed neutrons. For operational reasons, the goal is to relate the count rate of the delayed neutron monitor to the condition of the breach in order that appropriate action might be taken

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

    International Nuclear Information System (INIS)

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

    1976-11-01

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

  20. Scattering of 14.6 MeV neutrons from Fe and evidence for structure in the emitted neutron spectra

    International Nuclear Information System (INIS)

    Gul, K.; Anwar, M.; Ahmad, M.; Saleem, S.M.; Khan, N.A.

    1984-06-01

    Structure in the spectra of neutrons emitted from iron on bombardment with 14.6 MeV neutrons has been investigated and explained in terms of excitation of levels in iron 56. The energies of scattered neutrons have been measured by the time-of-flight technique based on the associated particle method. The observed excitations have been correlated with the reported levels in a satisfactory manner. Evidence for new excitations at 8.8 +- 0.02, 9.8 +- 0.1, 10.2 +- 0.1, 12.44 +- 0.03 and 12.52 +- 0.03 MeV has been obtained. The excitation of possible components of Ml giant resonance in iron 56 is discussed. (author)

  1. Theoretical analysis of time-dependent neutron spectra in bulk assemblies

    International Nuclear Information System (INIS)

    Akimoto, Tadashi; Ogawa, Yuichi; Togawa, Orihiko.

    1988-01-01

    Time-dependent neutron spectra in an iron assembly and in a graphite assembly are obtained with the one-dimensional S N calculation, in order an attempt to investigate the availability of these spectra to the benchmark test by the LINAC-TOF method for evaluation of nuclear data and numerical methods. The group constants are taken from the JAERI FAST SET Version 1, 2 and the ABBN SET. It was demonstrated by a sensitivity test that the time-dependent neutron spectra are sensitive to changes in the inelastic scattering cross section data in the iron assembly and to changes in the elastic scattering cross section data in the graphite assembly. Moreover, it is shown that the time-dependent spectra in the graphite assembly are sensitive to the group structure. Because some information about the neutron transport phenomena which has not been obtained in the stationary spectra is observed in the time-dependent spectra, the availability of the benchmark test based on the time-dependent spectra is indicated from the theoretical analysis. (author)

  2. Evaluation of energy collapsing effect on reactor kinetics parameters by diffusion theory

    International Nuclear Information System (INIS)

    Unesaki, Hironobu

    1989-01-01

    Reactor kinetics parameters play an important role as scaling factors between observed and calculated reactivities in the analysis of reactor physics experiments. In this report, energy collapsing errors in two kinetic parameters, the effective delayed neutron fraction and the neutron life time, are investigated by means of the diffusion theory. Coarse group calculations are made for various energy group structures. Cores of various moderator-to-fuel volume ratios are selected to investigate the influence of neutron spectrum changes on the energy collapsing error. The energy collapsing errors in the effective delayed neutron fraction and neutron life time are much larger than those in k eff . This might be because the former two parameters are functions of both the foward and adjoint flux, whereas the latter is a function of the forward flux alone. The use of coarse constants will cause errors in both fluxes, and the resulting errors in the former will be much more emphasized. As the effective delayed neutron fraction is sensitive to the treatment of an energy region in the vicinity of the fission spectrum peak, the coarse group error in it might differ between cores with different enrichment and composition. Inaccurate weighting of group constants leads to neutron spectra which do not conserve the fine group spectra, and those errors will be emphasized in calculated integral parameters. (N.K.)

  3. Two reports: (i) Correlation properties of delayed neutrons from fast neutron induced fission. (ii) Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Isaev, S.G.; Goverdovski, A.A.; Pshakin, G.M.

    1998-10-01

    The document includes the following two reports: 'Correlation properties of delayed neutrons from fast neutron induced fission' and 'Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting. A separate abstract was prepared for each report

  4. Artificial neural network for the determination of neutron spectra in the bunker of a Linac of 18 MV

    International Nuclear Information System (INIS)

    Banuelos F, A.; Borja H, C. G.; Valero L, C.; Guzman G, K. A.; Hernandez D, V. M.; Vega C, H. R.

    2011-11-01

    The neutron spectrum and equivalent of environmental dose H(10) were calculated for a radiotherapy room in 16 punctual detectors, 15 inside of and 1 outside of the same one. The calculations were carried out with the Monte Carlo method and with the code MCNP5 for a generic room model with a Linac of 18 MV, obtaining this way 16 spectra with 47 intervals of energy class, starting from these spectra the values of H(10) were calculated. On the other hand, an artificial neural network was designed and trained to determine the spectra by neutrons in 15 different locations inside the radiotherapy room starting from the value of H(10) in the detector 16 located in the exterior of the room, using as training data the spectra and calculated dose by neutrons, of which a medium quadratic error was obtained (m se) in the adjustment between the objective data and the exit data of m se=1E(-8). The results demonstrate that the use of the artificial intelligence as technique is an useful tool in the spectrometry and dosimetry of neutrons, since it simplifies the characterization process of neutron fields in radiotherapy rooms without the use of spectrometry systems, and that once the energy distribution of the neutrons produced by the Linac is known and the corresponding doses be calculated H(10), they can take the appropriate cautions for the security patient in treatment as well as for the personnel in the room. (Author)

  5. Neutron and gamma-ray spectra of 239PuBe and 241AmBe

    International Nuclear Information System (INIS)

    Vega-Carrillo, H.R.; Manzanares-Acuna, Eduardo; Becerra-Ferreiro, A.M.; Carrillo-Nunez, Aureliano

    2002-01-01

    Neutron and gamma-ray spectra of 239 PuBe and 241 AmBe were measured and their dosimetric features were calculated. Neutron spectra were measured using a multisphere neutron spectrometer with a 6 LiI(Eu) scintillator. The 239 PuBe neutron spectrum was measured in an open environment, while the 241 AmBe neutron spectrum was measured in a closed environment. Gamma-ray spectra were measured using a NaI(Tl) scintillator using the same experimental conditions for both sources. The effect of measuring conditions for the 241 AmBe neutron spectrum indicates the presence of epithermal and thermal neutrons. The low-resolution neutron spectra obtained with the multisphere spectrometer allows one to calculate the dosimetric features of neutron sources. At 100 cm both sources produce approximately the same count rate as that of the 4.4 MeV gamma-ray per unit of alpha emitter activity

  6. Measurement and calculation of neutron leakage spectra from slab samples of beryllium, gallium and tungsten irradiated with 14.8 MeV neutrons

    Science.gov (United States)

    Nie, Y. B.; Ruan, X. C.; Ren, J.; Zhang, S.; Han, R.; Bao, J.; Huang, H. X.; Ding, Y. Y.; Wu, H. C.; Liu, P.; Zhou, Z. Y.

    2017-09-01

    In order to make benchmark validation of the nuclear data for gallium (Ga), tungsten (W) and beryllium (Be) in existing modern evaluated nuclear data files, neutron leakage spectra in the range from 0.8 to 15 MeV from slab samples were measured by time-of-flight technique with a BC501 scintillation detector. The measurements were performed at China Institute of Atomic Energy (CIAE) using a D-T neutron source. The thicknesses of the slabs were 0.5 to 2.5 mean free path for 14.8 MeV neutrons, and the measured angles were chosen to be 60∘ and 120∘. The measured spectra were compared with those calculated by the continuous energy Monte-Carlo transport code MCNP, using the data from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 nuclear data files, the comparison between the experimental and calculated results show that: The results from all three libraries significantly underestimate the cross section in energy range of 10-13 MeV for Ga; For W, the calculated spectra using data from CENDL-3.1 and JENDL-4.0 libraries show larger discrepancies with the measured ones, especially around 8.5-13.5 MeV; and for Be, all the libraries led to underestimation below 3 MeV at 120∘.

  7. Neutron spectra characteristics for the intense neutron source, INS

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  8. Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. M.; Sum, G. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified.

  9. Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar

    International Nuclear Information System (INIS)

    Lee, K. M.; Sum, G. M.

    2016-01-01

    Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  12. Comparison of dynamic compensation methods for delayed self-powered neutron detector

    International Nuclear Information System (INIS)

    In, Wang Kee; Kim, Joon Sung; Auh, Geun Sun; Yoon, Tae Young

    1993-01-01

    Dynamic compensation methods for rhodium self-powered neutron detector have been developed by Banda and Hoppe to compensate for the time delay associated with detector signals. The time delay is due to the decay of the neutron-activated rhodium and results in delayed detector response. Two digital dynamic compensation methods, were compared for step change of neutron flux in this paper. The inverse kinetics method gave slightly better response time and noise gain. However, the inverse kinetics method also showed overshooting of neutron flux for the step change. (Author)

  13. Neutron-energy-dependent cell survival and oncogenic transformation.

    Science.gov (United States)

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

    1999-12-01

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

  14. Investigation of the neutron emission spectra of some deformed nuclei for (n, xn) reactions up to 26 MeV energy

    International Nuclear Information System (INIS)

    Kaplan, A.; Bueyuekuslu, H.; Tel, E.; Aydin, A.; Boeluekdemir, M.H.

    2011-01-01

    In this study, neutron-emission spectra produced by (n, xn) reactions up to 26 MeV for some deformed target nuclei as 165 Ho, 181 Ta, 184 W, 232 Th and 238 U have been investigated. Also, the mean free path parameter's effect for 9n, xn) neutron-emission spectra has been examined. In the calculations, pre-equilibrium neutron-emission spectra have been calculated by using new evaluated hybrid model and geometry dependent hybrid model, full exciton model and cascade exciton model. The reaction equilibrium component has been calculated by Weisskopf-Ewing model. The obtained results have been discussed and compared with the available experimental data and found agreement with each other. (author)

  15. Delayed neutrons in ANSTO

    International Nuclear Information System (INIS)

    Wall, T.

    1988-01-01

    Delayed neutron analysis carried out at the Australian Nuclear Scientific and Technology Organization facilities, provides a fast, high sensitivity, low cost, reliable method, particularly suitable for large batches of samples, and for non destructive analysis of a range of materials. While its main use has been in uranium exploration, other applications include archeological investigations, agriculture, oceanography and biology

  16. First measurement of several $\\beta$-delayed neutron emitting isotopes beyond N=126

    CERN Document Server

    Caballero-Folch, R.; Agramunt, J.; Algora, A.; Ameil, F.; Arcones, A.; Ayyad, Y.; Benlliure, J.; Borzov, I.N.; Bowry, M.; Calvino, F.; Cano-Ott, D.; Cortés, G.; Davinson, T.; Dillmann, I.; Estrade, A.; Evdokimov, A.; Faestermann, T.; Farinon, F.; Galaviz, D.; García, A.R.; Geissel, H.; Gelletly, W.; Gernhäuser, R.; Gómez-Hornillos, M.B.; Guerrero, C.; Heil, M.; Hinke, C.; Knöbel, R.; Kojouharov, I.; Kurcewicz, J.; Kurz, N.; Litvinov, Y.; Maier, L.; Marganiec, J.; Marketin, T.; Marta, M.; Martínez, T.; Martínez-Pinedo, G.; Montes, F.; Mukha, I.; Napoli, D.R.; Nociforo, C.; Paradela, C.; Pietri, S.; Podolyák, Zs.; Prochazka, A.; Rice, S.; Riego, A.; Rubio, B.; Schaffner, H.; Scheidenberger, Ch.; Smith, K.; Sokol, E.; Steiger, K.; Sun, B.; Taín, J.L.; Takechi, M.; Testov, D.; Weick, H.; Wilson, E.; Winfield, J.S.; Wood, R.; Woods, P.; Yeremin, A.

    2016-01-01

    The $\\beta$-delayed neutron emission probabilities of neutron rich Hg and Tl nuclei have been measured together with $\\beta$-decay half-lives for 20 isotopes of Au, Hg, Tl, Pb and Bi in the mass region N$\\gtrsim$126. These are the heaviest species where neutron emission has been observed so far. These measurements provide key information to evaluate the performance of nuclear microscopic and phenomenological models in reproducing the high-energy part of the $\\beta$-decay strength distribution. In doing so, it provides important constraints to global theoretical models currently used in $r$-process nucleosynthesis.

  17. An In-Pile Kinetic Method for Determining the Delayed Neutron Fraction βeff

    International Nuclear Information System (INIS)

    Gilad, E.; Rivin, O.; Ettedgui, H.; Yaar, I.; Geslot, B.; Pepino, A.; Di Salvo, J.; Gruel, A.; Blaise, P.

    2014-01-01

    Delayed neutrons are of fundamental importance in the field of nuclear reactor dynamics and control. Although only a small fraction of the neutrons emitted by fission are not prompt, the knowledge of the delayed neutrons parameters is essential for transient analysis, such as startup or shutdown of the reactor, as well as for accidents analysis and control system design [1]. One of the main delayed neutron parameters used in the point reactor model equations is the effective delayed neutron fraction, which incorporates both delayed neutron spectral properties and core geometrical configuration [1,2]. Additional delayed neutron parameters include the fraction of fission neutrons emitted in each delayed group, and the delayed neutron precursors decay constants . Experimental efforts aimed at determining the value ofβ, which provide experimental support for the evaluation of delayed neutron parameters, are extremely valuable. This is due to the fact that unlike other fields in reactor physics, e.g. criticality safety or shielding, the availability of experimental data and benchmark problems for validating delayed neutron parameters and its implementation in different models is highly limited. Furthermore, the existing experimental data exhibit significant discrepancies between the different sets of parameter, which lead to substantial disparity in the analysis of kinetic experiments and reactor dynamic behavior]. In this work, a method for determining the effective delayed neutron fraction using in-pile reactivity oscillation and Fourier analysis is presented. The method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. Knowledge of the reactor transfer function enables the estimation of theβ value using multi-parameter nonlinear fit. The method accounts for higher harmonics, which are excited by the trapezoidal

  18. Fast neutron and gamma-ray spectra measurements with a NE-213 spectrometer in the FNG Copper Benchmark Experiment

    International Nuclear Information System (INIS)

    Klix, Axel; Angelone, Maurizio; Fischer, Ulrich; Pillon, Mario

    2016-01-01

    Highlights: • Fast neutron and gamma-ray spectra were measured in a copper assembly irradiated with DT neutrons. • The results were compared with MCNP calculations. • Primary aim was to provide experimental data for checking and validation of nuclear data evaluations of copper. - Abstract: A neutronics benchmark experiment on a pure Copper assembly was performed at the Frascati Neutron Generator. The work aimed at testing of recent nuclear data libraries. This paper focuses on the measurement of fast neutron and gamma-ray flux spectra in the Copper assembly under DT neutron irradiation in two selected positions with a spectrometer based on the organic liquid scintillator NE-213. The measurement results were compared with Monte Carlo radiation transport calculations using MCNP and nuclear data from the JEFF-3.1.1 library. Calculations have been done with Cu data from JEFF-3.1.1, JEFF-3.2, FENDL-3 and ENDF/B-7.0. Discrepancies appear in the intermediate neutron energy range between experiment and calculation. Large discrepancies were observed in the gamma-ray spectra calculated with JEFF-3.2.

  19. Fast neutron and gamma-ray spectra measurements with a NE-213 spectrometer in the FNG Copper Benchmark Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Klix, Axel, E-mail: axel.klix@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Angelone, Maurizio [ENEA Dipartimento Fusione e Tecnologie per la Sicurezza Nucleare, C.R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Fischer, Ulrich [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pillon, Mario [ENEA Dipartimento Fusione e Tecnologie per la Sicurezza Nucleare, C.R. Frascati, via E. Fermi 45, 00044 Frascati (Italy)

    2016-11-01

    Highlights: • Fast neutron and gamma-ray spectra were measured in a copper assembly irradiated with DT neutrons. • The results were compared with MCNP calculations. • Primary aim was to provide experimental data for checking and validation of nuclear data evaluations of copper. - Abstract: A neutronics benchmark experiment on a pure Copper assembly was performed at the Frascati Neutron Generator. The work aimed at testing of recent nuclear data libraries. This paper focuses on the measurement of fast neutron and gamma-ray flux spectra in the Copper assembly under DT neutron irradiation in two selected positions with a spectrometer based on the organic liquid scintillator NE-213. The measurement results were compared with Monte Carlo radiation transport calculations using MCNP and nuclear data from the JEFF-3.1.1 library. Calculations have been done with Cu data from JEFF-3.1.1, JEFF-3.2, FENDL-3 and ENDF/B-7.0. Discrepancies appear in the intermediate neutron energy range between experiment and calculation. Large discrepancies were observed in the gamma-ray spectra calculated with JEFF-3.2.

  20. Measurement of the fast neutron background at the China Jinping Underground Laboratory

    Science.gov (United States)

    Du, Q.; Lin, S. T.; Liu, S. K.; Tang, C. J.; Wang, L.; Wei, W. W.; Wong, H. T.; Xing, H. Y.; Yue, Q.; Zhu, J. J.

    2018-05-01

    We report on the measurements of the fluxes and spectra of the environmental fast neutron background at the China Jinping Underground Laboratory (CJPL) with a rock overburden of about 6700 meters water equivalent, using a liquid scintillator detector doped with 0.5% gadolinium. The signature of a prompt nuclear recoil followed by a delayed high energy γ-ray cascade is used to identify neutron events. The large energy deposition of the delayed γ-rays from the (n , γ) reaction on gadolinium, together with the excellent n- γ discrimination capability provides a powerful background suppression which allows the measurement of a low intensity neutron flux. The neutron flux of (1 . 51 ± 0 . 03(stat .) ± 0 . 10(syst .)) × 10-7cm-2s-1 in the energy range of 1-10 MeV in the Hall A of CJPL was measured based on 356 days of data. In the same energy region, measurement with the same detector placed in a room surrounding with one meter thick polyethylene shielding gives a significantly lower flux of (4 . 9 ± 0 . 9(stat .) ± 0 . 5(syst .)) × 10-9cm-2s-1 with 174 days of data. This represents a measurement of the lowest environmental fast neutron background among the underground laboratories in the world, prior to additional experiment-specific attenuation. Additionally, the fast neutron spectra both in the Hall A and the polyethylene room were reconstructed with the help of GEANT4 simulations.

  1. APPLE-2: an improved version of APPLE code for plotting neutron and gamma ray spectra and reaction rates

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitsu; Seki, Yasushi.

    1982-07-01

    A computer code APPLE-2 which plots the spatial distribution of energy spectra of multi-group neutron and/or gamma ray fluxes, and reaction rates has been developed. This code is an improved version of the previously developed APPLE code and has the following features: (1) It plots energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT and MORSE. (2) It calculates and plots the spatial distribution of neutron and gamma ray fluxes and various types of reaction rates such as nuclear heating rates, operational dose rates, displacement damage rates. (3) Input data specification is greatly simplified by the use of standard, response libraries and by close coupling with radiation transport calculation codes. (4) Plotting outputs are given in camera ready form. (author)

  2. Experimental determination of neutron capture cross sections of fast reactor structure materials integrated in intermediate energy spectra. Vol. 2: description of experimental structure

    International Nuclear Information System (INIS)

    Tassan, S.

    1978-01-01

    A selection of technical documents is given concerning the experimental determination of the neutron capture cross-sections of fast reactor structural materials (Fe, Cr, Ni...) integrated over the intermediate energy spectra. The experimental structure project and modifications of the reactor RB2 for this experiment, together with criticality and safety calculations, are presented

  3. Spallation Neutron Emission Spectra in Some Amphoter Target Nuclei by Proton Beam Up to 140 MeV Energy

    International Nuclear Information System (INIS)

    Yildirim, G.

    2008-01-01

    In the present study, the (p,xn) reaction neutron-emission spectra for some amphoter target nuclei as 27 A l, 64 Z n, 120 S n, and 208 P b were investigated up to 140 MeV incident proton energy. The pre-equilibrium calculations were calculated by using the hybrid model, the geometry dependent hybrid model, the full exciton model and the cascade exciton model. The reaction equilibrium component was calculated with a traditional compound nucleus model developed by Weisskopf Ewing. Calculation results have been discussed and compared with the available experimental data in literature

  4. Theory of neutron spectra from d-d-reactions in the linear z-pinch and the plasma focus

    International Nuclear Information System (INIS)

    Deutsch, R.; Kaeppeler, H.J.

    1982-05-01

    Because of a finite gyroradius effect, the equilibrium probability density function of the ions in the azimuthal magnetic field of a linear z-pinch becomes anisotropic. This density function was derived by solving the Vlasov equation and used to determine the neutron spectra produced in the deuterium plasma of a z-pinch. The neutron spectra were calculated for two models, differing in the energy distribution of the fast ions. A background plasma with 'slow' ions was also considered. The interactions of the fast ions with the slow ions and 'beam-beam' interactions between fast ions were considered. Typical spectra for arbitrary directions to the cylindrical axis are given. The anisotropy factors were calculated. Considering the influence of the azimuthal magnetic field on the equilibrium density function of the deuterons, the well known particularities of the neutron spectra are obtained without any of the contradictions typical of the traditional models. (orig.)

  5. Logic based feature detection on incore neutron spectra

    International Nuclear Information System (INIS)

    Bende-Farkas, S.; Kiss, S.; Racz, A.

    1992-09-01

    A methodology is proposed to investigate neutron spectra in such a way which is similar to human thinking. The goal was to save experts from tedious, mechanical tasks of browsing a large amount of signals in order to recognize changes in the underlying mechanisms. The general framework for detecting features of incore neutron spectra with a rulebased methodology is presented. As an example, the meaningful peaks in the APSDs are determined. This method is a part of a wider project to develop a noise diagnostic expert system. (R.P.) 6 refs.; 6 figs.; 1 tab

  6. Use of the foil activation method with arbitrary trial functions to determine neutron energy spectra

    International Nuclear Information System (INIS)

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

    1987-01-01

    Neutron Spectra have been measured by the foil activation method in thirteen different environments in and around the Sandia Pulsed Reactor (SPR-III), the White Sands Missile Range FBR, and the Annular Core Research Reactor (ACRR). The unfolded spectra were obtained by using the SANDII code in a manner which was not dependent on the initial trial. This altered technique is, therefore, better suited for the determination of spectra in environments that are difficult to predict by calculation, and it tends to reveal features that may be biased out by the use of standard trial functions

  7. Reconstruction of neutron spectra through neural networks; Reconstruccion de espectros de neutrones mediante redes neuronales

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E. [Cuerpo Academico de Radiobiologia, Estudios Nucleares, Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico)] e-mail: rvega@cantera.reduaz.mx [and others

    2003-07-01

    A neural network has been used to reconstruct the neutron spectra starting from the counting rates of the detectors of the Bonner sphere spectrophotometric system. A group of 56 neutron spectra was selected to calculate the counting rates that would produce in a Bonner sphere system, with these data and the spectra it was trained the neural network. To prove the performance of the net, 12 spectra were used, 6 were taken of the group used for the training, 3 were obtained of mathematical functions and those other 3 correspond to real spectra. When comparing the original spectra of those reconstructed by the net we find that our net has a poor performance when reconstructing monoenergetic spectra, this attributes it to those characteristic of the spectra used for the training of the neural network, however for the other groups of spectra the results of the net are appropriate with the prospective ones. (Author)

  8. Study of Beta-Delayed Neutron Emission by Neutron-Rich Nuclei and Analysis of the Nuclear Reaction Mechanism responsible for the Yields of these Nuclei

    International Nuclear Information System (INIS)

    Bazin, D.

    1987-07-01

    Among the nuclear mechanisms used for the production of nuclei far from stability, the projectile fragmentation process has recently proved its efficiency. However, at Fermi energies, one has to take into account some collective and relaxation effects which drastically modify the production cross-sections. The spectroscopic study of very neutron-rich nuclei is very dependent of these production rates. A study of beta-delayed neutron emission which leads to new measurements of half-lives and neutron delayed emission probabilities is achieved with a liquid scintillator detector. The results which are then compared to different theories are of interest for the understanding of natural production of heavy elements (r processus) [fr

  9. Compilation of neutron flux density spectra and reaction rates in different neutron fields

    International Nuclear Information System (INIS)

    Ertek, C.

    1979-07-01

    Upon the recommendation of International Working Group of Reactor Radiation Measurements (IWGRRM), the compilation of neutron flux density spectra and the reaction rates obtained by activation and fission foils in different neutron fields is presented. The neutron fields considered are as follows: 1/E; iron block; LWR core and pressure vessel; LMFBR core and blanket; CTR first wall and blanket; fission spectrum

  10. Particle unstable excited states in /sup 9/Be influence of beta recoil and width on delayed particle spectra

    CERN Document Server

    Nyman, G H; Jonson, B; Kratz, K L; Larsson, P O; Mattsson, S; Ziegert, W

    1981-01-01

    The light nucleus /sup 9/Be has been studied through the emission of beta-delayed neutrons and alpha particles from /sup 9/Li. The activity is produced at the ISOLDE facility in fragmentation reactions induced either by 600 MeV proton or 910 MeV /sup 3/He beams from the CERN Synchro-cyclotron. After mass separation neutron spectra are recorded using /sup 3/He-filled proportional counters, while surface barrier detectors are used for the spectroscopy of alpha particles. Effects on the spectrum shape induced by recoil and polarization phenomena as well as large widths of the intermediate states are discussed. (14 refs).

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  12. The structure of the Gamow-Teller giant resonance and consequences for beta-delayed neutron spectra and element synthesis

    International Nuclear Information System (INIS)

    Klapdor, H.V.

    1976-01-01

    Recent results in β-delayed neutron emission are interpreted by structure of the Gamow-Teller giant resonance not included in the 'gross-theory' of β-decay. Inclusion of this structure of the β-decay function is important for calculations of β-decay production rates for heavy nuclides by astrophysical processes and thermonuclear explosions. (Auth.)

  13. Measurement of secondary neutrons and gamma rays produced by neutron interactions in aluminum over the incident energy range 1 to 20 MeV

    International Nuclear Information System (INIS)

    Morgan, G.L.

    1975-11-01

    The spectra of secondary neutrons and gamma rays produced by neutron interaction in a thin sample (approximately 1/6 mean free path) of aluminum have been measured as a function of the incident neutron energy over the range 1 to 20 MeV. Data were taken at an angle of 125 0 . A linac (ORELA) was used as a neutron source with a 47-m flight path. Incident energy was determined by time-of-flight, while secondary spectra were determined by pulse-height unfolding techniques. The results of the measurements are presented in forms suitable for comparison to calculations based on the evaluated data files. (6 tables, 4 figures)

  14. Statistical and non statistical models for delayed neutron emission: applications to nuclei near A = 90

    International Nuclear Information System (INIS)

    De Oliveira, Z.M.

    1980-01-01

    A detailed analysis of the simple statistical model description for delayed neutron emission of 87 Br, 137 I, 85 As and 135 Sb has been performed. In agreement with experimental findings, structure in the #betta#-strength function is required to reproduce the envelope of the neutron spectrum from 87 Br. For 85 As and 135 Sb the model is found incapable of simultaneously reproducing envelopes of delayed neutron spectra and neutron branching ratios to excited states in the final nuclei for any choice of #betta#-strength function. The results indicate that partial widths for neutron emission are not compatible with optical-model transmission coefficients. The simple shell model with pairing is shown to qualitatively describe the main features of the #betta#-strength functions for decay of 87 Br and 91 93 95 97 Rb. It is found that the location of apparent resonances in the experimental data are in rough agreement with the location of centroids of strength calculated with this model. An extension of the shell model picture which includes the Gamow-Teller residual interaction is used to investigate decay properties of 84 86 As, 86 92 Br and 88 102 Rb. For a realistic choice of interaction strength, the half lives of these isotopes are fairly well reproduced and semiquantitative agreement with experimental #betta#-strength functions is found. Delayed neutron emission probabilities are reproduced for precursors nearer stability with systematic deviations being observed for the heavier nuclei. Contrary to the assumption of a structureless Gamow-Teller giant resonance as embodied gross theory of #betta#-decay, we find that structures in the tail of the Gamow-Teller giant resonances are expected which strongly influence the decay properties of nuclides in this region

  15. A possible island of beta-delayed neutron precursors in heavy nucleus region

    International Nuclear Information System (INIS)

    Zhang Li

    1991-01-01

    The possible Beta-Delayed neutron precursors in the elements Tl, Hg, and Au were predicted following a systematic research on the known Beta-Delayed neutron precursors. The masses of the unknown nuclei and neutron emission probabilities were calculated

  16. Measurements of the energy spectrum of backscattered fast neutrons

    International Nuclear Information System (INIS)

    Segal, Y.

    1976-03-01

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

  17. 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.

  18. Measurement of keV-neutron capture cross sections and capture gamma-ray spectra of Er isotopes

    International Nuclear Information System (INIS)

    Harun-Ar-Rashid, A.K.M.; Igashira, Masayuki; Ohsaki, Toshiro

    2000-01-01

    Neutron capture cross sections and capture γ-ray spectra of 166,167, 168 Er were measured in the energy region of 10 to 550 keV. The measurements were performed with a pulsed 7 Li(p,n) 7 Be neutron source and a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique and the standard capture cross sections of gold were used to derive the capture cross sections. The errors of the derived cross sections were about 5%. The present results were compared with other measurements and evaluations. The observed capture γ-ray pulse-height spectra were unfolded to obtain the corresponding γ-ray spectra. An anomalous shoulder was observed around 3 MeV in each of the capture γ-ray spectra. (author)

  19. Extension of the calibration of an NE-213 liquid scintillator based pulse height response spectrometer up to 18 MeV neutron energy and leakage spectrum measurements on bismuth at 8 MeV and 18 MeV neutron energies

    International Nuclear Information System (INIS)

    Fenyvesi, A.; Valastyan, I.; Olah, L.; Csikai, J.; Plompen, A.; Jaime, R.; Loevestam, G.; Semkova, V.

    2011-01-01

    Monoenergetic neutrons were produced at the Van de Graaff accelerator of the EC-JRC-Institute for Reference Materials and Measurements (IRMM, Geel, Belgium). An air-jet cooled D_2-gas target (1.2 bar, ΔE_d = 448 keV) was bombarded with E_d =4976 keV deuterons to produce neutrons up to E_n = 8 MeV energy via the D(d,n)"3He reaction. Higher energy neutrons up to E_n = 18 MeV were produced via the T(d,n)"4He reaction by bombarding a TiT target with E_d =1968 keV deuterons. Pulse height spectra were measured at different neutron energies from E_n = 8 MeV up to E_n = 18 MeV with the NE-213 liquid scintillator based Pulse Height Response Spectrometer (PHRS) of UD-IEP. The energy calibration of the PHRS system has been extended up to E_n = 18 MeV. Pulse height spectra induced by gamma photons have been simulated by the GRESP7 code. Neutron induced pulse height spectra have been simulated by the NRESP7 and MCNP-POLIMI codes. Comparison of the results of measurements and simulations enables the improvement of the parameter set of the function used by us to describe the light output dependence of the resolution of the PHRS system at light outputs of L > 2 light units. Also, it has been shown that the derivation method for unfolding neutron spectra from measured pulse height spectra performs well when relative measurements are done up to E_n = 18 MeV neutron energy. For matrix unfolding purposes, the NRESP7 code has to be preferred to calculate the pulse height response matrix of the PHRS system. Leakage spectra of neutrons behind bismuth slabs of different thicknesses have been measured with the PHRS system by using monoenergetic neutrons. The maximum slab thickness was d = 14 cm. Simulations of the measurements have been carried out with the MCNP-4c code. The necessary nuclear cross-sections were taken from the from the ENDF/B-VII and JEFF.3.1 data libraries. For both libraries, the agreement of measured and simulated neutron spectra is good for the 5 MeV ≤ En ≤ 18 Me

  20. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    International Nuclear Information System (INIS)

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

  1. Population of delayed-neutron granddaughter states and the optical potential

    International Nuclear Information System (INIS)

    Schenter, R.E.; Mann, F.M.; Warner, R.A.; Reeder, P.L.

    1982-08-01

    Using a statistical treatment of beta decay and the Hauser-Feshbach model of nuclear reactions, calculations were made and compared to recent experimental measurements of the population of granddaughter states of several delayed neutron precursors ( 144 145 147 Cs and 96 Rb). Emphasis of this paper is on the sensitivity and interpretation of experimental results to various standard low energy neutron optical model potentials and variations in their forms and parameters. Results for these precursors show qualitative agreement with experiment for all the optical potential models used and good quantitative agreement for two (Moldauer and Becchetti-Greenlees). Questions such as (N-Z) terms, deformation and nonlocality dependence are presented

  2. 8-group relative delayed neutron yields for epithermal neutron induced fission of 235U and 239Pu

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. High energy proton simulation of 14-MeV neutron damage in Al2O3

    International Nuclear Information System (INIS)

    Muir, D.W.; Bunch, J.M.

    1975-01-01

    High-energy protons are a potentially useful tool for simulating the radiation damage produced by 14-MeV neutrons in CTR materials. A comparison is given of calculations and measurements of the relative damage effectiveness of these two types of radiation in single-crystal Al 2 O 3 . The experiments make use of the prominent absorption band at 206 nm as an index to lattice damage, on the assumption that peak absorption is proportional to the concentration of lattice vacancies. The induced absorption is measured for incident proton energies ranging from 5 to 15 MeV and for 14-MeV neutrons. Recoil-energy spectra are calculated for elastic and inelastic scattering using published angular distributions. Recoil-energy spectra also are calculated for the secondary alpha particles and 12 C nuclei produced by (p,p'α) reactions on 16 O. The recoil spectra are converted to damage-energy spectra and then integrated to yield the damage-energy cross section at each proton energy and for 14 MeV neutrons. A comparison of the calculations with experimental results suggests that damage energy, at least at high energies, is a reasonable criterion for estimating this type of radiation damage. (auth)

  4. Leakage monitoring equipment of fuel element by delayed neutron method

    International Nuclear Information System (INIS)

    Ji Changsong; Zhang Shulan; Zhang Shuheng

    1999-01-01

    Based on monitoring results of delayed neutrons from reactor first circle water, the leakage of reactor fuel elements is monitored. A monitoring equipment consisted of an array of 3 He proportional counter tubes with 75 s delay has been developed. The neutron detection efficiency of 6.1% is obtained

  5. Temperature-tuned Maxwell-Boltzmann neutron spectra for kT ranging from 30 up to 50 keV for nuclear astrophysics studies.

    Science.gov (United States)

    Martín-Hernández, G; Mastinu, P F; Praena, J; Dzysiuk, N; Capote Noy, R; Pignatari, M

    2012-08-01

    The need of neutron capture cross section measurements for astrophysics motivates present work, where calculations to generate stellar neutron spectra at different temperatures are performed. The accelerator-based (7)Li(p,n)(7)Be reaction is used. Shaping the proton beam energy and the sample covering a specific solid angle, neutron activation for measuring stellar-averaged capture cross section can be done. High-quality Maxwell-Boltzmann neutron spectra are predicted. Assuming a general behavior of the neutron capture cross section a weighted fit of the spectrum to Maxwell-Boltzmann distributions is successfully introduced. Copyright © 2012 Elsevier Ltd. All rights reserved.

  6. Use of one delayed-neutron precursor group in transient analysis

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1983-01-01

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

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

    Science.gov (United States)

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

    2015-09-11

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

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

    Science.gov (United States)

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

    2018-04-01

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

  9. The delayed neutron method of uranium analysis

    International Nuclear Information System (INIS)

    Wall, T.

    1989-01-01

    The technique of delayed neutron analysis (DNA) is discussed. The DNA rig installed on the MOATA reactor, the assay standards and the types of samples which have been assayed are described. Of the total sample throughput of about 55,000 units since the uranium analysis service began, some 78% has been concerned with analysis of uranium ore samples derived from mining and exploration. Delayed neutron analysis provides a high sensitivity, low cost uranium analysis method for both uranium exploration and other applications. It is particularly suitable for analysis of large batch samples and for non-destructive analysis over a wide range of matrices. 8 refs., 4 figs., 3 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-11

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

  11. Calculational analysis of errors for various models of an experiment on measuring leakage neutron spectra

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Deeva, V.V.; Prokof'eva, Z.A.

    1990-01-01

    Analysis is made for the effect of mathematical model accuracy of the system concerned on the calculation results using the BRAND program system. Consideration is given to the impact of the following factors: accuracy of neutron source energy-angular characteristics description, various degrees of system geometry approximation, adequacy of Monte-Carlo method estimation to a real physical neutron detector. The calculation results analysis is made on the basis of the experiments on leakage neutron spectra measurement in spherical lead assemblies with the 14 MeV-neutron source in the centre. 4 refs.; 2 figs.; 10 tabs

  12. Some properties of zero power neutron noise in a time-varying medium with delayed neutrons

    International Nuclear Information System (INIS)

    Kitamura, Y.; Pal, L.; Pazsit, I.; Yamamoto, A.; Yamane, Y.

    2008-01-01

    The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed

  13. Fusion materials high energy-neutron studies. A status report

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  14. Very High Energy Neutron Scattering from Hydrogen

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  15. The universal library of fission products and delayed neutron group yields

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Science.gov (United States)

    Hosseini, Seyed Abolfazl; Esmaili Paeen Afrakoti, Iman

    2018-01-17

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

  17. Analysis of accelerator based neutron spectra for BNCT using proton recoil spectroscopy

    International Nuclear Information System (INIS)

    Wielopolski, L.; Ludewig, H.; Powell, J.R.; Raparia, D.; Alessi, J.G.; Lowenstein, D.I.

    1998-01-01

    Boron Neutron Capture Therapy (BNCT) is a promising binary treatment modality for high-grade primary brain tumors (glioblastoma multiforme, GM) and other cancers. BNCT employs a boron-10 containing compound that preferentially accumulates in the cancer cells in the brain. Upon neutron capture by 10 B energetic alpha particles and triton released at the absorption site kill the cancer cell. In order to gain penetration depth in the brain Fairchild proposed, for this purpose, the use of energetic epithermal neutrons at about 10 keV. Phase I/II clinical trials of BNCT for GM are underway at the Brookhaven Medical Research Reactor (BMRR) and at the MIT Reactor, using these nuclear reactors as the source for epithermal neutrons. In light of the limitations of new reactor installations, e.g. cost, safety and licensing, and limited capability for modulating the reactor based neutron beam energy spectra alternative neutron sources are being contemplated for wider implementation of this modality in a hospital environment. For example, accelerator based neutron sources offer the possibility of tailoring the neutron beams, in terms of improved depth-dose distributions, to the individual and offer, with relative ease, the capability of modifying the neutron beam energy and port size. In previous work new concepts for compact accelerator/target configuration were published. In this work, using the Van de Graaff accelerator the authors have explored different materials for filtering and reflecting neutron beams produced by irradiating a thick Li target with 1.8 to 2.5 MeV proton beams. However, since the yield and the maximum neutron energy emerging from the Li-7(p,n)Be-7 reaction increase with increase in the proton beam energy, there is a need for optimization of the proton energy versus filter and shielding requirements to obtain the desired epithermal neutron beam. The MCNP-4A computer code was used for the initial design studies that were verified with benchmark experiments

  18. A contribution for the problematic of measurements of fast-neutron-energy spectrum in thermal reactor-systems

    International Nuclear Information System (INIS)

    Dederichs, H.

    1978-06-01

    The aims of this work are to check the experimental conditions for using of a 6 Li-semiconductor-spectrometer at thermal reactor-systems and to measure the neutron-energy-spectra at the critical experiment KAHTER comparing with the theory. Using the spectrometer at thermal-neutraon-experiments questions will be attended as resolution, statistic and selection of usable nuclear data. The nuclear data will be gauged by qualified measurements, the statistic will be estimated by simulated calculations and the resolution will be improved by using the Fredholm-equation in the calculations. The calculated spectra show a good agreement with the measured spectra. Only in the energy region of maximum distribution of fission-neutrons there are little difference. The measurements show the using of the spectrometer is recommended at systems with preponderant thermal neutron-spectra, although the resolution and statistic are optimized for the spectrometer by measurements at experiments with fast neutron-spectra. (orig.) 891 RW [de

  19. On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Jordan, K. A. [Paul Scherrer Institut, Villigen, 5232 (Switzerland)

    2011-07-01

    Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

  20. Detection of SNM by delayed gamma rays from induced fission

    International Nuclear Information System (INIS)

    Rennhofer, H.; Crochemore, J.-M.; Roesgen, E.; Pedersen, B.

    2011-01-01

    The Pulsed Neutron Interrogation Test Assembly (PUNITA) is an experimental device for research in NDA methods and field applicable instrumentation for nuclear safeguards and security applications. PUNITA incorporates a standard 14-MeV (D-T) pulsed neutron generator inside a large graphite mantle. The generator target is surrounded by a thick tungsten filter with the purpose to increase the neutron output and to tailor the neutron energy spectrum. In this configuration a sample may be exposed to a relatively high average thermal neutron flux of about (2.2±0.1)x10 3 s -1 cm -2 at only 10% of the maximum target neutron emission. The sample cavity is large enough to allow variation of the experimental setup including the fissile sample, neutron and gamma detectors, and shielding materials. The response from SNM samples of different fissile material content was investigated with various field-applicable scintillation gamma detectors such as the 3x2 in. LaBr 3 detector. Shielding in the form of tungsten and cadmium was applied to the detector to improve the signal to background ratio. Gamma and neutron shields surrounding the samples were also tested for the purpose of simulating clandestine conduct. The energy spectra of delayed gamma rays were recorded in the range 100 keV-9 MeV. In addition time spectra of delayed gamma rays in the range 3.3-8 MeV were recorded in the time period of 10 ms-120 s after the 14-MeV neutron burst. The goal of the experiment was to optimize the sample/detector configuration including the energy range and time period for SNM detection. The results show, for example, that a 170 g sample of depleted uranium can be detected with the given setup in less than 3 min of investigation. Samples of higher enrichment or higher mass are detected in much shorter time.

  1. Spectra of γ-rays from capture of 2 eV to 9 x 104 eV neutrons by 181Ta

    International Nuclear Information System (INIS)

    Stelts, M.L.

    Using new experimental techniques, the spectra of γ-rays from the capture of neutrons by 181 Ta were measured at the Livermore 100-MeV linac for neutrons from 2 eV to 9 x 10 4 eV with a (Ge(Li)-NaI) three-crystal spectrometer. Individual primary γ-ray lines were resolved to 1778-keV excitation in 182 Ta. Neutron resonances were resolved to 200-eV neutron energy. Data analysis techniques and codes were developed to extract positions and intensities of resolved transitions from the large data matrices accumulated in this experiment. Techniques were developed to unfold the unresolved γ-ray spectra using the simple response of the three-crystal spectrometer. The resolved transition data were used to place 110 states with spin and parity assignments in the 182 Ta level diagram below 1780-keV excitation. A set of 1240 E1 transition strengths were analyzed to extract 1.38 +- 0.11 degrees of freedom for the most likely chisquared fit to the distribution of widths. The E1 strength function was extracted for E/sub gamma/ = 4 to 6 MeV and compared with previous results. The γ-ray spectra for E/sub gamma/ = 1.5 to 6.1 MeV were unfolded for neutron energy groups between 20 and 9 x 10 4 eV. Below 5-MeV γ-ray energy no dependence of the spectral shape on neu []ron energy was observed. (30 figures, 4 tables) (auth)

  2. Neutron and gamma dose and spectra measurements on the Little Boy replica

    International Nuclear Information System (INIS)

    Hoots, S.; Wadsworth, D.

    1984-01-01

    The radiation-measurement team of the Weapons Engineering Division at Lawrence Livermore National Laboratory (LLNL) measured neutron and gamma dose and spectra on the Little Boy replica at Los Alamos National Laboratory (LANL) in April 1983. This assembly is a replica of the gun-type atomic bomb exploded over Hiroshima in 1945. These measurements support the National Academy of Sciences Program to reassess the radiation doses due to atomic bomb explosions in Japan. Specifically, the following types of information were important: neutron spectra as a function of geometry, gamma to neutron dose ratios out to 1.5 km, and neutron attenuation in the atmosphere. We measured neutron and gamma dose/fission from close-in to a kilometer out, and neutron and gamma spectra at 90 and 30 0 close-in. This paper describes these measurements and the results. 12 references, 13 figures, 5 tables

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  4. Effects of Hot-Spot Geometry on Backscattering and Down-Scattering Neutron Spectra

    Science.gov (United States)

    Mohamed, Z. L.; Mannion, O. M.; Forrest, C. J.; Knauer, J. P.; Anderson, K. S.; Radha, P. B.

    2017-10-01

    The measured neutron spectrum produced by a fusion experiment plays a key role in inferring observable quantities. One important observable is the areal density of an implosion, which is inferred by measuring the scattering of neutrons. This project seeks to use particle-transport simulations to model the effects of hot-spot geometry on backscattering and down-scattering neutron spectra along different lines of sight. Implosions similar to those conducted at the Laboratory of Laser Energetics are modeled by neutron transport through a DT plasma and a DT ice shell using the particle transport codes MCNP and IRIS. Effects of hot-spot geometry are obtained by ``detecting'' scattered neutrons along different lines of sight. This process is repeated for various hot-spot geometries representing known shape distortions between the hot spot and the shell. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  5. Reconstruction of neutron spectra using neural networks starting from the Bonner spheres spectrometric system

    International Nuclear Information System (INIS)

    Ortiz R, J.M.; Martinez B, M.R.; Arteaga A, T.; Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.

    2005-01-01

    The artificial neural networks (RN) have been used successfully to solve a wide variety of problems. However to determine an appropriate set of values of the structural parameters and of learning of these, it continues being even a difficult task. Contrary to previous works, here a set of neural networks is designed to reconstruct neutron spectra starting from the counting rates coming from the detectors of the Bonner spheres system, using a systematic and experimental strategy for the robust design of multilayer neural networks of the feed forward type of inverse propagation. The robust design is formulated as a design problem of Taguchi parameters. It was selected a set of 53 neutron spectra, compiled by the International Atomic Energy Agency, the counting rates were calculated that would take place in a Bonner spheres system, the set was arranged according to the wave form of those spectra. With these data and applying the Taguchi methodology to determine the best parameters of the network topology, it was trained and it proved the same one with the spectra. (Author)

  6. Neutron thermalization and spectra; Thermalisation et spectres de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Cadilhac, M; Soule, J L; Tretiakoff, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The investigation of the neutron spectra in thermal reactors is taking an increasing importance due to the role played in them by Plutonium. Whatever the absorption law, it has been remarked that the scattering law in the.moderator affects the spectrum only through certain overall properties. It would thus seem possible to develop a simplified representation of this effect which would lead to a clear understanding of the phenomena, reducing at the same time the volume of numerical calculations required.. The synthetic model employed by the authors presents the advantage of reducing the determination of the spectra in an homogeneous medium to the resolution of a second order differential equation, like the Wigner-Wilkins model (monoatomic gaseous hydrogen) and the generalized heavy gas model of J. Horowitz which, incidentally, are both special cases. The model is, on the other hand, sufficiently general to allow a correct treatment of the situations met with in practice and in particular the important case where the presence of Plutonium introduces absorption resonances at low energy. Actually, the chemical or crystalline bonds of the moderator are introduced into the proposed model through two energy functions. These functions have been adjusted for the usual moderators (Graphite heavy water, light water) by means of known theoretical scattering laws. In a heterogeneous medium, the most important factor is the mean spectrum in the fuel of one cell, the knowledge of which is allowed by a generalization of the Amouyal-Benoist-Horowitz method. The proposed model lends itself particularly well to such calculations and also allows the effects of re-thermalization (for instance when the cooling system and the moderator are at different temperatures) to be treated. Finally, some examples are given of practical applications: codes for spectra and effective cross sections computations (editing of tables), codes for the treatment of neutron balance in a lattice or for the

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  8. MONTE CARLO CALCULATION OF THE ENERGY RESPONSE OF THE NARF HURST-TYPE FAST- NEUTRON DOSIMETER

    Energy Technology Data Exchange (ETDEWEB)

    De Vries, T. W.

    1963-06-15

    The response function for the fast-neutron dosimeter was calculated by the Monte Carlo technique (Code K-52) and compared with a calculation based on the Bragg-Gray principle. The energy deposition spectra so obtained show that the response spectra become softer with increased incident neutron energy ahove 3 Mev. The K-52 calculated total res nu onse is more nearly constant with energy than the BraggGray response. The former increases 70 percent from 1 Mev to 14 Mev while the latter increases 135 percent over this energy range. (auth)

  9. Beta-delayed fission and neutron emission calculations for the actinide cosmochronometers

    International Nuclear Information System (INIS)

    Meyer, B.S.; Howard, W.M.; Mathews, G.J.; Takahashi, K.; Moeller, P.; Leander, G.A.

    1989-01-01

    The Gamow-Teller beta-strength distributions for 19 neutron-rich nuclei, including ten of interest for the production of the actinide cosmochronometers, are computed microscopically with a code that treats nuclear deformation explicitly. The strength distributions are then used to calculate the beta-delayed fission, neutron emission, and gamma deexcitation probabilities for these nuclei. Fission is treated both in the complete damping and WKB approximations for penetrabilities through the nuclear potential-energy surface. The resulting fission probabilities differ by factors of 2 to 3 or more from the results of previous calculations using microscopically computed beta-strength distributions around the region of greatest interest for production of the cosmochronometers. The indications are that a consistent treatment of nuclear deformation, fission barriers, and beta-strength functions is important in the calculation of delayed fission probabilities and the production of the actinide cosmochronometers. Since we show that the results are very sensitive to relatively small changes in model assumptions, large chronometric ages for the Galaxy based upon high beta-delayed fission probabilities derived from an inconsistent set of nuclear data calculations must be considered quite uncertain

  10. Calculations of the relative effectiveness of alanine for neutrons with energies up to 17.1 MeV

    International Nuclear Information System (INIS)

    Gerstenberg, H.M.; Coyne, J.J.

    1990-01-01

    The relative effectiveness (RE) of alanine has been calculated for neutrons using the RE of alanine for charged particles. The neutrons interact with one or more of the elements (hydrogen, carbon, nitrogen and oxygen) that compose the alanine. These interactions produce spectra of secondary charged particles consisting of ions of H, D, He, Be, B, C, N and O. From a combination of the calculated secondary charged particle spectra generated by the slowing down neutrons, and the calculated RE of the ions produced, a RE for the neutrons can be obtained. In addition, lineal energy spectra were determined for neutrons with energies up to 17.1 MeV interacting with alanine. An analytical code was used to calculate these spectra for a 1 μm diameter alanine cell surrounded by an alanine medium. For comparison, similar calculations were made for muscle tissue. Finally, the calculated differential RE was folded with dose distributions to obtain RE-weighted distributions for alanine. (author)

  11. Design of an artificial neural network, with the topology oriented to the reconstruction of neutron spectra

    International Nuclear Information System (INIS)

    Arteaga A, T.; Ortiz R, J.M.; Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.; Mercado S, G.A.

    2006-01-01

    People that live in high places respect to the sea level, in latitudes far from the equator or that they travel by plane, they are exposed to atmospheres of high radiation generated by the cosmic rays. Another atmosphere with radiation is the medical equipment, particle accelerators and nuclear reactors. The evaluation of the biological risk for neutron radiation requires an appropriate and sure dosimetry. A commonly used system is the Bonner Sphere Spectrometer (EEB) with the purpose of reconstructing the spectrum that is important because the equivalent dose for neutrons depends strongly on its energy. The count rates obtained in each sphere are treated, in most of the cases, for iterative methods, Monte Carlo or Maximum Entropy. Each one of them has difficulties that it motivates to the development of complementary procedures. Recently it has been used Artificial Neural Networks, ANN) and not yet conclusive results have been obtained. In this work it was designed an ANN to obtain the neutron energy spectrum neutrons starting from the counting rate of count of an EEB. The ANN was trained with 129 reference spectra obtained of the IAEA (1990, 2001), 24 were built as defined energy, including isotopic sources of neutrons of reference and operational, of accelerators, reactors, mathematical functions, and of defined energy with several peaks. The spectrum was transformed from lethargy units to energy and were reaccommodated in 31 energies using the Monte Carlo code 4C. The reaccommodated spectra and the response matrix UTA4 were used to calculate the prospective count rates in the EEB. These rates were used as entrance and its respective spectrum was used as output during the net training. The net design is Retropropagation type with 5 layers of 7, 140, 140, 140 and 31 neurons, transfer function logsig, tansig, logsig, logsig, logsig respectively. Training algorithm, traingdx. After the training, the net was proven with a group of training spectra and others that

  12. 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

  13. Compilation of neutron flux density spectra and reaction rates in different neutron fields. V.3

    International Nuclear Information System (INIS)

    Ertek, C.

    1980-04-01

    Upon the recommendation of the International Working Group of Reactor Radiation Measurements (IWGRRM) a compilation of documents containing neutron flux density spectra and the reaction rates obtained by activiation and fission foils in different neutron fields is presented

  14. Study and building of a detection array for delayed neutrons: TONNERRE; Etude et realisation d`un ensemble de detection pour neutrons retardes: TONNERRE

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Thierry [Lab. de Physique Corpusculaire, Caen Univ., 14 - Caen (France)

    1998-11-09

    This work has been undertaken within a French-Romanian collaboration in order to build a high efficiency detector array for delayed neutrons: barrel-shaped TONNERRE. Some neutron-rich nuclei decay through 1, 2 or 3 neutron emission after {beta}{sup -} decay. More exotic nuclei will be produced by SPIRAL at GANIL. An array with high efficiency and good resolution is then required. Thirty two BC400 plastic scintillators (160 x 20 x 4 cm{sup 3}) allow us to get the time of flight neutron spectra. They are bent for uniform flight path and viewed by a photomultiplier tube at both ends. Simulations have allowed to establish scintillator size and to minimize light attenuation. Intrinsic efficiency and crosstalk have been measured with {sup 252}Cf and compared to GEANT. 1 to 5 MeV neutrons are detected with good timing and position properties. Other counters will be built for neutrons from 300 keV to 1 MeV. Planned to run at several particle accelerators (GANIL, CERN, and others), TONNERRE is modular and many geometries are possible. (author) 48 refs., 78 figs., 20 tabs.

  15. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

    International Nuclear Information System (INIS)

    Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

    2010-01-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

  16. The study of prompt neutron spectra of 238U fission induced by fast neutron

    International Nuclear Information System (INIS)

    Li Anli; Bai Xixiang; Wang Yufeng; Wang Xiaozhong; Men Jiangchen; Huang Shengnian

    1990-01-01

    The measurements of prompt neutron time-of-flight spectra of U fission induced by 11 MeV neutrons were carried out at HI-13 Tandem Van de Graaff Accelerator Laboratory in 1989. The block diagram of the electronics is shown. A fission neutron TOF spectrum for the sixth section of the fission plates and the left detector at low bias is given. The data accumulation time is 60 h

  17. Neutron and gamma-ray spectra of {sup 239}PuBe and {sup 241}AmBe

    Energy Technology Data Exchange (ETDEWEB)

    Vega-Carrillo, H.R. E-mail: rvega@cantera.reduaz.mx; Manzanares-Acuna, Eduardo; Becerra-Ferreiro, A.M.; Carrillo-Nunez, Aureliano

    2002-08-01

    Neutron and gamma-ray spectra of {sup 239}PuBe and {sup 241}AmBe were measured and their dosimetric features were calculated. Neutron spectra were measured using a multisphere neutron spectrometer with a {sup 6}LiI(Eu) scintillator. The {sup 239}PuBe neutron spectrum was measured in an open environment, while the {sup 241}AmBe neutron spectrum was measured in a closed environment. Gamma-ray spectra were measured using a NaI(Tl) scintillator using the same experimental conditions for both sources. The effect of measuring conditions for the {sup 241}AmBe neutron spectrum indicates the presence of epithermal and thermal neutrons. The low-resolution neutron spectra obtained with the multisphere spectrometer allows one to calculate the dosimetric features of neutron sources. At 100 cm both sources produce approximately the same count rate as that of the 4.4 MeV gamma-ray per unit of alpha emitter activity.

  18. Measurement and analysis of angular neutron spectra in a manganese pile

    International Nuclear Information System (INIS)

    Selvi, S.; Hayashi, S.A.; Kimura, I.; Kobayashi, K.; Yamamoto, S.; Mori, T.; Nishihara, H.; Kanazawa, S.; Nakagawa, M.

    1984-01-01

    The energy and angular distribution of neutrons in a Mn pile were measured by the linac time-of-flight method. A cylindrical Pb target for the production of photoneutrons was placed at the center of the pile. The experimental results were compared with the theoretical calculations using the group constants from the nuclear data files, JENDL-2 and ENDF/B-IV. Good agreement can be seen in the general shapes between calculated and measured angular spectra in three decades of energy range form a few keV to a few MeV. As far as can be concluded from the intercomparison, the neutron cross section data for Mn in ENDF/B-IV may be applicable to reactor design: however, several improvements for its resonance parameters can be recommended. A little more improvements are recommended for that in JENDL-2 from this intercomparison. (orig.) [de

  19. Measurement and analysis of double-differential neutron emission spectra in (P,N) and (α,N) reactions

    International Nuclear Information System (INIS)

    Okamoto, K.; Mehta, M.K.

    1988-05-01

    The second IAEA Research Co-ordination Meeting on Measurement and Analysis of Double-Differential Neutron Emission Spectra in (p,n) and (α,n) Reactions was convened by the IAEA Nuclear Data Section at the IAEA Headquarters in Vienna during 8-10 February, 1988. The main objectives of the Co-ordinated Research Project for which this meeting was held are (i) to extract systematic information about nuclear level densities as a function of excitation energy by analysing the neutron emission spectra from (p,n) and (α,n) reactions on properly selected targets and bombarding energy range, and (ii) to parametrize this information into appropriate phenomenological models to enable reliable extrapolation for general use of level density information in basic and applied nuclear physics related problems. Detailed conclusions and recommendations, together with a summary of the programme during 1988/1989 are attached in the Appendices

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  1. Analysis of the neutron energy spectra from the sup(208)Pb (p,n) sup(208)Bi reaction at Esub(p)=200 MeV

    International Nuclear Information System (INIS)

    Ershov, S.N.; Fayans, S.A.; Gareev, F.A.; Pyatov, N.I.

    1986-01-01

    Microscopic calculation of the small-angle neutron energy spectra from the 208 Pb (p, n) 208 Bi reaction at Esub(p)=200 MeV are presented. It is shown that the distorted-wave impulse approximation and the microscopic theory of finite Fermi systems can be employed for describing the low-energy excitation region 0 <= Q <= 30 MeV with small momentum transfers. A quantitative estimate is obtained for the local charge of quasiparticles esub(q)[σtau]=0.8 that characterizes the quenching of the integral strength of spin-flip low-energy transitions and the relevant effects are discussed

  2. Neutron dose measurements with the GSI ball at high energy accelerators

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-09-01

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

  4. 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.)

  5. Evaluation of response function of moderating-type neutron detector and application to environmental neutron measurement

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakamura, Takashi; Iwai, Satoshi; Katsuki, Shinji; Kamata, Masashi.

    1983-08-01

    The energy-dependent response function of a multi-cylinder moderating-type BF 3 counter, so-called Bonner counter, was calculated by the time-dependent multi-group Monte Carlo code, TMMCR. The calculated response function was evaluated experimentally for neutron energy below about 50 keV down to epithermal energy by the time-of-flight method combining with a large lead pile at the Nuclear Engineering Research Laboratory, University of Tokyo and also above 50 keV by using the monoenergetic neutron standard field a t the Electrotechnical Laboratory. The time delay in the polyethylene moderator of the Bonner counter due to multiple collisions with hydrogen was analyzed by the TMMCR code and used for the time-spectrum analysis of the time-of-flight measurement. The response function obtained by these two experiments showed good agreement with the calculated results. This Bonner counter having a response function evaluated from thermal to MeV energy range was used for spectrometry and dosimetry of environmental neutrons around some nuclear facilities. The neutron spectra and dose measured in the environment around a 252 Cf fission source, fast neutron source reactor and electron synchrotron were all in good agreement with the calculated results and the measured results with other neutron detectors. (author)

  6. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

    International Nuclear Information System (INIS)

    Schickler, R.A.; Marcum, W.R.; Reese, S.R.

    2013-01-01

    Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

  7. Energy-resolved fast neutron resonance radiography at CSNS

    Science.gov (United States)

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

    2018-05-01

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

  8. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Dickens, J.K.; Hill, N.W.; Hou, F.S.; McConnell, J.W.; Spencer, R.R.; Tsang, F.Y.

    1985-08-01

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report

  9. Integral test of neutron cross section data for future reactor materials through measurement and analysis of neutron spectra

    International Nuclear Information System (INIS)

    Mori, Takamasa

    1985-05-01

    In order to assess the cross section data for future reactor materials, such as molybdenum, niobium, titanium, lithium and fluorine, the angular neutron spectra in test piles of these materials or their chemical compounds have been measured in the energy range from a few keV to a few MeV by the linac time-of-flight method. The results have been compared with those theoretically calculated from the evaluated cross section data in such as JENDL-2 (or JENDL-1, JENDL-3PR1) and ENDF/B-IV. For both of molybdenum and niobium, it has been found that the energy distribution of inelastically scattered neutrons plays an important role in the analysis, and the JENDL library gives better predictions of spectrum shapes than ENDF/B-IV for both cases. In the case of niobium, however, it appears that the values of inelastic scattering cross section in JENDL-2 are too small around 2 MeV. It has been also found for niobium that the cross section data below 100 keV in ENDF/B-IV are inadequate. In a titanium pile, a discrepancy between the measured spectrum and the calculated one from ENDF/B-IV has been found in the energy range from about 60 keV to a few 100 keV. In order to investigate the cause of this discrepancy, the total cross sections for titanium have been measured by the transmission method. In the case of lithium, the discrepancy between the measured and calculated spectra is considerably reduced by adopting the angular distribution for 7 Li from ENDF/B-IV above about 500 keV. In the case of fluorine, spatial distributions of neutrons and X-rays have been also measured in both piles by the activation method to estimate the influence of photoneutrons generated in the sample material on the neutron distribution, and it has been found that their influence below 1 MeV is not so large as is necessary to be taken into account for the present assessment. (J.P.N)

  10. ZZ DOSCROS, Neutron Cross-Section Library for Spectra Unfolding and Integral Parameter Evaluation

    International Nuclear Information System (INIS)

    Zijp, Willem L.; Nolthenius, Henk J.; Rieffe, Henk Ch.

    1987-01-01

    1 - Description of problem or function: Format: SAND-II; Number of groups: 640 fine group cross section values; Nuclides: Li, B, F, Na, Mg, Al, S, Sc, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Br, Nb, Mo, Rh, Pd, Ag, In, Sb, I, Cs, La, Eu, Sm, Dy, Lu, Ta, W, Re, Au, Th, U, Np, Pu. Origin: ENDF/B-V mainly, ENDF/B-IV, INDL/V. This library forms in combination with the DAMSIG81 library a convenient source of evaluated energy dependent cross section sets which may be used in the determination of neutron spectra by means of adjustment (or unfolding) procedures or which can be used for the determination of integral parameters (such as damage-to-activation ratio) useful in characterising the neutron spectra. The energy dependent fine group cross section data are presented in a 640 group structure of the SAND-II type. This group structure has 45 energy groups per energy decade below 1 MeV and a group width of 100 KeV above 1 MeV. The total energy span of this group structure is from 10 -10 MeV to 20 MeV. The library has the SAND-II format, which implies that a special part of the library has to contain cover cross section data sets. These cross section data sets are required in the SAND-II program for taking into account the influence of special detector surroundings which may be used during an irradiation. 2 - Method of solution: The selection of the reactions from the evaluated nuclear data libraries was determined by various properties of the reactions for neutron metrology. For this reason all the well- known reactions of the ENDF/B-V dosimetry file are included but these data are supplemented with cross section sets for less well known metrology reactions which may become of interest

  11. Neutron spectrum unfolding using neural networks

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  12. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  13. Characteristic Investigation of Unfolded Neutron Spectra with Different Priori Information and Gamma Radiation Interference

    International Nuclear Information System (INIS)

    Kim, Bong Hwan

    2006-01-01

    Neutron field spectrometry using multi spheres such as Bonner Spheres (BS) has been almost essential in radiation protection dosimetry for a long time at workplace in spite of poor energy resolution because it is not asking the fine energy resolution but requiring easy operation and measurement performance over a wide range of energy interested. KAERI has developed and used extended BS system based on a LiI(Eu) scintillator as the representative neutron spectrometry system for workplace monitoring as well as for the quantification of neutron calibration fields such as those recommended by ISO 8529. Major topics in using BS are how close the unfolded spectra is the real one and to minimize the interference of gamma radiation in neutron/gamma mixed fields in case of active instrument such as a BS with a LiI(Eu) scintillator. The former is related with choosing a priori information when unfolding the measured data and the latter is depend on how to discriminate it in intense gamma radiation fields. Influence of a priori information in unfolding and effect of counting loss due to pile-up of signals for the KAERI BS system were investigated analyzing the spectral measurement results of Scattered Neutron Calibration Fields (SNCF)

  14. Proton energy spectra during ground level enhancements as measured by EPHIN aboard SOHO

    Energy Technology Data Exchange (ETDEWEB)

    Heber, Bernd; Kuehl, Patrick; Klassen, Andreas; Dresing, Nina [Christian-Albrechts-Universitaet zu Kiel, 24118 Kiel (Germany); Gomez-Herrero, Raul [Universidad de Alcala (Spain)

    2016-07-01

    Ground Level Enhancements (GLEs) are solar energetic particle (SEP) events that are recorded by ground-based instrumentation. The energy of the particles is so high that they produce secondary particles in the Earth's atmosphere, i.e. protons and neutrons, which are detected as sudden increases in cosmic ray intensities measured by e.g. neutron monitors. Since the launch of SOHO in December 1995 the neutron monitor network recorded 16 GLEs. The Electron Proton Helium INstrument on board SOHO has been designed to measure protons and helium up to 53 MeV/nucleon as well as electrons up to 8.3 MeV. Above these energies, particles penetrate all detector elements and thus, a separation between different particle species becomes more complicated. Recently we developed a method that allows deriving the energy spectrum for penetrating protons up to more than 1 GeV. In this contribution we present the proton energy spectra and time profiles of above mentioned GLEs and compare them to previous measurements. Although there are differences of up to a factor two the overall shape of the energy spectra agree surprisingly well. Thus it has been demonstrated that EPHIN measurements are a valuable tool for understanding GLE.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  16. Influence of delayed neutron parameter calculation accuracy on results of modeled WWER scram experiments

    International Nuclear Information System (INIS)

    Artemov, V.G.; Gusev, V.I.; Zinatullin, R.E.; Karpov, A.S.

    2007-01-01

    Using modeled WWER cram rod drop experiments, performed at the Rostov NPP, as an example, the influence of delayed neutron parameters on the modeling results was investigated. The delayed neutron parameter values were taken from both domestic and foreign nuclear databases. Numerical modeling was carried out on the basis of SAPFIR 9 5andWWERrogram package. Parameters of delayed neutrons were acquired from ENDF/B-VI and BNAB-78 validated data files. It was demonstrated that using delay fraction data from different databases in reactivity meters led to significantly different reactivity results. Based on the results of numerically modeled experiments, delayed neutron parameters providing the best agreement between calculated and measured data were selected and recommended for use in reactor calculations (Authors)

  17. A code for leakage neutron spectra through thick shields

    International Nuclear Information System (INIS)

    Nagarajan, P.S.; Sethulakshmi, P.; Raghavendran, C.P.

    1975-01-01

    An exponential transform Monte Carlo code has been developed for deep penetration of neutrons and the results of leakage neutron spectra of this code have been compared with those of a basic Monte Carlo code for small thickness. The development of the code and optimisation of certain transform parameters are discussed and results are presented for a few thick shields of concrete and water in the context of neutron monitoring in the environs of accelerator and reactor shields. (author)

  18. Results of neutron dose measurements at the Rossendorf research reactors taking the actual neutron spectra into account

    International Nuclear Information System (INIS)

    Rimpler, A.; Kneschke, H.

    1985-01-01

    Based on a systematic evaluation of area dose studies at the beginning of the seventies, no individual routine neutron monitoring has been performed at the Rossendorf research reactors. To check this decision, a limited number of persons has been monitored with solid-state nuclear track detectors for several years. The dosemeters were calibrated on the basis of neutron spectra determined at the working places by means of the Bonner sphere method. Intermediate neutrons with a 1/E/sup α/ Fermi distribution were dominating. The fraction of fast neutrons was practically negligible. The obtained spectra, radiation, field quantities and results of individual dose measurements are presented. The dosemeter most appropriate for such neutron fields would be a 12-inch Bonner sphere rem counter. As the mean annual neutron exposure of research workers at the reactor amounted to only 2% of the maximum permissible dose, individual routine monitoring will, also in the future, not be neccessary. (author)

  19. THE PRODUCTION OF LOW-ENERGY NEUTRONS IN SOLAR FLARES AND THE IMPORTANCE OF THEIR DETECTION IN THE INNER HELIOSPHERE

    International Nuclear Information System (INIS)

    Murphy, R. J.; Kozlovsky, B.; Share, G. H.

    2012-01-01

    Neutron detectors on spacecraft in the inner heliosphere can observe the low-energy ( ion –1 ) most important for producing low-energy neutrons from these reactions. We calculate escaping-neutron spectra and neutron-capture line yields from ions propagating in a magnetic loop with various kinetic-energy spectra. This study provides the basis for planning inner-heliospheric missions having a low-energy neutron detector. The MESSENGER spacecraft orbiting Mercury has such a detector. We conclude that a full understanding of ion acceleration, transport, and interaction at the Sun requires observation of both neutrons and gamma rays with detectors of comparable sensitivity. We find that the neutron-capture line fluence at 1 AU is comparable to the 1-10 MeV neutron fluence at 0.5 AU, and therefore as effective for revealing low-energy ion acceleration. However, as the distance from the Sun to the neutron detector decreases, the tremendous increase of the low-energy neutron flux allows exploration of ion acceleration in weak flares not previously observable and may reveal acceleration at other sites not previously detected where low-energy neutrons could be the only high-energy signature of ion acceleration. Also, a measurement of the low-energy neutron spectrum will provide important information about the accelerated-ion spectrum that is not available from the capture line fluence measurement alone.

  20. Synovectomy by Neutron capture; Sinovectomia por captura de neutrones

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  1. Measurements and calculations of neutron spectra and neutron dose distribution in human phantoms

    International Nuclear Information System (INIS)

    Palfalvi, J.

    1984-11-01

    The measurement and calculation of the radiation field around and in a phantom, with regard to the neutron component and the contaminating gamma radiation, are essential for radiation protection and radiotherapy purposes. The final report includes the development of the simple detector system, automized detector measuring facilities and a computerized evaluating system. The results of the depth dose and neutron spectra experiments and calculations in a human phantom are given

  2. 235U Determination using In-Beam Delayed Neutron Counting Technique at the NRU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, M. T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bentoumi, G. [Canadian Nuclear Labs., Chalk River, ON (Canada); Corcoran, E. C. [Royal Military College of Canada, Kingston, ON (United States); Dimayuga, I. [Canadian Nuclear Labs., Chalk River, ON (Canada); Kelly, D. G. [Royal Military College of Canada, Kingston, ON (United States); Li, L. [Canadian Nuclear Labs., Chalk River, ON (Canada); Sur, B. [Canadian Nuclear Labs., Chalk River, ON (Canada); Rogge, R. B. [Canadian Nuclear Labs., Chalk River, ON (Canada)

    2015-11-17

    This paper describes a collaborative effort that saw the Royal Military College of Canada (RMC)’s delayed neutron and gamma counting apparatus transported to Canadian Nuclear Laboratories (CNL) for use in the neutron beamline at the National Research Universal (NRU) reactor. Samples containing mg quantities of fissile material were re-interrogated, and their delayed neutron emissions measured. This collaboration offers significant advantages to previous delayed neutron research at both CNL and RMC. This paper details the determination of 235U content in enriched uranium via the assay of in-beam delayed neutron magnitudes and temporal behavior. 235U mass was determined with an average absolute error of ± 2.7 %. This error is lower than that obtained at RMCC for the assay of 235U content in aqueous solutions (3.6 %) using delayed neutron counting. Delayed neutron counting has been demonstrated to be a rapid, accurate, and precise method for special nuclear material detection and identification.

  3. Theoretical model application to the evaluation of fission neutron data up to 20 MeV incidence energy

    International Nuclear Information System (INIS)

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

    1990-01-01

    A complex statistical theory of fission neutron emission combined with a phenomenological fission model has been used to calculate fission neutron data for 238 U. Obtained neutron multiplicities and energy spectra as well as average fragment energies for incidence energies from threshold to 20 MeV (including multiple-chance fission) are compared with traditional data representations. (author). 19 refs, 6 figs

  4. 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)

  5. Neutron irradiation effects in fusion or spallation structural materials: Some recent insights related to neutron spectra

    International Nuclear Information System (INIS)

    Garner, F.A.; Greenwood, L.R.

    1998-01-01

    A review is presented of recent insights on the role of transmutation in the development of radiation-induced changes in dimension or radiation-induced changes in physical or mechanical properties. It is shown that, in some materials and some neutron spectra, transmutation can significantly affect or even dominate a given property change process. When the process under study is also sensitive to displacement rate, and especially if it involves radiation-induced segregation and precipitation, it becomes much more difficult to separate the transmutation and displacement rate dependencies. This complicates the application of data derived from 'surrogate' spectra to predictions in other flux-spectra environments. It is also shown in this paper that one must be sensitive to the impact of previously -ignored 'small' variations in neutron spectra within a given reactor. In some materials these small variations have major consequences. (author)

  6. Study of spectral response of a neutron filter. Design of a method to adjust spectra

    International Nuclear Information System (INIS)

    Colomb-Dolci, F.

    1999-02-01

    The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (author)

  7. Measurement of neutron spectra generated from bombardment of 4 to 24 MeV protons on a thick 9Be target and estimation of neutron yields

    International Nuclear Information System (INIS)

    Paul, Sabyasachi; Sahoo, G. S.; Tripathy, S. P.; Sunil, C.; Bandyopadhyay, T.; Sharma, S. C.; Ramjilal,; Ninawe, N. G.; Gupta, A. K.

    2014-01-01

    A systematic study on the measurement of neutron spectra emitted from the interaction of protons of various energies with a thick beryllium target has been carried out. The measurements were carried out in the forward direction (at 0° with respect to the direction of protons) using CR-39 detectors. The doses were estimated using the in-house image analyzing program autoTRAK-n, which works on the principle of luminosity variation in and around the track boundaries. A total of six different proton energies starting from 4 MeV to 24 MeV with an energy gap of 4 MeV were chosen for the study of the neutron yields and the estimation of doses. Nearly, 92% of the recoil tracks developed after chemical etching were circular in nature, but the size distributions of the recoil tracks were not found to be linearly dependent on the projectile energy. The neutron yield and dose values were found to be increasing linearly with increasing projectile energies. The response of CR-39 detector was also investigated at different beam currents at two different proton energies. A linear increase of neutron yield with beam current was observed

  8. A suggestion for specification of the neutron energy scale in measurements, analyses and evaluations of differential reaction cross-section data

    International Nuclear Information System (INIS)

    Smith, D.L.

    1992-01-01

    Neutron energy distributions (spectra) in differential neutron cross section measurements are represented conventionally by giving the energies (not well defined) and full widths at half maximum of the spectra. In most modern experiments, it is possible for investigators to provide a much more thorough representation of neutron spectra. However, reporting such information would entail more numerical detail than appears practical to document. Therefore, it is suggested that such spectra be represented in publications and files of experimental data by providing the following four moments of each distribution: mean value (energy), standard deviation (resolution), skewness (asymmetry), and kurtosis (sharpness). Implementation of such a standard for data reporting would offer evaluators a much more complete and rational basis for comparing reported values and performing evaluations than previously possible. Some examples are provided to illustrate the concepts

  9. Calculation of the pulsed Feynman- and Rossi-alpha formulae with delayed neutrons

    International Nuclear Information System (INIS)

    Kitamura, Y.; Pazsit, I.; Wright, J.; Yamamoto, A.; Yamane, Y.

    2005-01-01

    In previous works, the authors have developed an effective solution technique for calculating the pulsed Feynman- and Rossi-alpha formulae. Through derivation of these formulae, it was shown that the technique can easily handle various pulse shapes of the pulsed neutron source. Furthermore, it was also shown that both the deterministic (i.e., synchronizing with the pulsing of neutron source) and stochastic (non-synchronizing) Feynman-alpha formulae can be obtained with this solution technique. However, for mathematical simplicity and the sake of insight, the formal derivation was performed in a model without delayed neutrons. In this paper, to demonstrate the robustness of the technique, the pulsed Feynman- and Rossi-alpha formulae were re-derived by taking one group of delayed neutrons into account. The results show that the advantages of this technique are retained even by inclusion of the delayed neutrons. Compact explicit formulae are derived for the Feynman- and Rossi-alpha methods for various pulse shapes and pulsing methods

  10. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke; Favalli, Andrea; Hunt, Alan W.; Reedy, Edward T.E.; Seipel, Heather

    2015-01-01

    High-energy, beta-delayed gamma-ray spectroscopy is a potential, non-destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-energy delayed gamma rays from 235 U, 239 Pu, and 241 Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241 Pu, a significant fissile constituent in spent fuel, was measured and compared to 239 Pu. The 241 Pu/ 239 Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-ray emission was developed and demonstrated on a limited 235 U data set. De-convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an

  11. Neutron spectrometry with artificial neural networks

    International Nuclear Information System (INIS)

    Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.; Rodriguez, J.M.; Mercado S, G.A.; Iniguez de la Torre Bayo, M.P.; Barquero, R.; Arteaga A, T.

    2005-01-01

    An artificial neural network has been designed to obtain the neutron spectra from the Bonner spheres spectrometer's count rates. The neural network was trained using 129 neutron spectra. These include isotopic neutron sources; reference and operational spectra from accelerators and nuclear reactors, spectra from mathematical functions as well as few energy groups and monoenergetic spectra. The spectra were transformed from lethargy to energy distribution and were re-bin ned to 31 energy groups using the MCNP 4C code. Re-binned spectra and UTA4 response matrix were used to calculate the expected count rates in Bonner spheres spectrometer. These count rates were used as input and the respective spectrum was used as output during neural network training. After training the network was tested with the Bonner spheres count rates produced by a set of neutron spectra. This set contains data used during network training as well as data not used. Training and testing was carried out in the Mat lab program. To verify the network unfolding performance the original and unfolded spectra were compared using the χ 2 -test and the total fluence ratios. The use of Artificial Neural Networks to unfold neutron spectra in neutron spectrometry is an alternative procedure that overcomes the drawbacks associated in this ill-conditioned problem. (Author)

  12. Study on keV-neutron capture cross sections and capture γ-ray spectra of 117,119Sn

    International Nuclear Information System (INIS)

    Nishiyama, J.; Igashira, M.; Ohsaki, T.; Kim, G.N.; Chung, W.C.; Ro, T.I.

    2006-01-01

    The capture cross sections and capture γ-ray spectra of 117,119 Sn were measured in an incident neutron energy region from 10 to 100 keV and at 570 keV, using a 1.5-ns pulsed neutron source by the 7 Li(p,n) 7 Be reaction and a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross sections of 117,119 Sn were obtained with the error of about 5% by using the standard capture cross sections of 197 Au. The present cross sections were compared with previous experimental data and the evaluated values in JENDL-3.3 and ENDF/B-VI. The capture γ-ray spectra of 117,119 Sn were derived by unfolding the observed capture γ-ray pulse-height spectra. The calculations of capture cross sections and capture γ-ray spectra of 117,119 Sn were performed with the EMPIRE-II code. The calculated results were compared with the present experimental ones. (author)

  13. Uncertainties related to numerical methods for neutron spectra unfolding

    International Nuclear Information System (INIS)

    Glodic, S.; Ninkovic, M.; Adarougi, N.A.

    1987-10-01

    One of the often used techniques for neutron detection in radiation protection utilities is the Bonner multisphere spectrometer. Besides its advantages and universal applicability for evaluating integral parameters of neutron fields in health physics practices, the outstanding problems of the method are data analysis and the accuracy of the results. This paper briefly discusses some numerical problems related to neutron spectra unfolding, such as uncertainty of the response matrix as a source of error, and the possibility of real time data reduction using spectrometers. (author)

  14. Comparison of (alpha, n) thick-target neutron yields and spectra from ORIGEN-S and SOURCES

    International Nuclear Information System (INIS)

    Brown, T.H.; Wilson, W.B.; Perry, R.T.; Charlton, W.S.

    1998-01-01

    Both ORIGEN-S and SOURCES generate thick-target neutron yields and energy spectra from (α,n) reactions in homogeneous materials. SOURCES calculates yield and spectra for any material containing α-emitting and (α,n) target elements by simulating reaction physics, using α-emission energy spectra, elemental stopping cross sections, (α,n) cross sections for target nuclei, and branching fractions to produce-nuclide energy levels. This methodology results in accurate yield and spectra. ORIGEN-S has two options for calculating yields and spectra. The UO 2 option (default) estimates yields and spectra assuming the input α-emitters to be infinitely dilute in UO 2 . The borosilicate-glass option estimates yields from the total input material composition and generates spectra purportedly representative of spectra generated by 238 Pu, 241 Am, 242 Cm, and 244 Cm infinitely dilute in borosilicate glass, even if none of these four α-emitters are present in the input material composition. Because yields from the borosilicate-glass option in ORIGEN-S are based on entire input material composition and are reasonably accurate, the same is often assumed to be true for spectra. The input/output functionality of the borosilicate-glass option, along with ambiguity in ORIGEN-S documentation, gives the incorrect impression that spectra representative of input compositions are generated. This impression is reinforced by wide usage of the SCALE code system and its ORIGEN-S module and their sponsorship by the US Nuclear Regulatory Commission

  15. Neutron spectrometry using artificial neural networks

    International Nuclear Information System (INIS)

    Vega-Carrillo, Hector Rene; Martin Hernandez-Davila, Victor; Manzanares-Acuna, Eduardo; Mercado Sanchez, Gema A.; Pilar Iniguez de la Torre, Maria; Barquero, Raquel; Palacios, Francisco; Mendez Villafane, Roberto; Arteaga Arteaga, Tarcicio; Manuel Ortiz Rodriguez, Jose

    2006-01-01

    An artificial neural network has been designed to obtain neutron spectra from Bonner spheres spectrometer count rates. The neural network was trained using 129 neutron spectra. These include spectra from isotopic neutron sources; reference and operational spectra from accelerators and nuclear reactors, spectra based on mathematical functions as well as few energy groups and monoenergetic spectra. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. The re-binned spectra and the UTA4 response matrix were used to calculate the expected count rates in Bonner spheres spectrometer. These count rates were used as input and their respective spectra were used as output during the neural network training. After training, the network was tested with the Bonner spheres count rates produced by folding a set of neutron spectra with the response matrix. This set contains data used during network training as well as data not used. Training and testing was carried out using the Matlab ( R) program. To verify the network unfolding performance, the original and unfolded spectra were compared using the root mean square error. The use of artificial neural networks to unfold neutron spectra in neutron spectrometry is an alternative procedure that overcomes the drawbacks associated with this ill-conditioned problem

  16. Influence of cross-section structure on unfolded neutron spectra

    International Nuclear Information System (INIS)

    Ertek, C.; Vlasov, M.F.; Cross, B.; Smith, P.M.

    1979-01-01

    The influence of cross-section structure on neutron spectra unfolded by multiple foil activation technique, SAND-II case, has been studied. For three reactions with evident structure in neutron cross-section above threshold: 27Al(n,α)24Na, 31P(n,p)31Si and 32S(n,p)32P, two remarkably different sets of evaluated data were selected from the available evaluations; one set of data was ''smooth'', the structure having been averaged over by a smooth curve; the other set was ''sharp'' with structure given in detail. These data were used in unfolding procedure together with other reactions, the same in both cases (as well as input spectra and measured reaction rates). It was found that during unfolding calculations less iteration steps were needed to unfold the neutron flux spectrum with the set of ''sharp'' data. In case of ''smooth'' data it was difficult to obtain an agreement between measured and calculated activity values even by increasing the number of iteration steps. Contrary to expectations, considerable deformation of unfolded neutron flux spectrum has been observed in the case of the ''smooth'' data set. (author)

  17. The Dynamic Method for Time-of-Flight Measurement of Thermal Neutron Spectra from Pulsed Sources

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Tulaev, A.B.; Bobrakov, V.F.

    1994-01-01

    The time-of-flight method for a measurement of thermal neutron spectra in the pulsed neutron sources with high efficiency of neutron registration, more than 10 5 times higher in comparison with traditional one, is described. The main problems connected with the electric current technique for time-of-flight spectra measurement are examined. The methodical errors, problems of a special neutron detector design and other questions are discussed. Some experimental results, spectra from surfaces of the water and solid methane moderators, obtained in the pulsed reactor IBR-2 (Dubna, Russia) are presented. 4 refs., 5 figs

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  19. Calibration of the delayed-gamma neutron activation facility

    International Nuclear Information System (INIS)

    Ma, R.; Zhao, X.; Rarback, H.M.; Yasumura, S.; Dilmanian, F.A.; Moore, R.I.; Lo Monte, A.F.; Vodopia, K.A.; Liu, H.B.; Economos, C.D.; Nelson, M.E.; Aloia, J.F.; Vaswani, A.N.; Weber, D.A.; Pierson, R.N. Jr.; Joel, D.D.

    1996-01-01

    The delayed-gamma neutron activation facility at Brookhaven National Laboratory was originally calibrated using an anthropomorphic hollow phantom filled with solutions containing predetermined amounts of Ca. However, 99% of the total Ca in the human body is not homogeneously distributed but contained within the skeleton. Recently, an artificial skeleton was designed, constructed, and placed in a bottle phantom to better represent the Ca distribution in the human body. Neutron activation measurements of an anthropomorphic and a bottle (with no skeleton) phantom demonstrate that the difference in size and shape between the two phantoms changes the total body calcium results by less than 1%. To test the artificial skeleton, two small polyethylene jerry-can phantoms were made, one with a femur from a cadaver and one with an artificial bone in exactly the same geometry. The femur was ashed following the neutron activation measurements for chemical analysis of Ca. Results indicate that the artificial bone closely simulates the real bone in neutron activation analysis and provides accurate calibration for Ca measurements. Therefore, the calibration of the delayed-gamma neutron activation system is now based on the new bottle phantom containing an artificial skeleton. This change has improved the accuracy of measurement for total body calcium. Also, the simple geometry of this phantom and the artificial skeleton allows us to simulate the neutron activation process using a Monte Carlo code, which enables us to calibrate the system for human subjects larger and smaller than the phantoms used as standards. copyright 1996 American Association of Physicists in Medicine

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

    CERN Document Server

    Mukherjee, B

    1999-01-01

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

  1. Measurement and analysis of fast neutron spectra in reactor materials by time-of-flight method

    International Nuclear Information System (INIS)

    Hayashi, Shuhei; Kimura, Itsuro; Kobayashi, Shohei; Yamamoto, Shuji; Nishihara, Hiroshi.

    1982-01-01

    The LINAC-TOF experiments have been done to measure the neutron energy spectra in the assemblies of reactor materials. The sample materials to be measured were iron, stainless steel, aluminum, nickel, zirconium, thorium, lithium, and so on. The shapes of assemblies were piles (rectangular parallelopiped, sphere, and polyhedron) and slab. A photoneutron target was set at the center of the pile assemblies. Each assembly has an electron injection hole and a re-entrant hole. In case of a slab, a photo neutron target was placed at the outside of the slab. Neutrons were generated by using an electron linear accelerator (LINAC). The length of the flight path was 20 m. The neutron detectors were a Li-6 glass scintillator and a B-10 vaseline-NaI(Tl) scintillator. The spatial distributions of neutrons in the piles were measured by the foil activation method. The neutron transport calculation was performed, and the evaluation of group constants was made. (Kato, T.)

  2. 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.

  3. Spectroscopy of light neutron deficient nuclei: 31Ar and 27S decays

    International Nuclear Information System (INIS)

    Borrel, V.

    1991-01-01

    Light neutron-deficient nuclei exhibit several interesting decay modes. In some cases, beta-delayed alpha emission becomes possible, and beta-delayed three-proton emission can be allowed. The energy spectra of the emitted protons should give informations on the position of the isobaric analog state and on its deexcitation modes. Experimental results obtained at GANIL with the LISE spectrometer on the decay of the isotopes 31 Ar and 27 S are presented. These data are discussed and compared with the predicted isobaric analog state excitation energies and with shell-model calculations. (G.P.) 13 refs.; 7 figs.; 1 tab

  4. Evaluation of Kalman filters and genetic algorithms for delayed-neutron nondestructive assay data analyses

    International Nuclear Information System (INIS)

    Aumeier, S.E.; Forsmann, J.H.

    1998-01-01

    The ability to nondestructively determine the presence and quantity of fissile/fertile nuclei in various matrices is important in several areas of nuclear applications, including international and domestic safeguards, radioactive waste characterization, and nuclear facility operations. An analysis was performed to determine the feasibility of identifying the masses of individual fissionable isotopes from a cumulative delayed-neutron signal resulting form the neutron irradiation of several uranium and plutonium isotopes. The feasibility of two separate data-processing techniques was studied: Kalman filtering and genetic algorithms. The basis of each technique is reviewed, and the structure of the algorithms as applied to the delayed-neutron analysis problem is presented. The results of parametric studies performed using several variants of the algorithms are presented. The effect of including additional constraining information such as additional measurements and known relative isotopic concentration is discussed. The parametric studies were conducted using simulated delayed-neutron data representative of the cumulative delayed-neutron response following irradiation of a sample containing 238 U, 235 U, 239 Pu, and 240 Pu. The results show that by processing delayed-neutron data representative of two significantly different fissile/fertile fission ratios, both Kalman filters and genetic algorithms are capable of yielding reasonably accurate estimates of the mass of individual isotopes contained in a given assay sample

  5. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

    2013-09-15

    Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

  6. The criteria for selecting a method for unfolding neutron spectra based on the information entropy theory

    International Nuclear Information System (INIS)

    Zhu, Qingjun; Song, Fengquan; Ren, Jie; Chen, Xueyong; Zhou, Bin

    2014-01-01

    To further expand the application of an artificial neural network in the field of neutron spectrometry, the criteria for choosing between an artificial neural network and the maximum entropy method for the purpose of unfolding neutron spectra was presented. The counts of the Bonner spheres for IAEA neutron spectra were used as a database, and the artificial neural network and the maximum entropy method were used to unfold neutron spectra; the mean squares of the spectra were defined as the differences between the desired and unfolded spectra. After the information entropy of each spectrum was calculated using information entropy theory, the relationship between the mean squares of the spectra and the information entropy was acquired. Useful information from the information entropy guided the selection of unfolding methods. Due to the importance of the information entropy, the method for predicting the information entropy using the Bonner spheres' counts was established. The criteria based on the information entropy theory can be used to choose between the artificial neural network and the maximum entropy method unfolding methods. The application of an artificial neural network to unfold neutron spectra was expanded. - Highlights: • Two neutron spectra unfolding methods, ANN and MEM, were compared. • The spectrum's entropy offers useful information for selecting unfolding methods. • For the spectrum with low entropy, the ANN was generally better than MEM. • The spectrum's entropy was predicted based on the Bonner spheres' counts

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. Informational-computer system for the neutron spectra analysis

    International Nuclear Information System (INIS)

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

    1979-01-01

    In this article basic principles of the build-up of the informational-computer system for the neutron spectra analysis on a basis of measured reaction rates are given. The basic data files of the system, needed software and hardware for the system operation are described

  9. Photon and photoneutron spectra produced in radiotherapy Linacs

    International Nuclear Information System (INIS)

    Vega C, H. R.; Martinez O, S. A.; Benites R, J. L.; Lallena, A. M.

    2011-10-01

    A Monte Carlo calculation, using the MCNPX code, was carried out in order to estimate the photon and neutron spectra in two locations of two linacs operating at 15 and 18 MV. Detailed models of both linac heads were used in the calculations. Spectra were estimated below the flattening filter and at the isocenter. Neutron spectra show two components due to evaporation and knock-on neutrons. Lethargy spectra under the filter were compared to the spectra calculated from the function quoted by Tosi et al. that describes reasonably well neutron spectra beyond 1 MeV, though tends to underestimate the energy region between 10 -6 and 1 MeV. Neutron and Bremsstrahlung spectra show the same features regardless of the linac voltage. The amount of photons and neutrons produced by the 15 MV linac is smaller than that found for the 18 MV linac. As expected, Bremsstrahlung spectra ends according to the voltage used to accelerate the electrons. (Author)

  10. Photon and photoneutron spectra produced in radiotherapy Linacs

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Av. Central del Norte Km. 1, Via Paipa Tunja, Boyaca (Colombia); Benites R, J. L. [Universidad Autonoma de Nayarit, Postgrado CBAP, Carretera Tepic Compostela Km. 9, Xalisco, Nayarit (Mexico); Lallena, A. M., E-mail: fermineutron@yahoo.com [Universida de Granada, Departamento de Fisica Atomica, Molecular y Nuclear, E-18071 Granada (Spain)

    2011-10-15

    A Monte Carlo calculation, using the MCNPX code, was carried out in order to estimate the photon and neutron spectra in two locations of two linacs operating at 15 and 18 MV. Detailed models of both linac heads were used in the calculations. Spectra were estimated below the flattening filter and at the isocenter. Neutron spectra show two components due to evaporation and knock-on neutrons. Lethargy spectra under the filter were compared to the spectra calculated from the function quoted by Tosi et al. that describes reasonably well neutron spectra beyond 1 MeV, though tends to underestimate the energy region between 10{sup -6} and 1 MeV. Neutron and Bremsstrahlung spectra show the same features regardless of the linac voltage. The amount of photons and neutrons produced by the 15 MV linac is smaller than that found for the 18 MV linac. As expected, Bremsstrahlung spectra ends according to the voltage used to accelerate the electrons. (Author)

  11. High energy neutron radiography

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. Inclusive zero-angle neutron spectra at the ISR and OPER-model

    International Nuclear Information System (INIS)

    Grigoryan, A.A.

    1977-01-01

    The invlusive zero-angle neutron spectra in pp-collisions measured at the ISR are compared with the OPER-model predictions. OPER-model rather well describes the experimental data. Some features of the spectra behaviour at fixed transverse momentum and large x are considered

  13. Prototype Neutron Energy Spectrometer

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  14. Prototype Neutron Energy Spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Stephen Mitchell, Sanjoy Mukhopadhyay, Richard Maurer, Ronald Wolff

    2010-06-16

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

  15. Solar Energetic Particle Spectra

    Science.gov (United States)

    Ryan, J. M.; Boezio, M.; Bravar, U.; Bruno, A.; Christian, E. R.; de Nolfo, G. A.; Martucci, M.; Mergè, M.; Munini, R.; Ricci, M.; Sparvoli, R.; Stochaj, S.

    2017-12-01

    We report updated event-integrated spectra from several SEP events measured with PAMELA. The measurements were made from 2006 to 2014 in the energy range starting at 80 MeV and extending well above the neutron monitor threshold. The PAMELA instrument is in a high inclination, low Earth orbit and has access to SEPs when at high latitudes. Spectra have been assembled from these high-latitude measurements. The field of view of PAMELA is small and during the high-latitude passes it scans a wide range of asymptotic directions as the spacecraft orbits. Correcting for data gaps, solid angle effects and improved background corrections, we have compiled event-integrated intensity spectra for twenty-eight SEP events. Where statistics permit, the spectra exhibit power law shapes in energy with a high-energy exponential roll over. The events analyzed include two genuine ground level enhancements (GLE). In those cases the roll-over energy lies above the neutron monitor threshold ( 1 GV) while the others are lower. We see no qualitative difference between the spectra of GLE vs. non-GLE events, i.e., all roll over in an exponential fashion with rapidly decreasing intensity at high energies.

  16. Neutron matter, symmetry energy and neutron stars

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  17. Logic based feature detection on incore neutron spectra

    Energy Technology Data Exchange (ETDEWEB)

    Racz, A.; Kiss, S.; Bende-Farkas, S. (Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics)

    1993-04-01

    A general framework for detecting features of incore neutron spectra with a rule-based methodology is presented. As an example, we determine the meaningful peaks in the APSD-s. This work is part of a larger project, aimed at developing a noise diagnostic expert system. (Author).

  18. Investigation of capture reactions far off stability by β-delayed neutron emission

    International Nuclear Information System (INIS)

    Wiescher, M.; Leist, B.; Ziegert, W.; Gabelmann, H.; Steinmueller, B.; Ohm, H.; Kratz, K.h.; Thielemann, F.h.; Hillebrandt, W.

    1985-01-01

    Beta-delayed neutron spectroscopy is applied to determine reaction rates of neutron capture on several neutron rich nuclei. The results of these experiments are presented and discussed in the light of their astrophysical implications. Furthermore, the experimental possibilities and limits of planned measurements are advertised

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. Information about the new 8-group delayed neutron set preparation

    International Nuclear Information System (INIS)

    Svarny, J.

    1998-01-01

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

  1. Neutron spectra unfolding with maximum entropy and maximum likelihood

    International Nuclear Information System (INIS)

    Itoh, Shikoh; Tsunoda, Toshiharu

    1989-01-01

    A new unfolding theory has been established on the basis of the maximum entropy principle and the maximum likelihood method. This theory correctly embodies the Poisson statistics of neutron detection, and always brings a positive solution over the whole energy range. Moreover, the theory unifies both problems of overdetermined and of underdetermined. For the latter, the ambiguity in assigning a prior probability, i.e. the initial guess in the Bayesian sense, has become extinct by virtue of the principle. An approximate expression of the covariance matrix for the resultant spectra is also presented. An efficient algorithm to solve the nonlinear system, which appears in the present study, has been established. Results of computer simulation showed the effectiveness of the present theory. (author)

  2. High energy neutron source for materials research and development

    International Nuclear Information System (INIS)

    Odera, M.

    1989-01-01

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

  3. Rapid uranium analysis by delayed neutron counting of neutron activated samples

    International Nuclear Information System (INIS)

    Papadopoulos, N.N.

    1985-01-01

    The uranium analyzer at the Nuclear Research Center ''Demokritos'' and the delayed neutron method have been used to determine the uranium content in lignite, in chemically enriched samples and in solutions of extractable uranium. The results are compared with data obtained by other methods. In the case of dissolved extractable uranium. The results are in good agreement with X-ray fluorescence data in the range 100 ppm to 2000 ppm while beyond these limits the discrepancies between neutron and spectrophotometric data are observed. The results for lignite samples are in good agreement with gamma spectrometric data. Discrepancies indicate that more extensive intercomparisons are needed to check the reliability of various methods

  4. Synovectomy by Neutron capture

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  5. Measurement of the high-energy neutron flux on the surface of the natural uranium target assembly QUINTA irradiated by deuterons of 4- and 8-GeV energy

    International Nuclear Information System (INIS)

    Adam, J.; Baldin, A.A.; Chilap, V.

    2014-01-01

    Experiments with a natural uranium target assembly QUINTA exposed to 4- and 8 GeV deuteron beams of the Nuclotron accelerator at the Joint Institute for Nuclear Research (Dubna) are analyzed. The 129 I, 232 Th, 233 U, 235 U, nat U, 237 Np, 238 Pu, 239 Pu and 241 Am radioactive samples were installed on the surface of the QUINTA set-up and irradiated with secondary neutrons. The neutron flux through the RA samples was monitored by Al foils. The reaction rates of 27 Al(n, y 1 ) 24 Na, 27 Al(n, y 2 ) 22 Na and 27 Al(n, y 3 ) 7 Be reactions with the effective threshold energies of 5, 27 and 119 MeV were measured at both 4- and 8-GeV deuteron beam energies. The average neutron fluxes between the effective threshold energies and the effective ends of the neutron spectra (which are 800 or 1000 MeV for energy of 4- or 8-GeV deuterons) were determined. The evidence for the intensity shift of the neutron spectra to higher neutron energies with the increase of the deuteron energy from 4 to 8 GeV was found from the ratios of the average neutron fluxes. The reaction rates and the average neutron fluxes were calculated with MCNPX2.7 and MARS15 codes.

  6. 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)

  7. MMAPDNG: A new, fast code backed by a memory-mapped database for simulating delayed γ-ray emission with MCNPX package

    Science.gov (United States)

    Lou, Tak Pui; Ludewigt, Bernhard

    2015-09-01

    The simulation of the emission of beta-delayed gamma rays following nuclear fission and the calculation of time-dependent energy spectra is a computational challenge. The widely used radiation transport code MCNPX includes a delayed gamma-ray routine that is inefficient and not suitable for simulating complex problems. This paper describes the code "MMAPDNG" (Memory-Mapped Delayed Neutron and Gamma), an optimized delayed gamma module written in C, discusses usage and merits of the code, and presents results. The approach is based on storing required Fission Product Yield (FPY) data, decay data, and delayed particle data in a memory-mapped file. When compared to the original delayed gamma-ray code in MCNPX, memory utilization is reduced by two orders of magnitude and the ray sampling is sped up by three orders of magnitude. Other delayed particles such as neutrons and electrons can be implemented in future versions of MMAPDNG code using its existing framework.

  8. Data reduction for a high-throughput neutron activation analysis system

    International Nuclear Information System (INIS)

    Bowman, W.W.

    1979-01-01

    To analyze samples collected as part of a geochemical survey for the National Uranium Resource Evaluation program, Savannah River Laboratory has installed a high-throughput neutron activation analysis system. As part of that system, computer programs have been developed to reduce raw data to elemental concentrations in two steps. Program RAGS reduces gamma-ray spectra to lists of photopeak energies, peak areas, and statistical errors. Program RICHES determines the elemental concentrations from photopeak and delayed-neutron data, detector efficiencies, analysis parameters (neutron flux and activation, decay, and counting times), and spectrometric and cross-section data from libraries. Both programs have been streamlined for on-line operation with a minicomputer, each requiring approx. 64 kbytes of core. 3 tables

  9. One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry, E-mail: zzhong@anl.gov, E-mail: alby@anl.gov, E-mail: gohar@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, IL (United States)

    2011-07-01

    The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β{sub eff} has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β{sub eff} was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β{sub eff}, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have

  10. One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity

    International Nuclear Information System (INIS)

    Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry

    2011-01-01

    The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β_e_f_f has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β_e_f_f was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β_e_f_f, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have been

  11. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Mozin, Vladimir [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Campbell, Luke [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hunt, Alan W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reedy, Edward T.E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Seipel, Heather [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235U, 239Pu, and 241Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241Pu, a significant fissile constituent in spent fuel, was measured and compared to 239Pu. The 241Pu/239Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr3 detectors

  12. Diabatic emission of neutrons: A probe for the energy dissipation mechanism in nucleus-nucleus collisions

    International Nuclear Information System (INIS)

    Noerenberg, W.; Cassing, W.

    1984-05-01

    The precompound emission of neutrons in central nucleus-nucleus collisions is investigated within the framework of dissipative diabatic dynamics. For 92 Mo + 92 Mo at bombarding energies between 7.5 and 20 MeV/u the differential neutron multiplicities dMsub(n)/dEsub(n) are estimated from the decay of highly excited diabatic single-particle states. The energy spectra have an almost exponential high-energy tail with effective temperatures up to 10 MeV for 20 MeV/u bombarding energy. (orig.)

  13. Effect of isospin degree of freedom on transverse momentum spectra

    International Nuclear Information System (INIS)

    Kaur, Sukhjit; Swati

    2013-01-01

    We study the effect of isospin degree of freedom, incident energy as well as system mass on the behavior of transverse momentum spectra, dN/p t dp t , of neutrons and protons. We find that most of the nucleons suffer soft collisions. The effect of isospin degree of freedom on transverse spectra diminishes with the increase in the incident energy. In Fermi energy region, transverse momentum spectra of both protons and neutrons show sensitivity toward the density dependence of symmetry energy. (author)

  14. Phonon spectra in the parent superconducting iron-tuned telluride F e1 +xTe from inelastic neutron scattering and ab initio calculations

    Science.gov (United States)

    Zbiri, Mohamed; Viennois, Romain

    2017-10-01

    We report inelastic neutron scattering measurements of phonon spectra in the parent superconductor iron-tuned chalcogenide F e1 +xTe for two different x contents (x ≤0.11 ) using neutron time-of-flight technique. Thermal neutron spectroscopy allowed the collection of the low-temperature Stokes spectra over an extended Q range at 2, 40, and 120 K, hence covering both the magnetic monoclinic and the paramagnetic tetragonal phases, whereas cold neutrons allowed the measurement of high-resolution anti-Stokes spectra at 140, 220, and 300 K, thus covering the tetragonal phase. Our results evidence a spin-phonon coupling behavior towards the observed noticeable temperature-dependent change of the Stokes spectra across the transition temperatures. On the other hand, the anti-Stokes spectra reveal a pronounced hardening of the low-energy, acoustic region of the phonon spectrum upon heating, indicating a strong anharmonicity and a subtle dependence of phonons on structural evolution within the tetragonal phase. Experimental results are accompanied by ab initio calculations of phonon spectra of the tetragonal stoichiometric phase for a comparison with the high-resolution anti-Stokes spectra. Calculations included different density functional methods. Spin polarization and van der Waals interaction were either considered or neglected, individually or concomitantly, in order to study their respective effect on lattice dynamics description. Our results suggest that including van der Waals interaction has only a slight effect on phonon dynamics; however, phonon spectra are better described when spin polarization is included in a cooperative way with van der Waals interactions.

  15. Influence of Neutron Spectra Unfolding Method on Fast Neutron Dose Determination

    International Nuclear Information System (INIS)

    Marinkovic, P.

    1991-01-01

    Full text: Accuracy of knowing the fast neutron spectra has great influence on equivalent dose determination. In usual fast neutron spectrum measurements with scintillation detectors based on proton recoil, the main difficulty is confidence of unfolding method. In former ones variance of obtained result is usually great and negative values are possible too, which does means that we don't now exactly is obtained neutron spectrum real one. The new unfolding method based on Shanon's information theory, which gives non-negative spectrum and relative low variance, is obtained and appropriate numerical code for application in fast neutron spectrometry based on proton recoil is realized. In this method principle of maximum entropy and maximum likelihood are used together. Unknown group density distribution functions, which are considered as desired normalized mean neutron group flux, are constl u cted using only constrain of knowing mean value. Obtained distributions are consistent to available information (counts in NCA from proton recoil), while being maximally noncommittal with respect to all other unknown circumstances. For maximum likelihood principle, distribution functions around mean value of counts in the channels of MCA are taken to be Gauss function shape. Optimal non-negative solution is searched by means of Lagrange parameter method. Nonlinear system of equations, is solved using gradient and Newton iterative algorithm. Error covariance matrix is obtained too. (author)

  16. The dependence of radiation damage analysis on neutron dosimetry

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  17. Development of a BaF2 scintillation spectrometer for evaluation of photon energy spectra in workplaces around nuclear facilities

    International Nuclear Information System (INIS)

    Urabe, Itsumasa; Yoshimoto, Taka-aki; Kobayashi, Katsuhei; Akiyoshi, Tsunekazu; Tsujimoto, Tadashi; Nakashima, Yoshiyuki; Oda, Keiji.

    1997-01-01

    A BaF 2 scintillation spectrometer has been constructed for the determination of photon energy spectra in workplaces around nuclear facilities. Energy absorption spectra by the BaF 2 detector were calculated with the EGS4 Monte Carlo code in the energy region from 0.1 to 100 MeV and a response matrix of the spectrometer was obtained from the energy absorption spectra, of which the energy resolutions were modified to fit to the experimental results. With the irradiation experiments using neutron-capture gamma rays and those from radioactive sources, it became clear that photon energy spectra can be evaluated within an error of about 10% in the energy region 0.1 MeV to a few tens of megaelectronvolts. (author)

  18. The energy spectrum of cosmic-ray induced neutrons measured on an airplane over a wide range of altitude and latitude

    International Nuclear Information System (INIS)

    Goldhagen, P.; Clem, J. M.; Wilson, J. W.

    2004-01-01

    Crews of high-altitude aircraft are exposed to radiation from galactic cosmic rays (GCRs). To help determine such exposures, the Atmospheric Ionizing Radiation Project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on a NASA ER-2 high-altitude airplane. The primary instrument was a sensitive extended-energy multisphere neutron spectrometer. Its detector responses were calculated for energies up to 100 GeV using the radiation transport code MCNPX 2.5.d with improved nuclear models and including the effects of the airplane structure. New calculations of GCR-induced particle spectra in the atmosphere were used to correct for spectrometer counts produced by protons, pions and light nuclear ions. Neutron spectra were unfolded from the corrected measured count rates using the deconvolution code MAXED 3.1. The results for the measured cosmic-ray neutron spectrum (thermal to >10 GeV), total neutron fluence rate, and neutron dose equivalent and effective dose rates, and their dependence on altitude and geomagnetic cut-off agree well with results from recent calculations of GCR-induced neutron spectra. (authors)

  19. DANTE, Activation Analysis Neutron Spectra Unfolding by Covariance Matrix Method

    International Nuclear Information System (INIS)

    Petilli, M.

    1981-01-01

    1 - Description of problem or function: The program evaluates activation measurements of reactor neutron spectra and unfolds the results for dosimetry purposes. Different evaluation options are foreseen: absolute or relative fluxes and different iteration algorithms. 2 - Method of solution: A least-square fit method is used. A correlation between available data and their uncertainties has been introduced by means of flux and activity variance-covariance matrices. Cross sections are assumed to be constant, i.e. with variance-covariance matrix equal to zero. The Lagrange multipliers method has been used for calculating the solution. 3 - Restrictions on the complexity of the problem: 9 activation experiments can be analyzed. 75 energy groups are accepted

  20. Validation of neutron data libraries by backscattered spectra of Pu-Be Neutrons

    CERN Document Server

    El-Agib, I

    1999-01-01

    Elastically backscattered spectra of Pu-Be neutrons have been measured for SiO sub 2 , water, graphite, paraffin oil and Al slabs using a proton recoil spectrometer. The results were compared with the calculated spectra obtained by the three-dimensional Monte-Carlo transport code MCNP-4B and point-wise cross sections from the ENDF/B-V, ENDF/B-VI, JENDL-3.1 and BROND-2 data libraries. The good agreement between the measured and calculated results indicates that this procedure can be used for validation of different data libraries. This simple method renders possible the detection of oxygen, carbon and hydrogen in bulk samples. (author)

  1. Neutron emission and fragment yield in high-energy fission

    International Nuclear Information System (INIS)

    Grudzevich, O. T.; Klinov, D. A.

    2013-01-01

    The KRIS special library of spectra and emission probabilities in the decays of 1500 nuclei excited up to energies between 150 and 250 MeV was developed for correctly taking into account the decay of highly excited nuclei appearing as fission fragments. The emission of neutrons, protons, and photons was taken into account. Neutron emission fromprimary fragments was found to have a substantial effect on the formation of yields of postneutron nuclei. The library was tested by comparing the calculated and measured yields of products originating from the fission of nuclei that was induced by high-energy protons. The method for calculating these yields was tested on the basis of experimental data on the thermal-neutroninduced fission of 235 U nuclei

  2. Use of delayed gamma rays for active non-destructive assay of {sup 235}U irradiated by pulsed neutron source (plasma focus)

    Energy Technology Data Exchange (ETDEWEB)

    Andola, Sanjay; Niranjan, Ram [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kaushik, T.C., E-mail: tckk@barc.gov.in [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Rout, R.K. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ashwani; Paranjape, D.B.; Kumar, Pradeep; Tomar, B.S.; Ramakumar, K.L. [Radioanalytical Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, S.C. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-07-01

    A pulsed neutron source based on plasma focus device has been used for active interrogation and assay of {sup 235}U by monitoring its delayed high energy γ-rays. The method involves irradiation of fissile material by thermal neutrons obtained after moderation of a burst of neutrons emitted upon fusion of deuterium in plasma focus (PF) device. The delayed gamma rays emitted from the fissile material as a consequence of induced fission were detected by a large volume sodium iodide (NaI(Tl)) detector. The detector is coupled to a data acquisition system of 2k input size with 2k ADC conversion gain. Counting was carried out in pulse height analysis mode for time integrated counts up to 100 s while the temporal profile of delayed gamma has been obtained by counting in multichannel scaling mode with dwell time of 50 ms. To avoid the effect of passive (natural) and active (from surrounding materials) backgrounds, counts have been acquired for gamma energy between 3 and 10 MeV. The lower limit of detection of {sup 235}U in the oxide samples with this set-up is estimated to be 14 mg.

  3. Adjusted neutron spectra of STEK cores for reactivity calculations

    International Nuclear Information System (INIS)

    Dekker, J.W.M.; Dragt, J.B.; Janssen, A.J.; Heijboer, R.J.; Klippel, H.Th.

    1978-02-01

    Neutron flux and adjoint flux spectra form a pre-requisite in the analysis of reactivity worth data measured in the STEK facility. First, a survey of all available information about these spectra is given. Next a special application of a general adjustment method is described. This method has been used to obtain adjusted STEK group flux and adjoint flux spectra, starting from calculated spectra. These theoretical spectra were adjusted to reactivity worths of natural boron (nat. B) and 235 U as well as a number of fission reaction rates. As a by-product in this adjustment calculation adjusted fission group cross sections of 235 U were obtained. The results, viz. group fluxes and adjoint fluxes and adjusted fission cross sections of 235 U are given. They have been used for the interpretation of fission product reactivity worth measurements made in STEK

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  5. 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

  6. Neutron reference spectra measurements with the Bonner multi-spheres spectrometer; Medidas de espectros de referencia de neutrons com o espectrometro de multiesferas de Bonner

    Energy Technology Data Exchange (ETDEWEB)

    Lemos Junior, Roberto Mendonca de

    2004-07-01

    This paper aims to define a procedure to use the Bonner Multisphere Spectrometer with a {sup 6}LiI(Eu) detector in order to determine of neutron spectra. It was measured {sup 238}PuBe spectra and same of reference ({sup 241}AmBe, {sup 252}Cf e {sup 252}Cf+D{sub 2}O) published in ISO 8529-1 (2001) Norm. The data were processed by a computer program (BUNKI), which presents the results in neutrons energy fluency. Each input parameter of the program was studied in order to establish their influence in the adjustment result. The environment dose equivalent rate obtained placing the detector 1 m from the {sup 241}AmBe source was 122 {+-} 4 {mu}Sv/h with 7% of uncertainty and 95% of confidence level. The procedure established in this work was tested with the {sup 238}PuBe spectrum, obtaining an environment dose equivalent rate of 286 {+-} 9 {mu}Sv/h, 8% lower than the value measured experimentally used as reference. Through this procedure will be possible to measure neutron spectra in different work places where neutrons sources are used. Knowing these spectra, it will be possible to evaluate which area monitors, are more suitable, as well as, to study better the response of individual neutron monitors, as for instance, to obtain a conversion coefficient more appropriate to the albedo dosimeter used in different work places. As the measurements need a long time to be accomplished, the work optimization is fundamental to reduce the exposing time of the Bonner spectrometer operator. For this reason, an important parameter examined in this paper was the possibility of reducing the number of spheres used during the measurement without changing the final result. Considering the radiation protection standards, this parameter has a huge importance when the measurements are performed in work places where the neutron fluency and gamma rate offer risks to the operator's health, as for instance, in nuclear centrals. Studying this parameter, it was possible to conclude that

  7. Selected aspects in the structure of beta-delayed particle spectra

    International Nuclear Information System (INIS)

    Honkanen, J.; Aeystoe, J.; Eskola, K.

    1986-01-01

    Some weak beta-delayed particle emitters in the T z =-3/2, -1, -1/2, +1/2 and +5/2 series are reviewed. Selected features of the delayed particle emission are discussed in terms of experimental delayed particle data and (p,γ), (p,p') and (p,n) reaction data. Experimental beta transition strengths are compared with the existing complete shell-model calculations for the sd-shell nuclei. The effect of the Gamow-Teller giant resonance on the structure of the delayed particle spectra is considered. The correlation between the widths of two decay channels, protons and alpha particles, and the preceeding beta decay is studied in the case of the 40 Sc decay. (orig.)

  8. High-order harmonic generation spectra and isolated attosecond pulse generation with a two-color time delayed pulse

    International Nuclear Information System (INIS)

    Feng Liqiang; Chu Tianshu

    2012-01-01

    Highlights: ► Investigation of HHG spectra and single isolated attosecond pulse generation. ► Irradiation from a model Ne atom by two-color time delayed pulse. ► Observation of time delay effect and relative phase effect. ► Revelation of the optimal condition for generating isolated attosecond pulse. ► Generation of a single isolated attosecond pulse of 45as. - Abstract: In this paper, we theoretically investigate the delay time effect on the high-order harmonic generation (HHG) when a model Ne atom is exposed to a two-color time delayed pulse, consisting of a 5fs/800 nm fundamental field and a 20fs/2000 nm controlling field. It shows that the HHG spectra are strongly sensitive to the delay time between the two laser fields, in particular, for the zero carrier-envelope phase (CEP) φ case (corresponding to the 800 nm fundamental field), the maximum cutoff energy has been achieved at zero delay time. However, with the introduction of the CEP (φ = 180°), the delay effect on HHG is changed, exhibiting a ‘U’ structure harmonic emission from −1 T to 1 T. In addition, the combinations of different controlling pulse frequencies and pulse intensities have also been considered, showing the similar results as the original controlling field case, but with some characteristics. Finally, by properly superposing the optimal harmonic spectrum, an isolated 45as pulse is generated without phase compensation.

  9. Consistency between data from the ENDF/B-V dosimetry file and corresponding experimental data for some fast neutron reference spectra

    International Nuclear Information System (INIS)

    Nolthenius, H.J.; Zijp, W.L.

    1981-11-01

    Results are given of a study on the consistency between 'integral' and 'differential' cross sections data for four benchmark neutron spectra and 36 neutron reactions of importance for reactor neutron metrology. The energy dependent cross section data and their uncertainty data are obtained from the ENDF/B-V dosimetry file. The reactions have been considered with respect to the following quantities: 1. the precision of the averaged cross sections, for a specified spectrum; 2. the discrepancy between the measured and the calculated average cross section values; 3. the consistency between the measured and calculated average cross section values, described by the chi 2 -parameter. It was possible to take into account the available cross section covariance information present in the ENDF/B-V dosimetry file. Covariance information on the benchmark flux density spectra was not taken into account in this study

  10. Spectra of neutrons and fusion charged products produced in a dense laser plasma

    International Nuclear Information System (INIS)

    Burtsev, V.A.; Dyatlov, V.D.; Krzhizhanovskij, R.E.; Levkovskij, A.A.

    1977-01-01

    The possibility of laser-produced plasma diagnostics has been investigated by measuring spectra of neutrons and alpha particles produced in the T(d,n) 4 He reaction. Using the Monte Carlo method the spectra have been calculated for nine states of the deuterium-tritium plasma with the temperature of 1;5 and 10 keV and the density of 0.2; 1 and 10 g/cm 3 respectively. The initial radius of the target was assumed to be 0.01 cm at the density of 0.2 g/cm 3 . It is shown that the neutron and alpha spectra can serve as plasma diagnostics parameters in laser fusion

  11. Determination of neutron spectra using the programs GNSR and SPECTRIX

    International Nuclear Information System (INIS)

    Weyrauch, M.; Dietz, E.; Matzke, M.

    2002-01-01

    We describe the capabilities and the application of two computer programs, which have been developed in order to facilitate common tasks in neutron spectrometry: GNSR (calculation of response matrices) and SPECTRIX (unfolding). Gas-filled Neutron Spectrometer Response calculates response functions and response matrices of various gas-filled neutron detectors. It can be configured to accommodate the appropriate gas-fillings and supports a number of different neutron beam configurations with a possibility to input calculated or measured neutron beam spectra. The program includes graphical capabilities as well as a context-sensitive help system. SPECTRIX implements several unfolding algorithms as well as support algorithms for unfolding and includes graphics capabilities and context-sensitive help. We apply both programs to a specific example: calculation of the response matrix of a 3 He detector and unfolding of the neutron spectrum of a thick accelerator target using the calculated response matrix

  12. Calculation of prompt neutron spectra for curium isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.

    1997-03-01

    With the aim of checking the existing evaluations contained in JENDL-3.2 and providing new evaluations based on a methodology proposed by the author, a series of calculations of prompt neutron spectra have been undertaken for curium isotopes. Some of the evaluations in JENDL-3.2 was found to be unphysically hard and should be revised. (author)

  13. Summary Report of Second Research Coordination Meeting on Prompt Fission Neutron Spectra of Major Actinides

    International Nuclear Information System (INIS)

    Capote Noy, R.

    2013-09-01

    A summary is given of the Second Research Coordination Meeting on Prompt Fission Neutron Spectra of Actinides. Experimental data and modelling methods on prompt fission neutron spectra were reviewed. Extensive technical discussions held on theoretical methods to calculate prompt fission spectra. Detailed coordinated research proposals have been agreed. Summary reports of selected technical presentations at the meeting are given. The resulting work plan of the Coordinated Research Programme is summarized, along with actions and deadlines. (author)

  14. Delayed gamma-ray spectroscopy with lanthanum bromide detector for non-destructive assay of nuclear material

    Science.gov (United States)

    Favalli, Andrea; Iliev, Metodi; Ianakiev, Kiril; Hunt, Alan W.; Ludewigt, Bernhard

    2018-01-01

    High-energy delayed γ-ray spectroscopy is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Requirements for the γ-ray detection system, up to ∼6 MeV, can be summarized as follows: high efficiency at high γ-ray energies, high energy resolution, good linearity between γ-ray energy and output signal amplitude, ability to operate at very high count rates, and ease of use in industrial environments such as nuclear facilities. High Purity Germanium Detectors (HPGe) are the state of the art and provide excellent energy resolution but are limited in their count rate capability. Lanthanum Bromide (LaBr3) scintillation detectors offer significantly higher count rate capabilities at lower energy resolution. Thus, LaBr3 detectors may be an effective alternative for nuclear spent-fuel applications, where count-rate capability is a requirement. This paper documents the measured performance of a 2" (length) × 2" (diameter) of LaBr3 scintillation detector system, coupled to a negatively biased PMT and a tapered active high voltage divider, with count-rates up to ∼3 Mcps. An experimental methodology was developed that uses the average current from the PMT's anode and a dual source method to characterize the detector system at specific very high count rate values. Delayed γ-ray spectra were acquired with the LaBr3 detector system at the Idaho Accelerator Center, Idaho State University, where samples of ∼3g of 235U were irradiated with moderated neutrons from a photo-neutron source. Results of the spectroscopy characterization and analysis of the delayed γ-ray spectra acquired indicate the possible use of LaBr3 scintillation detectors when high count rate capability may outweigh the lower energy resolution.

  15. Measurement of neutron spectra in varied environments by the foil-activation method with arbitrary trials

    International Nuclear Information System (INIS)

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

    1987-12-01

    Neutron spectra have been measured by the foil-activation method in 13 different environments in and around the Sandia Pulsed Reactor, the White Sands Missile Range Fast Burst Reactor, and the Sandia Annular Core Research Reactor. The spectra were obtained by using the SANDII code in a manner that was not dependent on the initial trial. This altered technique is better suited for the determination of spectra in environments that are difficult to predict by calculation, and it tends to reveal features that may be biased out by the use of standard trial-dependent methods. For some of the configurations, studies have also been made of how well the solution is determined in each energy region. The experimental methods and the techniques used in the analyses are thoroughly explained. 34 refs., 51 figs., 40 tabs

  16. Testing of a Code for the Calculation of Spectra of Neutrons Produced in a Target of a Neutron Generator

    Science.gov (United States)

    Gaganov, V. V.

    2017-12-01

    The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.

  17. A delayed neutron technique for measuring induced fission rates in fresh and burnt LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K.A., E-mail: kajordan@gmail.co [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland); Perret, G. [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland)

    2011-04-01

    The LIFE-PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel following re-irradiation in the zero-power PROTEUS research reactor. One such technique utilizes the measurement of delayed neutrons. To demonstrate the feasibility of the delayed neutron technique, fresh and burnt UO{sub 2} fuel samples were irradiated in different positions in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Fission rate ratios of the same sample irradiated in two different positions (inter-positional) and of two different samples irradiated in the same position (inter-sample) were derived from the measurements and compared with Monte Carlo predictions. Derivation of fission rate ratios from the delayed neutron measured signal requires correcting the signal for the delayed neutron source properties, the efficiency of the measurement setup, and the time dependency of the signal. In particular, delayed neutron source properties strongly depend on the fissile and fertile isotopes present in the irradiated sample and must be accounted for when deriving inter-sample fission rate ratios. Measured inter-positional fission rate ratios generally agree within 1{sigma} uncertainty (on the order of 1.0%) with the calculation predictions. For a particular irradiation position, however, a bias of about 2% is observed and is currently under investigation. Calculated and measured inter-sample fission rate ratios have C/E values deviating from unity by less than 1% and within 2{sigma} of the statistical uncertainties. Uncertainty arising from delayed neutron data is also assessed, and is found to give an additional 3% uncertainty factor. The measurement data indicate that uncertainty is overestimated.

  18. Using the SAND-II and MLM methods to reconstruct fast neutron spectra

    International Nuclear Information System (INIS)

    Bondars, Kh.Ya.; Kamnev, V.A.; Lapenas, A.A.; Troshin, V.S.

    1981-01-01

    The reconstruction of fast neutron spectra from measured reaction rates may be reduced to the solution of Fredholm's integral equation of the first kind. This problem falls in the category of incorrectly formulated problems, and so additional information is required concerning the unknown function i.e. concerning the differential energy dependence of the neutron, flux density sup(phi)(E). There are various methods for seeking a solution to the problem as formulated above. One of the best-known methods used in the USSR is the maximum likelihood method (MLM) (or directional difference method (DDM)), whereas SAND-II is commonly used abroad. The purpose of this paper is to compare the MLM and SAND-II methods, taking as an example the processing of measurement data which were obtained in the B-2 beam line at the BR-10 reactor in order to determine the composition of shielding for a fast reactor

  19. Study on thermal neutron spectra in reactor moderators by time-of-flight method

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1982-12-01

    Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

  20. Delayed Particle Study of Neutron Rich Lithium Isotopes

    CERN Multimedia

    Marechal, F; Perrot, F

    2002-01-01

    We propose to make a systematic complete coincidence study of $\\beta$-delayed particles from the decay of neutron-rich lithium isotopes. The lithium isotopes with A=9,10,11 have proven to contain a vast information on nuclear structure and especially on the formation of halo nuclei. A mapping of the $\\beta$-strength at high energies in the daughter nucleus will make possible a detailed test of our understanding of their structure. An essential step is the comparison of $\\beta$-strength patterns in $^{11}$Li and the core nucleus $^{9}$Li, another is the full characterization of the break-up processes following the $\\beta$-decay. To enable such a measurement of the full decay process we will use a highly segmented detection system where energy and emission angles of both charged and neutral particles are detected in coincidence and with high efficiency and accuracy. We ask for a total of 30 shifts (21 shifts for $^{11}$Li, 9 shifts $^{9}$Li adding 5 shifts for setting up with stable beam) using a Ta-foil target...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-15

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

  2. 239Pu prompt fission neutron spectra impact on a set of criticality and experimental reactor benchmarks

    International Nuclear Information System (INIS)

    Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.

    2014-01-01

    A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239 Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239 Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to

  3. Neutron energy spectrum in graphite blankets of fusion reactors

    International Nuclear Information System (INIS)

    Tsechanski, A.

    1981-09-01

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

  4. Characteristics of GaAs MESFET inverters exposed to high energy neutrons

    International Nuclear Information System (INIS)

    Bloss, W.L.; Yamada, W.E.; Young, A.M.; Janousek, B.K.

    1988-01-01

    GaAs MESFET circuits have been exposed to high energy neutrons with fluences ranging from 1x10/sup 14/ n/cm/sup 2/ to 2x10/sup 15/ m/cm/sup 2/. Discrete transistors, inverters, and ring oscillators were characterized at each fluence. While the MESFETs exhibit significant threshold voltage shifts and transconductance and saturation current degradation over this range of neutron fluences, the authors have observed improvement in the DC characteristics of Schottky Diode FET Logic (SDFL) inverters. This unusual result has been successfully simulated using device parameters extracted from FETs damaged by exposure to high energy neutrons. Although the decrease in device transconductance results in an increase in inverter gate delay, as reflected in ring oscillator frequency measurements, the authors conclude that GaAs ICs fabricated from this logic family will remain functional after exposure to extreme neutron fluences. This is a consequence of the observed improvement in inverter noise margin evident in both measured and simulated circuit performance

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

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi

    1983-01-01

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

  6. Prompt gamma-based neutron dosimetry for Am-Be and other workplace neutron spectra

    International Nuclear Information System (INIS)

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

    2016-01-01

    A new field-deployable technique for estimating the neutron ambient dose equivalent H*(10) by using the measured prompt gamma intensities emitted from borated high-density polyethylene (BHDPE) and the combination of normal HDPE and BHDPE with different configurations have been evaluated in this work. Monte Carlo simulations using the FLUKA code has been employed to calculate the responses from the prompt gammas emitted due to the monoenergetic neutrons interacting with boron, hydrogen, and carbon nuclei. A suitable linear combination of these prompt gamma responses (dose conversion coefficient (DCC)-estimated) is generated to approximate the International Commission on Radiological Protection provided DCC using the cross-entropy minimization technique. In addition, the shape and configurations of the HDPE and BHDPE combined system are optimized using the FLUKA code simulation results. The proposed method is validated experimentally, as well as theoretically, using different workplace neutron spectra with a satisfactory outcome. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  8. FIER: Software for analytical modeling of delayed gamma-ray spectra

    Science.gov (United States)

    Matthews, E. F.; Goldblum, B. L.; Bernstein, L. A.; Quiter, B. J.; Brown, J. A.; Younes, W.; Burke, J. T.; Padgett, S. W.; Ressler, J. J.; Tonchev, A. P.

    2018-05-01

    A new software package, the Fission Induced Electromagnetic Response (FIER) code, has been developed to analytically predict delayed γ-ray spectra following fission. FIER uses evaluated nuclear data and solutions to the Bateman equations to calculate the time-dependent populations of fission products and their decay daughters resulting from irradiation of a fissionable isotope. These populations are then used in the calculation of γ-ray emission rates to obtain the corresponding delayed γ-ray spectra. FIER output was compared to experimental data obtained by irradiation of a 235U sample in the Godiva critical assembly. This investigation illuminated discrepancies in the input nuclear data libraries, showcasing the usefulness of FIER as a tool to address nuclear data deficiencies through comparison with experimental data. FIER provides traceability between γ-ray emissions and their contributing nuclear species, decay chains, and parent fission fragments, yielding a new capability for the nuclear science community.

  9. YAP scintillators for resonant detection of epithermal neutrons at pulsed neutron sources

    International Nuclear Information System (INIS)

    Tardocchi, M.; Gorini, G.; Pietropaolo, A.; Andreani, C.; Senesi, R.; Rhodes, N.; Schooneveld, E. M.

    2004-01-01

    Recent studies indicate the resonance detector (RD) technique as an interesting approach for neutron spectroscopy in the electron volt energy region. This work summarizes the results of a series of experiments where RD consisting of YAlO 3 (YAP) scintillators were used to detect scattered neutrons with energy in the range 1-200 eV. The response of YAP scintillators to radiative capture γ emission from a 238 U analyzer foil was characterized in a series of experiments performed on the VESUVIO spectrometer at the ISIS pulsed neutron source. In these experiments a biparametric data acquisition allowed the simultaneous measurements of both neutron time-of-flight and γ pulse height (energy) spectra. The analysis of the γ pulse height and neutron time of flight spectra permitted to identify and distinguish the signal and background components. These measurements showed that a significant improvement in the signal-to-background ratio can be achieved by setting a lower level discrimination on the pulse height at about 600 keV equivalent photon energy. Present results strongly indicate YAP scintillators as the ideal candidate for neutron scattering studies with epithermal neutrons at both very low (<5 deg.) and intermediate scattering angles

  10. Neutron emission spectra of excited 126–140Sn nuclei

    International Nuclear Information System (INIS)

    Aggarwal, Mamta; Rajasekaran, M.

    2004-01-01

    We investigate one-neutron and two-neutron emission from 132 Sn and its neighboring isotopes due to thermal excitation. The rotational states of 132 Sn at different temperatures are investigated. The effects of separation energy and thermal excitation energy on neutron emission probability are studied. (author)

  11. The activation method for determining neutron spectra and fluences

    International Nuclear Information System (INIS)

    Hogel, J.; Vespalec, R.

    1980-01-01

    3 mm thick foils of 4 and 17 mm in diameter were used for measurements. NaI scintillation detectors 45 mm in diameter by 50 mm thick and 40 mm in diameter by 1 mm thick, and a Ge-Li spectrometer of 53 cm 3 in volume were used for gamma detection. A photopeak or a certain part of the integral spectrum was measured for each radionuclide. Computer code PIKAR was applied in automatic calculation of a simple gamma spectrum obtained using the semiconductor spectrometer. The FACT code was used for calculating foil activity. Codes SAND II and RFSP were used for neutron spectra unfolding. Ge-Li detector spectrometry was used for determining neutron fluence. Code FLUE was used for determining the mean value of neutron flux density and fluence. (J.P.)

  12. Gamma-Ray Emission Spectra as a Constraint on Calculations of 234,236,238U Neutron-Capture Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Ullmann, John Leonard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kawano, Toshihiko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bredeweg, Todd Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Baramsai, Bayarbadrakh [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Couture, Aaron Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haight, Robert Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jandel, Marian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); O' Donnell, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vieira, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wilhelmy, Jerry B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Becker, John A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wu, Ching-Yen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Krticka, Milan [Charles Univ., Prague (Czech Republic)

    2015-05-28

    Neutron capture cross sections in the “continuum” region (>≈1 keV) and gamma-emission spectra are of importance to basic science and many applied fields. Careful measurements have been made on most common stable nuclides, but physicists must rely on calculations (or “surrogate” reactions) for rare or unstable nuclides. Calculations must be benchmarked against measurements (cross sections, gamma-ray spectra, and <Γγ>). Gamma-ray spectrum measurements from resolved resonances were made with 1 - 2 mg/cm2 thick targets; cross sections at >1 keV were measured using thicker targets. The results show that the shape of capture cross section vs neutron energy is not sensitive to the form of the strength function (although the magnitude is); the generalized Lorentzian E1 strength function is not sufficient to describe the shape of observed gamma-ray spectra; MGLO + “Oslo M1” parameters produces quantitative agreement with the measured 238U(n,γ) cross section; additional strength at low energies (~ 3 MeV) -- likely M1-- is required; and careful study of complementary results on low-lying giant resonance strength is needed to consistently describe observations.

  13. Design of an artificial neural network, with the topology oriented to the reconstruction of neutron spectra; Diseno de una red neuronal artificial, con la topologia orientada a la reconstruccion del espectro de neutrones

    Energy Technology Data Exchange (ETDEWEB)

    Arteaga A, T.; Ortiz R, J.M.; Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.; Mercado S, G.A. [Unidades Academicas de Estudios Nucleares, Ingenieria Electrica y Matematicas, Universidad de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico)]. e-mail: tarcicio70@yahoo.co.uk

    2006-07-01

    People that live in high places respect to the sea level, in latitudes far from the equator or that they travel by plane, they are exposed to atmospheres of high radiation generated by the cosmic rays. Another atmosphere with radiation is the medical equipment, particle accelerators and nuclear reactors. The evaluation of the biological risk for neutron radiation requires an appropriate and sure dosimetry. A commonly used system is the Bonner Sphere Spectrometer (EEB) with the purpose of reconstructing the spectrum that is important because the equivalent dose for neutrons depends strongly on its energy. The count rates obtained in each sphere are treated, in most of the cases, for iterative methods, Monte Carlo or Maximum Entropy. Each one of them has difficulties that it motivates to the development of complementary procedures. Recently it has been used Artificial Neural Networks, ANN) and not yet conclusive results have been obtained. In this work it was designed an ANN to obtain the neutron energy spectrum neutrons starting from the counting rate of count of an EEB. The ANN was trained with 129 reference spectra obtained of the IAEA (1990, 2001), 24 were built as defined energy, including isotopic sources of neutrons of reference and operational, of accelerators, reactors, mathematical functions, and of defined energy with several peaks. The spectrum was transformed from lethargy units to energy and were reaccommodated in 31 energies using the Monte Carlo code 4C. The reaccommodated spectra and the response matrix UTA4 were used to calculate the prospective count rates in the EEB. These rates were used as entrance and its respective spectrum was used as output during the net training. The net design is Retropropagation type with 5 layers of 7, 140, 140, 140 and 31 neurons, transfer function logsig, tansig, logsig, logsig, logsig respectively. Training algorithm, traingdx. After the training, the net was proven with a group of training spectra and others that

  14. Neutron generator (HIRRAC) and dosimetry study.

    Science.gov (United States)

    Endo, S; Hoshi, M; Takada, J; Tauchi, H; Matsuura, S; Takeoka, S; Kitagawa, K; Suga, S; Komatsu, K

    1999-12-01

    Dosimetry studies have been made for neutrons from a neutron generator at Hiroshima University (HIRRAC) which is designed for radiobiological research. Neutrons in an energy range from 0.07 to 2.7 MeV are available for biological irradiations. The produced neutron energies were measured and evaluated by a 3He-gas proportional counter. Energy spread was made certain to be small enough for radiobiological studies. Dose evaluations were performed by two different methods, namely use of tissue equivalent paired ionization chambers and activation of method with indium foils. Moreover, energy deposition spectra in small targets of tissue equivalent materials, so-called lineal energy spectrum, were also measured and are discussed. Specifications for biological irradiation are presented in terms of monoenergetic beam conditions, dose rates and deposited energy spectra.

  15. 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

  16. Activity of the Delayed Neutron Working Group of JNDC and the International Evaluation Cooperation - WPEC/SG6

    International Nuclear Information System (INIS)

    Yoshida, Tadashi

    1999-01-01

    The Delayed Neutron Working Group was established in April 1997 within the Nuclear Data Subcommittee of JNDC. It has two principal missions. One is to coordinate the Japanese activities toward the WPEC/Subgroup-6 efforts, and the other is to recommend the delayed neutron data for JENDL-3.3. The final report of Subgroup-6, which in one of the subgroups of the NEA International Evaluation Cooperation (WPEC) and is in charge of the delayed neutron data, is to be completed in 1999. Here in Japan, JENDL-3.3 is planned to be released in early 2000. Delayed Neutron Working Group is, then, going to finalize its activity by the end of the fiscal year 1999 after recommending appropriate sets of data as coherently as possible with the of Subgroup-6 efforts. (author)

  17. $\\beta$-delayed neutron spectroscopy of $^{130-132}$ Cd isotopes with the ISOLDE decay station and the VANDLE array

    CERN Multimedia

    We propose to use the new ISOLDE decay station and the neutron detector VANDLE to measure the $\\beta$-delayed neutron emission of N=82-84 $^{130-132}$Cd isotopes. The large delayed neutron emission probability observed in a previous ISOLDE measurement is indicative of the Gamow-Teller transitions due to the decay of deep core neutrons. Core Gamow-Teller decay has been experimentally proven in the $^{78}$Ni region for the N>50 nuclei using the VANDLE array. The spectroscopic measurement of delayed neutron emission along the cadmium isotopic chain will allow us to track the evolution of the single particle states and the shell gap.

  18. Spatial and energy distributions of skyshine neutron and gamma radiation from nuclear reactors on the ground-air boundary

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Y.; Netecha, M.E.; Vasiliev, A.P.; Avaev, V.N.; Vasiliev, G.A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Zelensky, D.I.; Istomin, Y.L.; Cherepnin, Y.S. [Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan, Semipalatinsk-21 (Kazakhstan); Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-03-01

    A set of measurements on skyshine radiation was conducted at two special research reactors. A broad range of detectors was used in the measurements to record neutron and gamma radiations. Dosimetric and radiometric field measurements of the neutrons and gamma quanta of the radiation scattered in the air were performed at distances of 50 to 1000 m from the reactor during different weather conditions. The neutron spectra in the energy range of 1 eV to 10 MeV and the gamma quanta spectra in the range of 0.1-10 MeV were measured. (author)

  19. Neutron spectra of /sup 242/Cm-Be and /sup 244/Cm-Be neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A; Nagarajan, P S [Bhabha Atomic Research Centre, Bombay (India). Div. of Radiation Protection

    1977-02-15

    Neutron spectra of /sup 242/Cm-Be(..cap alpha..,n) and /sup 244/Cm-Be(..cap alpha..,n) sources have been calculated including the spontaneous fission contribution which is negligible for /sup 242/Cm and amounts to about 4% for /sup 244/Cm. The agreement of the present work with experimental results is poor.

  20. 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

  1. Use of borated polyethylene to improve low energy response of a prompt gamma based neutron dosimeter

    Energy Technology Data Exchange (ETDEWEB)

    Priyada, P.; Ashwini, U.; Sarkar, P.K., E-mail: pradip.sarkar@manipal.edu

    2016-05-21

    The feasibility of using a combined sample of borated polyethylene and normal polyethylene to estimate neutron ambient dose equivalent from measured prompt gamma emissions is investigated theoretically to demonstrate improvements in low energy neutron dose response compared to only polyethylene. Monte Carlo simulations have been carried out using the FLUKA code to calculate the response of boron, hydrogen and carbon prompt gamma emissions to mono energetic neutrons. The weighted least square method is employed to arrive at the best linear combination of these responses that approximates the ICRP fluence to dose conversion coefficients well in the energy range of 10{sup −8} MeV to 14 MeV. The configuration of the combined system is optimized through FLUKA simulations. The proposed method is validated theoretically with five different workplace neutron spectra with satisfactory outcome. - Highlights: • An improved method is proposed for estimating H⁎(10) using prompt gamma emissions. • A combination of BHDPE and HDPE cylinders is used as a sample. • Linear combination of prompt gamma intensities approximates ICRP-DCC closely. • Feasibility of the method was tested theoretically using workplace neutron spectra.

  2. Nuclear and activation characteristics of materials in 14.1-MeV and 2.5-MeV neutron field

    International Nuclear Information System (INIS)

    Seki, Yasushi; Takeyasu, Yuuichi.

    1988-11-01

    The nuclear and activation characteristics of various materials and elements of interest in terms of fusion reactor design are calculated and the results are graphically shown. The elements and materials are placed in a simple geometry modelling a blanket and shield of a fusion reactor. The neutrons with 14.1-MeV and 2.5-MeV energy are generated from the region represented as D-T and D-D plasma, respectively. The following activation characteristics after neutron irradiation are shown for each material and element; 1. Time evolution of induced activity, 2. Time evolution of decay heat, 3. Delayed gamma-ray dose distribution, 4. Decay heat distribution. In addition to the above activation characteristics, nuclear characteristics during the neutron irradiation, e.g. neutron energy spectra, neutron and gamma-ray flux distribution, nuclear heating distributions, and neutron and gamma-ray dose rate are also shown. (author)

  3. Neutron spectra measuring by magnetless hadron spectrometer

    International Nuclear Information System (INIS)

    Bayukov, Yu.D.; Buklej, A.E.; Gavrilov, V.B.

    1980-01-01

    The energy resolution, efficiency and background conditions of neutron recording in inclusive nuclear reactions by a magnetless hadron spectrometer (MHS) in the 8-300 MeV energy range. The scheme of apparatus lay-out for measuring neutron recording efficiency is shown. For recording colliding particles with the 3 GeV/c momentum four beam scintillation counters, the latter of which of 30x40 mm dimensions and 1 mm thickness defines the working beam range in the target centre, are used. Targets of the 80 mm diameter made of C and Pb (2.08 g/cm 2 and 3.04 g/cm 2 thickness, respectively) are located at the 45 deg angle in respect to the beam direction. Secondary particles escaping at the 90 deg angle are recorded by two telescopes of the scintillation counters. For neutron and γ quanta recording the special scintillation detector of the 20 cm thickness encircled by an anticoincidence counter is used. The neutron recording efficiency is determined on the basis of comparison of the neutron production differential cross sections of the π +- 12 C 6 → nX reactions and of proton production in isotopically symmetric reactions π +- 12 C 6 → pX. The experimental data are in good agreement with the calculation data [ru

  4. New data on prefission neutrons

    International Nuclear Information System (INIS)

    Boikov, G.S.; Dmitriev, V.D.; Kudyaev, G.A.; Ostapenko, Yu.B.; Svirin, M.I.; Smirenkin, G.N.

    1991-01-01

    The spectra of neutrons emitted in fission of 232 Th, 235 U and 238 U induced by 2.9 and 14.7 MeV neutrons (below and above the chance fission threshold, respectively) were measured by the time-of-flight method. Two effects were observed in the prefission neutron spectra: the high-energy wing is related to the nonequilibrium mechanism of emission up to the well pronounced upper boundary of ε max = 8.5 MeV; in the lower-energy wing ε < 2 MeV, neutron yield exceeds conventional statistical model description. The latter effect was attributed to the fission process dynamics. (author). 18 refs, 3 figs, 2 tabs

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

    International Nuclear Information System (INIS)

    Camargo, Dayana Q. de; Bodmann, Bardo E.J.; Vilhena, Marco T. de; Froehlich, Herberth B.

    2011-01-01

    In this work we developed a stochastic model to simulate neutron transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using the Monte Carlo method for the propagation of neutrons in different environments. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational time we introduced a variable control volume together with (pseudo-) periodic boundary conditions in order to overcome this problem. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude that there is need to consider the energy dependence and hence defined a spectral effective multiplication factor per Monte Carlo step. (author)

  6. 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

  7. FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra

    International Nuclear Information System (INIS)

    LaBauve, R.J.; George, D.C.; England, T.R.

    1980-03-01

    The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from 235 U irradiated with a pulse (10 -4 s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables

  8. Peculiar time dependence of unexpected lines in delayed beam-foil X-ray spectra of V, Fe and Ni

    International Nuclear Information System (INIS)

    Ahmad, Nissar; Karn, Ranjeet K.; Marketos, Pan; Nandi, T.

    2005-01-01

    Delayed beam-foil X-ray spectra of highly charged ions of V, Fe and Ni show a few lines at energies higher than the H-like Lyman α-line of the respective projectile ions. These can only be attributed to heavier ions. Further the time dependence of such unexpected lines display a peculiar behavior. This work presents the experimental observations systematically

  9. Effect of absorption discontinuity on neutron spectra of water assemblies poisoned with non-1/V absorbers

    International Nuclear Information System (INIS)

    Gupta, I.J.; Trikha, S.K.

    1977-01-01

    Calculations are presented of the diffusion of thermal neutrons (2.5 x 10 -4 to 7 x 10 -1 eV) across an absorption discontinuity in a water assembly, consisting of pure water on one side and aqueous solutions of three different non-1/V absorbers on the other, which were obtained by solving the Boltzmann transport equation in the diffusion approximation using the multigroup formalism. The gradual appearance and disappearance of the depletion region in the neutron spectra (caused by the resonance absorption peaks at energies 0.096 and 0.179 eV for samarium and cadmium respectively), as one moves from the pure water assembly to the poisoned water assembly and vice versa, have also been studied. The minimum concentrations of Sm and Cd atoms in water for which the depletion region in the spectra just starts building up are found to be 60 x 10 18 Sm atom cm -3 and 125 x 10 18 Cd atom cm -3 respectively. However no such depletion region is observed in gadolinium-poisoned water assembly. At the boundary, the equilibrium neutron distribution gets disturbed and is re-established to the equilibrium distribution of the second medium at some distance from the interface. The diffusion lengths so calculated from the total neutron density curves are in good agreement with the experimental results of Goddard and Johnson (Nucl. Sci. Eng.; 37:127 (1969)) at various concentrations of Gd and Cd atoms in water. (author)

  10. Errors in estimating neutron quality factor using lineal energy distributions measured in tissue-equivalent proportional counters

    International Nuclear Information System (INIS)

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

    1982-01-01

    Neutron dose equivalent is obtained from quality factors which are defined in terms of LET. It is possible to estimate the dose averaged quality factor, antiQ, directly from distributions in lineal energy, y, that are measured in tissue-equivalent proportional counters. This eliminates a mathematical transformation of the absorbed dose from D(y) to D(L). We evaluate the inherent error in computing Q from D(y) rather than D(L) for neutron spectra below 4 MeV. The effects of neutron energy and simulated tissue diameters within a gas cavity are examined in detail. (author)

  11. Theoretical and Experimental Research in Neutron Spectra and Nuclear Waste Transmutation on Fast Subcritical Assembly with MOX Fuel

    Science.gov (United States)

    Arkhipkin, D. A.; Buttsev, V. S.; Chigrinov, S. E.; Kutuev, R. Kh.; Polanski, A.; Rakhno, I. L.; Sissakian, A.; Zulkarneev, R. Ya.; Zulkarneeva, Yu. R.

    2003-07-01

    The paper deals with theoretical and experimental investigation of transmutation rates for a number of long-lived fission products and minor actinides, as well as with neutron spectra formed in a subcritical assembly driven with the following monodirectional beams: 660-MeV protons and 14-MeV neutrons. In this work, the main objective is the comparison of neutron spectra in the MOX assembly for different external driving sources: a 660-MeV proton accelerator and a 14-MeV neutron generator. The SAD project (JINR, Russia) has being discussed. In the context of this project, a subcritical assembly consisting of a cylindrical lead target surrounded by a cylindrical MOX fuel layer will be constructed. Present conceptual design of the subcritical assembly is based on the core with a nominal unit capacity of 15 kW (thermal). This corresponds to a multiplication coefficient, keff= 0.945, and an accelerator beam power of 0.5 kW. The results of theoretical investigations on the possibility of incinerating long-lived fission products and minor actinides in fast neutron spectrum and formation of neutron spectra with different hardness in subcritical systems based on the MOX subcritical assembly are discussed. Calculated neutron spectra emitted from a lead target irradiated by a 660-MeV protons are also presented.

  12. 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

  13. Bioassay method for Uranium in urine by Delay Neutron counting

    International Nuclear Information System (INIS)

    Suratman; Purwanto; Sukarman-Aminjoyo

    1996-01-01

    A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 μg

  14. Monte Carlo simulation of fast neutron scattering experiments including DD-breakup neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, D.; Siebert, B.R.L.

    1993-06-01

    The computational simulation of the deuteron breakup in a scattering experiment has been investigated. Experimental breakup spectra measured at 16 deuteron energies and at 7 angles for each energy served as the data base. Analysis of these input data and of the conditions of the scattering experiment made it possible to reduce the input data. The use of one weighted breakup spectrum is sufficient to simulate the scattering spectra at one incident neutron energy. A number of tests were carried out to prove the validity of this result. The simulation of neutron scattering on carbon, including the breakup, was compared with measured spectra. Differences between calculated and measured spectra were for the most part within the experimental uncertainties. Certain significant deviations can be attributed to erroneous scattering cross sections taken from an evaluation and used in the simulation. Scattering on higher-lying states in [sup 12]C can be analyzed by subtracting the simulated breakup-scattering from the experimental spectra. (orig.)

  15. Monte Carlo simulation of fast neutron scattering experiments including DD-breakup neutrons

    International Nuclear Information System (INIS)

    Schmidt, D.; Siebert, B.R.L.

    1993-06-01

    The computational simulation of the deuteron breakup in a scattering experiment has been investigated. Experimental breakup spectra measured at 16 deuteron energies and at 7 angles for each energy served as the data base. Analysis of these input data and of the conditions of the scattering experiment made it possible to reduce the input data. The use of one weighted breakup spectrum is sufficient to simulate the scattering spectra at one incident neutron energy. A number of tests were carried out to prove the validity of this result. The simulation of neutron scattering on carbon, including the breakup, was compared with measured spectra. Differences between calculated and measured spectra were for the most part within the experimental uncertainties. Certain significant deviations can be attributed to erroneous scattering cross sections taken from an evaluation and used in the simulation. Scattering on higher-lying states in 12 C can be analyzed by subtracting the simulated breakup-scattering from the experimental spectra. (orig.)

  16. $\\gamma$-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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-03-15

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

  18. Study on method of dose estimation for the Dual-moderated neutron survey meter

    International Nuclear Information System (INIS)

    Zhou, Bo; Li, Taosheng; Xu, Yuhai; Gong, Cunkui; Yan, Qiang; Li, Lei

    2013-01-01

    In order to study neutron dose measurement in high energy radiation field, a Dual-moderated survey meter in the range from 1 keV to 300 MeV mean energies spectra has been developed. Measurement results of some survey meters depend on the neutron spectra characteristics in different neutron radiation fields, so the characteristics of the responses to various neutron spectra should be studied in order to get more reasonable dose. In this paper the responses of the survey meter were calculated under different neutron spectra data from IAEA of Technical Reports Series No. 318 and other references. Finally one dose estimation method was determined. The range of the reading per H*(10) for the method estimated is about 0.7–1.6 for the neutron mean energy range from 50 keV to 300 MeV. -- Highlights: • We studied a novel high energy neutron survey meter. • Response characteristics of the survey meter were calculated by using a series of neutron spectra. • One significant advantage of the survey meter is that it can provide mean energy of radiation field. • Dose estimate deviation can be corrected. • The range of corrected reading per H*(10) is about 0.7–1.6 for the neutron fluence mean energy range from 0.05 MeV to 300 MeV

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  1. Reconstruction of neutron spectra using neural networks starting from the Bonner spheres spectrometric system; Reconstruccion de espectros de neutrones usando redes neuronales a partir del sistema espectrometrico de esferas de Bonner

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz R, J.M.; Martinez B, M.R.; Arteaga A, T.; Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E. [Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico)

    2005-07-01

    The artificial neural networks (RN) have been used successfully to solve a wide variety of problems. However to determine an appropriate set of values of the structural parameters and of learning of these, it continues being even a difficult task. Contrary to previous works, here a set of neural networks is designed to reconstruct neutron spectra starting from the counting rates coming from the detectors of the Bonner spheres system, using a systematic and experimental strategy for the robust design of multilayer neural networks of the feed forward type of inverse propagation. The robust design is formulated as a design problem of Taguchi parameters. It was selected a set of 53 neutron spectra, compiled by the International Atomic Energy Agency, the counting rates were calculated that would take place in a Bonner spheres system, the set was arranged according to the wave form of those spectra. With these data and applying the Taguchi methodology to determine the best parameters of the network topology, it was trained and it proved the same one with the spectra. (Author)

  2. Basics of Neutrons for First Responders

    Energy Technology Data Exchange (ETDEWEB)

    Rees, Brian G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-02-05

    These are slides from a presentation on the basics of neutrons. A few topics covered are: common origins of terrestrial neutron radiation, neutron sources, neutron energy, interactions, detecting neutrons, gammas from neutron interactions, neutron signatures in gamma-ray spectra, neutrons and NaI, neutron fluence to dose (msV), instruments' response to neutrons.

  3. A wide-range direction neutron spectrometer

    International Nuclear Information System (INIS)

    Luszik-Bhadra, M.; D'Errico, F.; Hecker, O.; Matzke, M.

    2002-01-01

    A new device is presented which has been developed for measuring the energy and direction of distribution of neutron fluence in fields of broad energy spectra (thermal to 100 MeV) and with a high background of photon, electron and muon radiation. The device was tested in reference fields with different energy and direction distributions of neutron fluence. The direction-integrated fluence spectra agree fairly well with reference spectra. In all cases, the ambient and personal dose equivalent values calculated from measured direction-differential spectra are within 35% of the reference values. Independent measurements of the directional dose equivalent were performed with a directional dose equivalent monitor based on superheated drop detectors

  4. High energy fast neutrons from the Harwell variable energy cyclotron. II. Biologic studies in mammalian systems

    International Nuclear Information System (INIS)

    Berry, R.J.; Bance, D.A.; Barnes, D.W.H.; Cox, R.; Goodhead, D.T.; Sansom, J.M.; Thacker, J.

    1977-01-01

    A high energy fast neutron beam potentially suitable for radiotherapy has been described in a companion paper. Its biologic effects have been studied in the following experimental systems: clonal survival and mutation induction after irradiation in vitro in Chinese hamster cells and human diploid fibroblasts; survival of reproductive capacity in vivo of murine hemopoietic colony-forming cells and murine intestinal crypts after irradiation in vivo; survival of reproductive capacity in vivo after irradiation in vitro or in vivo of murine lymphocytic leukemia cells; acute intestinal death following total body irradiation of mice and guinea pigs; and hemopoietic death following total body irradiation of mice and guinea pigs. The relative biologic effectiveness of these high energy neutrons varied among the different biologic systems, and in several cases varied with the size of the radiation dose. The oxygen enhancement ratio was studied in murine lymphocytic leukemia cells irradiated under aerobic or hypoxic conditions in vitro and assayed for survival of reproductive capacity in vivo. Compared with x-rays, the potential therapeutic gain factor for these neutrons was about 1.5. This work represents a ''radiobiologic calibration'' program which it is suggested should be undertaken before new and unknown fast neutron spectra are used for experimental radiotherapy. The results are compared with biologic studies carried out at high energy fast neutron generators in the United States

  5. Measurement of prompt fission gamma-ray spectra in fast neutron-induced fission

    International Nuclear Information System (INIS)

    Laborie, J.M.; Belier, G.; Taieb, J.

    2012-01-01

    Knowledge of prompt fission gamma-ray emission has been of major interest in reactor physics for a few years. Since very few experimental spectra were ever published until now, new measurements would be also valuable to improve our understanding of the fission process. An experimental method is currently being developed to measure the prompt fission gamma-ray spectrum from some tens keV up to 10 MeV at least. The mean multiplicity and total energy could be deduced. In this method, the gamma-rays are measured with a bismuth germanate (BGO) detector which has the advantage to present a high P/T ratio and a high efficiency compared to other gamma-ray detectors. The prompt fission neutrons are rejected by the time of flight technique between the BGO detector and a fission trigger given by a fission chamber or a scintillating active target. Energy and efficiency calibration of the BGO detector were carried out up to 10.76 MeV by means of the Al-27(p, gamma) reaction. First prompt fission gamma-ray spectrum measurements performed for the spontaneous fission of Cf-252 and for 1.7 and 15.6 MeV neutron-induced fission of U-238 at the CEA, DAM, DIF Van de Graaff accelerator, will be presented. (authors)

  6. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    International Nuclear Information System (INIS)

    Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

    2010-01-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  7. Application of semi-empirical modeling and non-linear regression to unfolding fast neutron spectra from integral reaction rate data

    International Nuclear Information System (INIS)

    Harker, Y.D.

    1976-01-01

    A semi-empirical analytical expression representing a fast reactor neutron spectrum has been developed. This expression was used in a non-linear regression computer routine to obtain from measured multiple foil integral reaction data the neutron spectrum inside the Coupled Fast Reactivity Measurement Facility. In this application six parameters in the analytical expression for neutron spectrum were adjusted in the non-linear fitting process to maximize consistency between calculated and measured integral reaction rates for a set of 15 dosimetry detector foils. In two-thirds of the observations the calculated integral agreed with its respective measured value to within the experimental standard deviation, and in all but one case agreement within two standard deviations was obtained. Based on this quality of fit the estimated 70 to 75 percent confidence intervals for the derived spectrum are 10 to 20 percent for the energy range 100 eV to 1 MeV, 10 to 50 percent for 1 MeV to 10 MeV and 50 to 90 percent for 10 MeV to 18 MeV. The analytical model has demonstrated a flexibility to describe salient features of neutron spectra of the fast reactor type. The use of regression analysis with this model has produced a stable method to derive neutron spectra from a limited amount of integral data

  8. Benchmark experiments of effective delayed neutron fraction βeff at FCA

    International Nuclear Information System (INIS)

    Sakurai, Takeshi; Okajima, Shigeaki

    1999-01-01

    Benchmark experiments of effective delayed neutron fraction β eff were performed at Fast Critical Assembly (FCA) in the Japan Atomic Energy Research Institute. The experiments were made in three cores providing systematic change of nuclide contribution to the β eff : XIX-1 core fueled with 93% enriched uranium, XIX-2 core fueled with plutonium and uranium (23% enrichment) and XIX-3 core fueled with plutonium (92% fissile Pu). Six organizations from five countries participated in these experiments and measured the β eff by using their own methods and instruments. Target accuracy in the β eff was achieved to be better than ±3% by averaging the β eff values measured using a wide variety of experimental methods. (author)

  9. Reduction of delayed-neutron contribution to variance-to-mean ratio by application of difference filter technique

    International Nuclear Information System (INIS)

    Hashimoto, Kengo; Mouri, Tomoaki; Ohtani, Nobuo

    1999-01-01

    The difference-filtering correlation analysis was applied to time-sequence neutron count data measured in a slightly subcritical assembly, where the Feynman-α analysis suffered from large contribution of delayed neutron to the variance-to-mean ratio of counts. The prompt-neutron decay constant inferred from the present filtering analysis agreed very closely with that by pulsed neutron experiment, and no dependence on the gate-time range specified could be observed. The 1st-order filtering was sufficient for the reduction of the delayed-neutron contribution. While the conventional method requires a choice of analysis formula appropriate to a gate-time range, the present method is applicable to a wide variety of gate-time ranges. (author)

  10. Neutron response study

    International Nuclear Information System (INIS)

    Endres, G.W.R.; Fix, J.J.; Thorson, M.R.; Nichols, L.L.

    1981-01-01

    Neutron response of the albedo type dosimeter is strongly dependent on the energy of the incident neutrons as well as the moderating material on the backside of the dosimeter. This study characterizes the response of the Hanford dosimeter for a variety of neutron energies for both a water and Rando phantom (a simulated human body consisting of an actual human skeleton with plastic for body muscles and certain organs). The Hanford dosimeter response to neutrons of different energies is typical of albedo type dosimeters. An approximate two orders of magnitude difference in response is observed between neutron energies of 100 keV and 10 MeV. Methods were described to compensate for the difference in dosimeter response between a laboratory neutron spectrum and the different spectra encountered at various facilities in the field. Generally, substantial field support is necessary for accurate neutron dosimetry

  11. Spectra and neutron dose of an 18 MV Linac using two geometric models of the head

    International Nuclear Information System (INIS)

    Barrera, M. T.; Pino, F.; Barros, H.; Sajo-Bohus, L.; Davila, J.; Salcedo, E.; Vega C, H. R.; Benites R, J. L.

    2015-10-01

    Full text: Using the Monte Carlo method, by MCNP5 code, simulations were performed with different source terms and 2 geometric models of the head to obtain spectra in energy, flow and doses of photo-neutrons at different positions on the stretcher and in the radiotherapy room. The simplest model was a spherical shell of tungsten; the second was the complete model of a heterogeneous head of an accelerator Varian ix. In both models Tosi function was used as a source term. In addition, for the second model Sheikh-Bagheri distribution was used for photons and photo-neutrons were generated. Also in both models the radiotherapy room of Gurve group of the Teaching Medical Center La Trinidad was included, which is equipped with an accelerator Varian Clinic 2100. In this Center passive detectors PADC (Cr-39) were irradiated with neutron converters, with 18 MeV photons radiation. The measured neutron flow was compared with that obtained with Monte Carlo calculations. The Monte Carlo flows are similar to those measured at the isocenter. The simplest model underestimates the neutron flow compared with the calculated flows with the heterogeneous model of the head. (Author)

  12. High energy neutron generator

    International Nuclear Information System (INIS)

    Barjon, R.; Breynat, G.

    1987-01-01

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

  13. RMF+BCS description of some traditional neutron magic isotones

    International Nuclear Information System (INIS)

    Saxena, G.; Singh, D.; Kaushik, M.

    2014-01-01

    The traditional neutron magic nuclei with N = 8, 20, 28, 50, 82 and 126, and those with neutron sub-magic number N = 40 are investigated within the relativistic mean-field plus BCS (RMF+BCS) approach. The results indicate appearance of new proton magic numbers as well as the disappearance of conventional magic numbers for nuclei with extreme isospin values. The calculated energies and densities do not indicate any tendency for the proton halo formations in any of the proton rich isotones due to Coulomb interaction and different single particle spectra. However, the potential barrier provided by the Coulomb interaction and that due to the centrifugal force may cause a long delay in the actual decay of proton rich nucleus resulting in the extended drip line. (authors)

  14. Study on calculation methods for the effective delayed neutron fraction

    International Nuclear Information System (INIS)

    Irwanto, Dwi; Obara, Toru; Chiba, Go; Nagaya, Yasunobu

    2011-03-01

    The effective delayed neutron fraction β eff is one of the important neutronic parameters from a view point of a reactor kinetics. Several Monte-Carlo-based methods to estimate β eff have been proposed to date. In order to quantify the accuracy of these methods, we study calculation methods for β eff by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate β eff with various definitions such as the fundamental value β 0 , the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other. Through the present study, we find the following: The largest difference among the standard definition of β eff , Nauchi's β eff and Meulekamp's β eff is approximately 10%. The fundamental value β 0 is quite larger than the others in several cases. For all the cases, Meulekamp's β eff is always higher than Nauchi's β eff . This is because Nauchi's β eff considers the average neutron multiplicity value per fission which is large in the high energy range (1MeV-10MeV), while the definition of Meulekamp's β eff does not include this parameter. Furthermore, we evaluate the multi-generation effect on β eff values and demonstrate that this effect should be considered to obtain the standard definition values of β eff . (author)

  15. Artificial neural networks in neutron dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.; Mercado, G.A.; Perales M, W.A.; Robles R, J.A. [Unidades Academicas de Estudios Nucleares, UAZ, A.P. 336, 98000 Zacatecas (Mexico); Gallego, E.; Lorente, A. [Depto. de Ingenieria Nuclear, Universidad Politecnica de Madrid, (Spain)

    2005-07-01

    An artificial neural network has been designed to obtain the neutron doses using only the Bonner spheres spectrometer's count rates. Ambient, personal and effective neutron doses were included. 187 neutron spectra were utilized to calculate the Bonner count rates and the neutron doses. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra, UTA4 response matrix and fluence-to-dose coefficients were used to calculate the count rates in Bonner spheres spectrometer and the doses. Count rates were used as input and the respective doses were used as output during neural network training. Training and testing was carried out in Mat lab environment. The artificial neural network performance was evaluated using the {chi}{sup 2}- test, where the original and calculated doses were compared. The use of Artificial Neural Networks in neutron dosimetry is an alternative procedure that overcomes the drawbacks associated in this ill-conditioned problem. (Author)

  16. Artificial neural networks in neutron dosimetry

    International Nuclear Information System (INIS)

    Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.; Mercado, G.A.; Perales M, W.A.; Robles R, J.A.; Gallego, E.; Lorente, A.

    2005-01-01

    An artificial neural network has been designed to obtain the neutron doses using only the Bonner spheres spectrometer's count rates. Ambient, personal and effective neutron doses were included. 187 neutron spectra were utilized to calculate the Bonner count rates and the neutron doses. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra, UTA4 response matrix and fluence-to-dose coefficients were used to calculate the count rates in Bonner spheres spectrometer and the doses. Count rates were used as input and the respective doses were used as output during neural network training. Training and testing was carried out in Mat lab environment. The artificial neural network performance was evaluated using the χ 2 - test, where the original and calculated doses were compared. The use of Artificial Neural Networks in neutron dosimetry is an alternative procedure that overcomes the drawbacks associated in this ill-conditioned problem. (Author)

  17. Experimental study on neutronics in bombardment of thick targets by high energy proton beams for accelerator-driven sub-critical system

    CERN Document Server

    Guo Shi Lun; Shi Yong Qian; Shen Qing Biao; Wan Jun Sheng; Brandt, R; Vater, P; Kulakov, B A; Krivopustov, M I; Sosnin, A N

    2002-01-01

    The experimental study on neutronics in the target region of accelerator-driven sub-critical system is carried out by using the high energy accelerator in Joint Institute for Nuclear Research, Dubna, Russia. The experiments with targets U(Pb), Pb and Hg bombarded by 0.533, 1.0, 3.7 and 7.4 GeV proton beams show that the neutron yield ratio of U(Pb) to Hg and Pb to Hg targets is (2.10 +- 0.10) and (1.76 +- 0.33), respectively. Hg target is disadvantageous to U(Pb) and Pb targets to get more neutrons. Neutron yield drops along 20 cm thick targets as the thickness penetrated by protons increases. The lower the energy of protons, the steeper the neutron yield drops. In order to get more uniform field of neutrons in the targets, the energy of protons from accelerators should not be lower than 1 GeV. The spectra of secondary neutrons produced by different energies of protons are similar, but the proportion of neutrons with higher energy gradually increases as the proton energy increases

  18. Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

    2012-10-01

    Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

  19. Cyclotron Lines in Accreting Neutron Star Spectra

    Science.gov (United States)

    Wilms, Jörn; Schönherr, Gabriele; Schmid, Julia; Dauser, Thomas; Kreykenbohm, Ingo

    2009-05-01

    Cyclotron lines are formed through transitions of electrons between discrete Landau levels in the accretion columns of accreting neutron stars with strong (1012 G) magnetic fields. We summarize recent results on the formation of the spectral continuum of such systems, describe recent advances in the modeling of the lines based on a modification of the commonly used Monte Carlo approach, and discuss new results on the dependence of the measured cyclotron line energy from the luminosity of transient neutron star systems. Finally, we show that Simbol-X will be ideally suited to build and improve the observational database of accreting and strongly magnetized neutron stars.

  20. Power measurement in the boiling capsules in R2 using delayed neutron detector

    International Nuclear Information System (INIS)

    Roennberg, G.

    1979-03-01

    LWR fuel testing is performed in the R2 reactor by irradiation in both loops and so-called boiling capsules. The loops have forced cooling, and the power can be measured calorimetrically by conventional instrumentation. The boiling capsules have convection cooling, and it has therefore been necessary to develop a special technique for power measurement, the delayed neutron detector (DND). The DND is a pneumatic rabbit system, which activates small uranium samples in the boiling capsules and counts the delayed neutrons for determination of the fission rate. This report describes the equipment used, the procedure of measurement, and the method of evaluation. (atuhor)

  1. Study of the delayed neutron emission through the time-of-flight method. Application to 49K, 50K and 51K

    International Nuclear Information System (INIS)

    Rachidi, J.

    1983-04-01

    This work is dedicated to the study of the emission of delayed neutrons observed in the decay of 49 K, 50 K and 51 K. Spectroscopic data are non-existent for these 3 isotopes, so we have had to design a specific detection system based on a large-surface scintillation counter. A series of n-γ coincidence measurement has allowed us to determine the energy levels of the non-bound states of 49 Ca, 50 Ca and 51 Ca and to establish the nature of the beta transitions (K → Ca). We have measured the energy of the delayed neutrons through the time-of-flight method. Our results are consistent with the model of the p-n states based on the Bansac-French's works. This model shows that the non-bound states of the calcium isotopes discovered in the experiment are represented by simple configurations of the (sd) -1 (fp) n type. (A.C.)

  2. Pilot study for the implantation of a high-energy neutrons field

    International Nuclear Information System (INIS)

    Pinto, Jose Julio de O.; Mendes, Adriane C.; Federico, Claudio A.; Passaro, Angelo; Gaspar, Felipe de B.; Pazianotto, Mauricio T.

    2013-01-01

    In this work a theoretical study is presented for the implementation of a high-energy neutron field (14.1 MeV) produced by a neutron generator type DT (deuterium-tritium), to be installed in the premises of the Laboratorio de Radiacoes Ionizantes (LRI) of the Instituto de Estudos Avancados (IEAv). This evaluation was performed by means of computer simulation by Monte Carlo method, using the computer code MCNP5 (Monte Carlo N-Particle). The neutron spectra were simulated computationally for pre-selected points of the installation, allowing to estimate the beam quality in the positions provided for use of the direct beam. These simulations also allow assist the basement of a project to install the consistent D-T generator with the guidelines for radiation protection and radiation safety standards determined by the Comissao Nacional de Energia Nuclear (CNEN), by estimating the dose rates provided in accessible points to Individuals Occupationally Exposed (IOE) in the facility. The computational determination of spectra, fluxes and doses produced in different positions previously selected within and outside the laboratory, will serve as guidance from previous studies for the future installation of this generator in the physical facilities of the LRI

  3. Statistical analysis for discrimination of prompt gamma ray peak induced by high energy neutron: Monte Carlo simulation study

    International Nuclear Information System (INIS)

    Do-Kun Yoon; Joo-Young Jung; Tae Suk Suh; Seong-Min Han

    2015-01-01

    The purpose of this research is a statistical analysis for discrimination of prompt gamma ray peak induced by the 14.1 MeV neutron particles from spectra using Monte Carlo simulation. For the simulation, the information of 18 detector materials was used to simulate spectra by the neutron capture reaction. The discrimination of nine prompt gamma ray peaks from the simulation of each detector material was performed. We presented the several comparison indexes of energy resolution performance depending on the detector material using the simulation and statistics for the prompt gamma activation analysis. (author)

  4. Response of Inorganic Scintillators to Neutrons of 3 and 15 MeV Energy

    CERN Document Server

    Lucchini, M; Pizzichemi, M; Chipaux, R; Jacquot, F; Mazue, H; Wolff, H; Lecoq, P; Auffray, E

    2014-01-01

    In the perspective of the development of future high energy physics experiments, homogeneous calorimeters based on inorganic scintillators can be considered for the detection of hadrons (e.g., calorimeter based on dual-readout technique). Although of high importance in the high energy physics framework as well as for homeland security applications, the response of these inorganic scintillators to neutrons has been only scarcely investigated. This paper presents results obtained using five common scintillating crystals (of size around 2x2x2 cm 3), namely lead tungstate (PbWO4), bismuth germanate (BGO), cerium fluoride (CeF3), Ce-doped lutetium-yttrium orthosilicate (LYSO:Ce) and lutetium aluminum garnet (LuAG:Ce) in a pulsed flux of almost mono-energetic (similar to 3 MeV and similar to 15 MeV) neutrons provided by the Van de Graff accelerator SAMES of CEA Valduc. Energy spectra have been recorded, calibrated and compared with Geant4 simulations computed with different physics models. The neutron detection eff...

  5. BASACF, Integral Neutron Spectra Adjustment and Dosimetry

    International Nuclear Information System (INIS)

    Tichy, Milos

    1996-01-01

    1 - Description of program or function: Adjustment of a neutron spectrum based on integral detector measurements and calculation of an integral dosimetric quantity (integral flux, d.p.a., dose equivalent) and its variance. The program requires measured data (activities and their covariance matrix) and a priori information (spectrum, dosimetry cross sections, integral quantity conversion factor and their covariance matrices). All a priori covariance matrices can be read in from a file prepared by some other code or can be generated by means of three different methods (by subroutines included in the program). A subroutine which can normalize the a priori flux to measured data is also included. The program provides also adjusted dosimetry cross sections (with covariance matrix) so that it can be used for an adjustment of cross sections (or response functions of e.g. Bonner balls) by measurements in well-known neutron spectra. 2 - Method of solution: Bayesian theorem on conditional probability applied to linearized relation between activities, dosimetry cross sections and flux. All probability distributions are supposed to be normal and this supposition leads to minimizing of the same functional as least squares method (STAY'SL). This task is solved by a covariance filter method which avoids any matrix inversion and is numerically robust and stable. 3 - Restrictions on the complexity of the problem: This version can use 45 energy groups and 5 detectors and occupies 310 kB of main memory. This restriction can be modified according to available memory. The covariance matrix of activities is supposed diagonal. A solution is produced for any set of input data but in the case of non-consistent data, when measured activities do not match the a priori flux, the solution is not very meaningful

  6. ULX spectra revisited: Accreting, highly magnetized neutron stars as the engines of ultraluminous X-ray sources

    Science.gov (United States)

    Koliopanos, Filippos; Vasilopoulos, Georgios; Godet, Olivier; Bachetti, Matteo; Webb, Natalie A.; Barret, Didier

    2017-12-01

    Aims: In light of recent discoveries of pulsating ultraluminous X-ray sources (ULXs) and recently introduced theoretical schemes that propose neutron stars (NSs) as the central engines of ULXs, we revisit the spectra of eighteen well known ULXs, in search of indications that favour this newly emerging hypothesis. Methods: We examine the spectra from high-quality XMM-Newton and NuSTAR observations. We use a combination of elementary black body and multicolour disk black body (MCD) models, to diagnose the predictions of classic and novel theoretical models of accretion onto NSs. We re-interpret the well established spectral characteristics of ULXs in terms of accretion onto lowly or highly magnetised NSs, and explore the resulting parameter space for consistency. Results: We confirm the previously noted presence of the low-energy (≲6 keV) spectral rollover and argue that it could be interpreted as due to thermal emission. The spectra are well described by a double thermal model consisting of a "hot" (≳1 keV) and a "cool" (≲0.7 keV) multicolour black body (MCB). Under the assumption that the "cool" MCD emission originates in a disk truncated at the neutron star magnetosphere, we find that all ULXs in our sample are consistent with accretion onto a highly magnetised (B ≳ 1012 G) neutron star. We note a strong correlation between the strength of the magnetic field, the temperature of the "hot" thermal component and the total unabsorbed luminosity. Examination of the NuSTAR data supports this interpretation and also confirms the presence of a weak, high-energy (≳15 keV) tail, most likely the result of modification of the MCB emission by inverse Compton scattering. We also note that the apparent high-energy tail, may simply be the result of mismodelling of MCB emission with an atypical temperature (T) versus radius (r) gradient, using a standard MCD model with a fixed gradient of T r-0.75. Conclusions: We have offered a new and robust physical interpretation for

  7. Correction of measured charged-particle spectra for energy losses in the target - A comparison of three methods

    CERN Document Server

    Soederberg, J; Alm-Carlsson, G; Olsson, N

    2002-01-01

    The experimental facility, MEDLEY, at the The Svedberg Laboratory in Uppsala, has been constructed to measure neutron-induced charged-particle production cross-sections for (n, xp), (n, xd), (n, xt), (n, x sup 3 He) and (n, x alpha) reactions at neutron energies up to 100 MeV. Corrections for the energy loss of the charged particles in the target are needed in these measurements, as well as for loss of particles. Different approaches have been used in the literature to solve this problem. In this work, a stripping method is developed, which is compared with other methods developed by Rezentes et al. and Slypen et al. The results obtained using the three codes are similar and they could all be used for correction of experimental charged-particle spectra. Statistical fluctuations in the measured spectra cause problems independent of the applied technique, but the way to handle it differs in the three codes.

  8. Calculation of Multisphere Neutron Spectrometer Response Functions in Energy Range up to 20 MeV

    CERN Document Server

    Martinkovic, J

    2005-01-01

    Multisphere neutron spectrometer is a basic instrument of neutron measurements in the scattered radiation field at charged-particles accelerators for radiation protection and dosimetry purposes. The precise calculation of the spectrometer response functions is a necessary condition of the propriety of neutron spectra unfolding. The results of the response functions calculation for the JINR spectrometer with LiI(Eu) detector (a set of 6 homogeneous and 1 heterogeneous moderators, "bare" detector within cadmium cover and without it) at two geometries of the spectrometer irradiation - in uniform monodirectional and uniform isotropic neutron fields - are given. The calculation was carried out by the code MCNP in the neutron energy range 10$^{-8}$-20 MeV.

  9. 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)

  10. Nondestructive analysis of the natural uranium mass through the measurement of delayed neutrons using the technique of pulsed neutron source

    International Nuclear Information System (INIS)

    Coelho, Paulo Rogerio Pinto

    1979-01-01

    This work presents results of non destructive mass analysis of natural uranium by the pulsed source technique. Fissioning is produced by irradiating the test sample with pulses of 14 MeV neutrons and the uranium mass is calculated on a relative scale from the measured emission of delayed neutrons. Individual measurements were normalised against the integral counts of a scintillation detector measuring the 14 MeV neutron intensity. Delayed neutrons were measured using a specially constructed slab detector operated in anti synchronism with the fast pulsed source. The 14 MeV neutrons were produced via the T(d,n) 4 He reaction using a 400 kV Van de Graaff accelerated operated at 200 kV in the pulsed source mode. Three types of sample were analysed, namely: discs of metallic uranium, pellets of sintered uranium oxide and plates of uranium aluminium alloy sandwiched between aluminium. These plates simulated those of Material Testing Reactor fuel elements. Results of measurements were reproducible to within an overall error in the range 1.6 to 3.9%; the specific error depending on the shape, size and mass of the sample. (author)

  11. Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2001-01-01

    The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

  12. 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

  13. Beta-delayed gamma and neutron emission near the double shell closure at 78Ni

    International Nuclear Information System (INIS)

    Rykaczewski, Krzysztof Piotr; Mazzocchi, C.; Grzywacz, R.; Batchelder, J. C.; Bingham, C.R.; Fong, D.; Hamilton, J.H.; Hwang, J.K.; Karny, M.; Krolas, W.; Liddick, S. N.; Morton, A. C.; Mantica, P. F.; Mueller, W. F.; Steiner, M.; Stolz, A.; Winger, J.A.

    2005-01-01

    An experiment was performed at the National Superconducting Cyclotron Laboratory at Michigan State University to investigate β decay of very neutron-rich cobalt isotopes. Beta-delayed neutron emission from 71-74 Co has been observed for the first time. Preliminary results are reported

  14. Test of the rem-counter WENDI-II from Eberline in different energy-dispersed neutron fields

    International Nuclear Information System (INIS)

    Gutermuth, F.; Radon, T.; Fehrenbacher, G.; Siekmann, R.

    2004-03-01

    The neutron rem-counter WENDI-II from Eberline was tested in high-energy particle accelerator produced neutron fields. A radioactive 241 Am-Be(αn) source was used as a reference. The experimentally determined responses are compared to Monte-Carlo simulations of the response function done by R. H. Olsher et al. (2000). The energy spectra of the accelerator produced neutron fields were determined employing Monte-Carlo simulations, too. According to the simulations done by C. Birattari et al. (1998) and in this work these neutron fields exhibit large contributions to the ambient dose equivalent resulting from neutrons with kinetic energy of more than 20 MeV up to a few 100 MeV. The WENDI-II detector proved to show a response of approximately 3.10 9 pulses per Sievert ambient dose equivalent. Considering the experimental and statistical uncertainties the results are consistent with the assumption that the dose response of the WENDI-II reproduces quite accurately the function for the ambient dose equivalent of the ICRP 74

  15. Mean energy polarized neutron source

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  16. Neutron energy measurement for practical applications

    Science.gov (United States)

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

    2018-03-01

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

  17. A Neutron Sensitive Microchannel Plate Detector with Cross Delay Line Readout

    International Nuclear Information System (INIS)

    Berry, Kevin D.; Bilheux, Hassina Z.; Crow, Lowell; Diawara, Yacouba; Feller, W. Bruce; Iverson, Erik B.; Martin, Adrian; Robertson, J. Lee

    2012-01-01

    Microchannel plates containing neutron absorbing elements such as boron and gadolinium in the bulk glass are used as the sensing element in high spatial resolution, high rate neutron imaging systems. In this paper we describe one such device, using both 10 B and natural Gd, which employs cross delay line signal readout, with time-of-flight capability. This detector has a measured spatial resolution under 40 m FWHM, thermal neutron efficiency of 19%, and has recorded rates in excess of 500 kHz. A physical and functional description is presented, followed by a discussion of measurements of detector performance and a brief survey of some practical applications.

  18. Characterisation of an accelerator-based neutron source for BNCT versus beam energy

    Science.gov (United States)

    Agosteo, S.; Curzio, G.; d'Errico, F.; Nath, R.; Tinti, R.

    2002-01-01

    Neutron capture in 10B produces energetic alpha particles that have a high linear energy transfer in tissue. This results in higher cell killing and a higher relative biological effectiveness compared to photons. Using suitably designed boron compounds which preferentially localize in cancerous cells instead of healthy tissues, boron neutron capture therapy (BNCT) has the potential of providing a higher tumor cure rate within minimal toxicity to normal tissues. This clinical approach requires a thermal neutron source, generally a nuclear reactor, with a fluence rate sufficient to deliver tumorcidal doses within a reasonable treatment time (minutes). Thermal neutrons do not penetrate deeply in tissue, therefore BNCT is limited to lesions which are either superficial or otherwise accessible. In this work, we investigate the feasibility of an accelerator-based thermal neutron source for the BNCT of skin melanomas. The source was designed via MCNP Monte Carlo simulations of the thermalization of a fast neutron beam, generated by 7 MeV deuterons impinging on a thick target of beryllium. The neutron field was characterized at several deuteron energies (3.0-6.5 MeV) in an experimental structure installed at the Van De Graaff accelerator of the Laboratori Nazionali di Legnaro, in Italy. Thermal and epithermal neutron fluences were measured with activation techniques and fast neutron spectra were determined with superheated drop detectors (SDD). These neutron spectrometry and dosimetry studies indicated that the fast neutron dose is unacceptably high in the current design. Modifications to the current design to overcome this problem are presented.

  19. Importance of the neutron spectrum for determination of radiation damage

    International Nuclear Information System (INIS)

    Hehn, G.; Stiller, P.; Mattes, M.

    1977-01-01

    Since the radiation effects of neutrons depend strongly on the neutron energy, the correlation between the induced damage and the fluence of the fast neutrons shows appreciable disadvantages. The measured values of changes in material properties resulted in large differences for the same fast neutron fluence, being partly due to different neutron spectra. The uncertainties in damage data led to strong overdesign of important structural components. Different neutron environment at surveillance sample position may give an underestimation of the embrittlement in the reactor pressure vessel, which has to be avoided. The application of damage functions combined with accurately calculated neutron spectra, promise to be a reasonable solution. The damage function has the advantage of a phenomenological quantity that all spectral effects are included. But the correlation quantity has to be determined of high experimental costs. Therefore approximations of its energy distributions are very important. For the keV energy region the kerma function is reasonably good. For the MeV energy region a higher effort is needed to calculate the displacement cross section. The same holds for the low energy part. In all three parts the formation of stable material property levels may vary, so that the final correlation can be determined only by measurements of material properties in different neutron spectra. In material samples the spectra distribution of the displacement production rate was determined at different local positions outside the reactor core of a PWR and a fast breeder showing the most important energy regions of both reactors. (orig.) [de

  20. 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)

  1. The covariance matrix of neutron spectra used in the REAL 84 exercise

    International Nuclear Information System (INIS)

    Matzke, M.

    1986-08-01

    Covariance matrices of continuous functions are discussed. It is pointed out that the number of non-vanishing eigenvalues corresponds to the number of random variables (parameters) involved in the construction of the continuous functions. The covariance matrices used in the REAL 84 international intercomparison of unfolding methods of neutron spectra are investigated. It is shown that a small rank of these covariance matrices leads to a restriction of the possible solution spectra. (orig.) [de

  2. Neutron energy distribution function reconstructed from time-of-flight signals in deuterium gas-puff Z-pinch

    Czech Academy of Sciences Publication Activity Database

    Klír, D.; Kravárik, J.; Kubeš, J.; Rezac, K.; Ananev, S.S.; Bakshaev, Y. L.; Blinov, P. I.; Chernenko, A. S.; Kazakov, E.D.; Korolev, V. D.; Ustroev, G. I.; Juha, Libor; Krása, Josef; Velyhan, Andriy

    2009-01-01

    Roč. 37, č. 3 (2009), s. 425-432 ISSN 0093-3813 R&D Projects: GA MŠk(CZ) LC528; GA MŠk LA08024 Grant - others:IAEA(XE) RC 14817 Institutional research plan: CEZ:AV0Z10100523 Keywords : deuterium * fusion reaction * gas puff * Monte Carlo reconstruction * neutron energy spectra * neutron s * Z-pinch Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.043, year: 2009

  3. High-capacity neutron activation analysis facility

    International Nuclear Information System (INIS)

    Hochel, R.C.; Bowman, W.W.; Zeh, C.W.

    1980-01-01

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program, which is sponsored and funded by the United States Department of Energy, Grand Junction Office. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  4. Beta-Delayed Neutron Spectroscopy of 72Co with VANDLE

    Science.gov (United States)

    Keeler, Andrew; Grzywacz, Robert; King, Thomas; Taylor, Steven; Paulauskas, Stanley; Zachary, Christopher; Vandle Collaboration

    2017-09-01

    Measurements of simple, closed-shell isotopes far from stability provide important benchmarks for nuclear models and are a key constraint in r-process calculations. In particular, r-process models are sensitive to beta decay lifetimes and branching ratios of these neutron-rich isotopes. In this experiment, the Versatile Array of Neutron Detectors at Low Energy (VANDLE) was used to observe decays of nuclei produced by the fragmentation of 82Se at the National Superconducting Cyclotron Laboratory (NSCL). The neutron and gamma emissions of 72Co were measured to map the beta strength distribution (S_beta) above the neutron separation energy and infer the size of the Z = 28 shell gap in the 78Ni region. An implantation detector made of a radiation-hardened, inorganic scintillator was used to correlate implanted ions with beta decays as well as provide a start signal for the neutron Time of Flight measurement. Funded by the National Nuclear Security Administration under the Stewardship Science Academic Alliances program through DOE Award No. DE-NA0002132 and by the Office of Nuclear Physics, U.S. Department of Energy under Awards No. DE-FG02-96ER40983 (UTK).

  5. Future neutron data activity on the neutron source IREN

    International Nuclear Information System (INIS)

    Janeva, N.B.; Koyumdjieva, N.T.; Grigoriev, Y.V.; Gundorin, N.A.; Mareev, Y.D.; Kopatch, Y.N.; Pikelner, L.B.; Shvetsov, V.N.; Sedyshev, P.V.; Zeinalov, S.; Ruskov, I.N.

    2011-01-01

    The global energy demand continues to rise and nuclear power has a potential to be part of the solution of energy problem. Complete and accurate information about the nuclear reactions ensures developing and operating nuclear reactors to reach high efficiencies and adequate safety standards. This demands many nuclear data of improved quality, including covariance nuclear data and correlations. The new neutron source IREN (1 stage) has been put in operation at the end of 2009. The first stage includes the construction of the LUE-200 linear accelerator and non multiplying target. The first measured TOF spectra have been presented recently. The facility is in continuous completion and improvement (according to the full version in the project). The program for neutron data investigation on the IREN neutron source is in preparation. The measuring targets for neutron cross-sections TOF spectra would be selected between isotopes of construction materials, fission products and minor actinides. Now the experimental facilities are in preparation - detectors, innovative electronics equipment and systems for data acquisition and analysis. (authors)

  6. Influence of neutron scattering and source extent on the measurement of neutron energy spectra at ASDEX

    International Nuclear Information System (INIS)

    Huebner, K.; Baetzner, R.; Roos, M.; Robouch, B.V.; Ingrosso, L.; Wurz, H.

    1987-08-01

    The problem of nuclear emulsion measurements at ASDEX is considered. Besides the application of the VINIA-3DAMC software, this needs a description of the plasma neutron source, a model of the ASDEX structure, and calculation of the response of the nuclear emulsion to the incoming spectral neutron fluence. The latter is essential for comparing the numerical results with measurements at ASDEX. To treat this part, the NEPMC software was developed. The aim of the present work is to demonstrate the feasibility, reliability and usefulness of the method. Therefore simplified treatments for the ASDEX model, the plasma neutron source and the track statistics in the NEPMC software were used. Such calculations are of interest not only for nuclear emulsion measurements as well as any other neutron diagnostics, but also for all problems of neutron shielding for other diagnostics. (orig./GG)

  7. Measurements of prompt fission neutron spectra and double-differential neutron inelastic-scattering cross sections for 238U and 232Th

    International Nuclear Information System (INIS)

    Baba, Mamoru; Itoh, Nobuo; Maeda, Kazuto; Hirakawa, Naohiro; Wakabayashi, Hidetaka.

    1989-10-01

    This report presents the summary of experimental studies of prompt fission neutron spectra and double-differential neutron inelastic-scattering cross sections of 238 U and 232 Th. The experiments were performed at Tohoku University Fast Neutron Laboratory employing a time-of-flight technique and Dynamitron accelerator as the pulsed neutron generator. From the experiments, we obtained the following data for both nuclei; 1. prompt fission neutron spectrum for 2 MeV neutrons, 2. double-differential neutron inelastic-scattering cross sections for 1.2, 2.0, 4.2, 6.1 and 14.1 MeV incident neutrons. Both in experiments and data processing, cares were taken to obtain reliable data by avoiding systematic uncertainty. The experimental data were compared with those by other experiments, evaluations and model calculations. Through the data comparison, some fundamental problems were found in the experiments by previous authors and the evaluations. The present data will provide useful data base for refinement of the evaluated data and theoretical models. (author)

  8. Measurement of fast neutron spectra inside reactors with a Li{sup 6} semiconductor counter spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, V S; Lalovic, B I; Petrovic, B P [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1963-12-15

    The possibility of using the Li{sup 6} semiconductor counter spectrometer for measuring fast neutron spectra inside reactors has been investigated in details and some solutions of the difficulties associated with the high interference of thermal neutrons in well-moderated reactors are suggested and checked experimentally (author)

  9. A method for prediction of prompt fission neutron spectra

    International Nuclear Information System (INIS)

    Grashin, A.F.; Lepeshkin, M.V.

    1988-01-01

    Three-parameter formula for the prompt-fission-neutron integral spectrum is derived from a thermodynamical model. Two parameters, scission-neutron weight p = 11 % and anisotropy factor for accelerated fragments b = 10 %, are determined from experimental data, the same values being assumed for any type of fission. The thermodynamical theory provides the value of the third parameter, temperature τ, thus prognozing neutron spectrum and average energy with an error about 1 %. (author)

  10. Modeling the neutron spin-flip process in a time-of-flight spin-resonance energy filter

    CERN Document Server

    Parizzi, A A; Klose, F

    2002-01-01

    A computer program for modeling the neutron spin-flip process in a novel time-of-flight (TOF) spin-resonance energy filter has been developed. The software allows studying the applicability of the device in various areas of spallation neutron scattering instrumentation, for example as a dynamic TOF monochromator. The program uses a quantum-mechanical approach to calculate the local spin-dependent spectra and is essential for optimizing the magnetic field profiles along the resonator axis. (orig.)

  11. 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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. Artificial neural networks in neutron dosimetry

    International Nuclear Information System (INIS)

    Vega-Carrillo, H. R.; Hernandez-Davila, V. M.; Manzanares-Acuna, E.; Mercado, G. A.; Gallego, E.; Lorente, A.; Perales-Munoz, W. A.; Robles-Rodriguez, J. A.

    2006-01-01

    An artificial neural network (ANN) has been designed to obtain neutron doses using only the count rates of a Bonner spheres spectrometer (BSS). Ambient, personal and effective neutron doses were included. One hundred and eighty-one neutron spectra were utilised to calculate the Bonner count rates and the neutron doses. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra, UTA4 response matrix and fluence-to-dose coefficients were used to calculate the count rates in the BSS and the doses. Count rates were used as input and the respective doses were used as output during neural network training. Training and testing were carried out in the MATLAB R environment. The impact of uncertainties in BSS count rates upon the dose quantities calculated with the ANN was investigated by modifying by ±5% the BSS count rates used in the training set. The use of ANNs in neutron dosimetry is an alternative procedure that overcomes the drawbacks associated with this ill-conditioned problem. (authors)

  14. The new high flux neutron source FRM-2 in Munich

    International Nuclear Information System (INIS)

    Roegler, H.J.; Wierheim, G.

    2002-01-01

    Quite some years ago in 1974 to be exact, the first consideration on a new neutron source started at the technical university of Munich (Germany). 27 years later the new high flux neutron source (FRM-2) was read for hot operation, now delayed by a refused approval for its third partial license by the federal government of Germany despite a wide support from the scientific community. FRM-2 is a tank-type research reactor cooled by water, moderated by heavy water and whose thermal power was limited to 20 MW maximum. The extreme compact core together with the applied inverse flux principle led to a neutron flux design value of 8.10 18 n/m 2 .s at the reflector peak. 10 beam tubes will allow an optimized use of the high neutron flux. A hot neutron source with graphite at about 2200 Celsius degrees and a cold neutron source with liquid D 2 at about 25 K will provide shifted energy spectra. The utilization of FRM-2 is many-fold: neutronography and tomography, medical irradiation, radio-nuclide production, doping of pure silicon, neutron activation analysis. (A.C.)

  15. Altitude variation of cosmic-ray neutrons

    International Nuclear Information System (INIS)

    Nakamura, T.; Uwamino, Y.; Ohkubo, T.; Hara, A.

    1987-01-01

    The altitude variation of the cosmic-ray neutron energy spectrum and the dose equivalent rate was measured at an average geomagnetic latitude of 24 degrees N by using the high-efficiency multi-sphere neutron spectrometer and neutron dose-equivalent counter developed by the authors. The data were obtained from a 2-h flight over Japan on 27 February 1985. The neutron energy spectra measured at sea level and at altitudes of 4880 m and at 11,280 m were compared with the calculated spectra of O'Brien and with other experimental spectra, and they are in moderately good agreement with them. The dose equivalent rate increases according to a quadratic curve up to about 6000 m and then increases linearly between 6000 m and 11,280 m. The dependence of dose equivalent rates at sea level and at an altitude of 12,500 m on geomagnetic latitude also is given by referring to other experimental results

  16. Future possibilities with intermediate-energy neutron beams

    International Nuclear Information System (INIS)

    Brady, F.P.

    1987-01-01

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

  17. Delayed power analysis

    International Nuclear Information System (INIS)

    Adamovich, L.A.; Azarov, V.V.

    1999-01-01

    Time dependent core power behavior in a nuclear reactor is described with well-known neutron kinetics equations. At the same time, two portions are distinguished in energy released from uranium nuclei fission; one released directly at fission and another delayed (residual) portion produced during radioactive decay of fission products. While prompt power is definitely described with kinetics equations, the delayed power presentation still remains outstanding. Since in operation the delayed power part is relatively small (about 6%) operation, it can be neglected for small reactivity disturbances assuming that entire power obeys neutron kinetics equations. In case of a high negative reactivity rapidly inserted in core (e.g. reactor scram initiation) the prompt and delayed components can be calculated separately with practically no impact on each other, employing kinetics equations for prompt power and known approximation formulas for delayed portion, named residual in this specific case. Under substantial disturbances the prompt component in the dynamic process becomes commensurable with delayed portion, thus making necessary to take into account their cross impact. A system of differential equations to describe time-dependent behavior of delayed power is presented. Specific NPP analysis shows a way to significantly simplify the task formulation. (author)

  18. Comparison of integral values for measured and calculated fast neutron spectra in lithium fluoride piles

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    1989-01-01

    The tritium production density, kerma heat production density, dose and certain integral values of scalar neutron spectra in bare and graphite-reflected lithium-fluoride piles irradiated with D-T neutrons were evaluated from the pulse height distribution of a miniature NE213 neutron spectrometer with UFO data processing code, and compared with the values calculated with MORSE-CV Monte Carlo code. (author). 8 refs.; 1 fig.; 2 tabs

  19. 14 MeV calibration of JET neutron detectors—phase 1: calibration and characterization of the neutron source

    Science.gov (United States)

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

    2018-02-01

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

  20. Automated uranium analysis by delayed-neutron counting

    International Nuclear Information System (INIS)

    Kunzendorf, H.; Loevborg, L.; Christiansen, E.M.

    1980-10-01

    Automated uranium analysis by fission-induced delayed-neutron counting is described. A short description is given of the instrumentation including transfer system, process control, irradiation and counting sites, and computer operations. Characteristic parameters of the facility (sample preparations, background, and standards) are discussed. A sensitivity of 817 +- 22 counts per 10 -6 g U is found using irradiation, delay, and counting times of 20 s, 5 s, and 10 s, respectively. Presicion is generally less than 1% for normal geological samples. Critical level and detection limits for 7.5 g samples are 8 and 16 ppb, respectively. The importance of some physical and elemental interferences are outlined. Dead-time corrections of measured count rates are necessary and a polynomical expression is used for count rates up to 10 5 . The presence of rare earth elements is regarded as the most important elemental interference. A typical application is given and other areas of application are described. (auther)