WorldWideScience

Sample records for calculated neutron-induced cross

  1. Calculation of neutron-induced single-event upset cross sections for semiconductor memory devices

    International Nuclear Information System (INIS)

    Ikeuchi, Taketo; Watanabe, Yukinobu; Nakashima, Hideki; Sun, Weili

    2001-01-01

    Neutron-induced single-event upset (SEU) cross sections for semiconductor memory devices are calculated by the Burst Generation Rate (BGR) method using LA150 data and QMD calculation in the neutron energy range between 20 MeV and 10 GeV. The calculated results are compared with the measured SEU cross sections for energies up to 160 MeV, and the validity of the calculation method and the nuclear data used is verified. The kind of reaction products and the neutron energy range that have the most effect on SEU are discussed. (author)

  2. Cross-section calculations for neutron-induced reactions up to 50 MeV

    International Nuclear Information System (INIS)

    Yamamuro, Nobuhiro.

    1996-01-01

    In the field of accelerator development, medium-energy reaction cross-section data for structural materials of accelerator and shielding components are required, especially for radiation protection purposes. For a d + Li stripping reaction neutron source used in materials research, neutron reaction cross sections up to 50 MeV are necessary for the design study of neutron irradiation facilities. The current version of SINCROS-II is able to calculate neutron and proton-induced reaction cross sections up to ∼ 50 MeV with some modifications and extensions of the cross-section calculation code. The production of isotopes when structural materials and other materials are bombarded with neutrons or protons is calculated using a revised code in the SINCROS-II system. The parameters used in the cross-section calculations are mainly examined with proton-induced reactions because the experimental data for neutrons above 20 MeV are rare. The status of medium mass nuclide evaluations for aluminum, silicon, chromium, manganese, and copper is presented. These data are useful to estimate the radiation and transmutation of nuclei in the materials

  3. Talys calculations for evaluation of neutron-induced single-event upset cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Bourselier, Jean-Christophe

    2005-08-15

    The computer code TALYS has been used to calculate interactions between cosmic-ray neutrons and silicon nuclei with the goal to describe single-event upset (SEU) cross sections in microelectronics devices. Calculations for the Si(n,X) reaction extend over an energy range of 2 to 200 MeV. The obtained energy spectra of the resulting residuals and light-ions have been integrated using several different critical charges as SEU threshold. It is found that the SEU cross section seems largely to be dominated by {sup 28}Si recoils from elastic scattering. Furthermore, the shape of the SEU cross section as a function of the energy of the incoming neutron changes drastically with decreasing critical charge. The results presented in this report stress the importance of performing studies at mono-energetic neutron beams to advance the understanding of the underlying mechanisms causing SEUs.

  4. Talys calculations for evaluation of neutron-induced single-event upset cross sections

    International Nuclear Information System (INIS)

    Bourselier, Jean-Christophe

    2005-08-01

    The computer code TALYS has been used to calculate interactions between cosmic-ray neutrons and silicon nuclei with the goal to describe single-event upset (SEU) cross sections in microelectronics devices. Calculations for the Si(n,X) reaction extend over an energy range of 2 to 200 MeV. The obtained energy spectra of the resulting residuals and light-ions have been integrated using several different critical charges as SEU threshold. It is found that the SEU cross section seems largely to be dominated by 28 Si recoils from elastic scattering. Furthermore, the shape of the SEU cross section as a function of the energy of the incoming neutron changes drastically with decreasing critical charge. The results presented in this report stress the importance of performing studies at mono-energetic neutron beams to advance the understanding of the underlying mechanisms causing SEUs

  5. Neutron-induced fission cross sections

    International Nuclear Information System (INIS)

    Weigmann, H.

    1991-01-01

    In the history of fission research, neutron-induced fission has always played the most important role. The practical importance of neutron-induced fission rests upon the fact that additional neutrons are produced in the fission process, and thus a chain reaction becomes possible. The practical applications of neutron-induced fission will not be discussed in this chapter, but only the physical properties of one of its characteristics, namely (n,f) cross sections. The most important early summaries on the subject are the monograph edited by Michaudon which also deals with the practical applications, the earlier review article on fission by Michaudon, and the review by Bjornholm and Lynn, in which neutron-induced fission receives major attention. This chapter will attempt to go an intermediate way between the very detailed theoretical treatment in the latter review and the cited monograph which emphasizes the applied aspects and the techniques of fission cross-section measurements. The more recent investigations in the field will be included. Section II will survey the properties of cross sections for neutron-induced fission and also address some special aspects of the experimental methods applied in their measurement. Section Ill will deal with the formal theory of neutron-induced nuclear reactions for the resolved resonance region and the region of statistical nuclear reactions. In Section IV, the fission width, or fission transmission coefficient, will be discussed in detail. Section V will deal with the broader structures due to incompletely damped vibrational resonances, and in particular will address the special case of thorium and neighboring isotopes. Finally, Section VI will briefly discuss parity violation effects in neutron-induced fission. 74 refs., 14 figs., 3 tabs

  6. Calculated neutron-induced cross sections for /sup 58,60/Ni from 1 to 20 MeV and comparisons with experiments

    International Nuclear Information System (INIS)

    Hetrick, D.M.; Fu, C.Y.; Larson, D.C.

    1987-06-01

    Nuclear model codes were used to compute cross sections for neutron-induced reactions on both 58 Ni and 60 Ni for incident energies from 1 to 20 MeV. The input parameters for the model codes were determined through analysis of experimental data in this energy region. Discussion of the models used, the input data, the resulting calculations, extensive comparisons to measured data, and comparisons to the Evaluated Nuclear Data File (ENDF/B-V) for Ni (MAT 1328) are included in this report. 118 refs., 101 figs., 19 tabs

  7. Calculated neutron-induced cross sections for /sup 63/ /sup 65/Cu from 1 to 20 MeV and comparisons with experiments

    Energy Technology Data Exchange (ETDEWEB)

    Hetrick, D.M.; Fu, C.Y.; Larson, D.C.

    1984-08-01

    Nuclear model codes were used to compute cross sections for neutron-induced reactions on both /sup 63/Cu and /sup 65/Cu for incident energies from 1 to 20 MeV. The input parameters for the model codes were determined through analysis of experimental data in this energy region. Discussion of the models used, the input data, the resulting calculations, extensive comparisons to measured data, and comparisons to the Evaluated Nuclear Data File (ENDF/B-V) for Cu (MAT 1329) are included in this report.

  8. The needs for program and cross-section library improvement in calculation of neutron-induced activity inventories

    International Nuclear Information System (INIS)

    Yavshitz, S.G.; Rubchenya, V.A.; Rimski-Korsakov, A.A.

    1993-01-01

    The authors demonstrate the possibility of an approach to evaluate the radioactive inventory - induced activity of structural materials and surface contamination of reactor components, that will fit well into ORIGEN code structure and could be used on a modest PC directly on the decommissioning site. This approach would also require only one well tested set of pre-calculated and adjusted by experiment cross-section libraries (averaged by typical neutron spectra outside the reactor core). 15 refs, 1 fig

  9. Neutron-induced cross-sections via the surrogate method

    International Nuclear Information System (INIS)

    Boutoux, G.

    2011-11-01

    The surrogate reaction method is an indirect way of determining neutron-induced cross sections through transfer or inelastic scattering reactions. This method presents the advantage that in some cases the target material is stable or less radioactive than the material required for a neutron-induced measurement. The method is based on the hypothesis that the excited nucleus is a compound nucleus whose decay depends essentially on its excitation energy and on the spin and parity state of the populated compound state. Nevertheless, the spin and parity population differences between the compound-nuclei produced in the neutron and transfer-induced reactions may be different. This work reviews the surrogate method and its validity. Neutron-induced fission cross sections obtained with the surrogate method are in general good agreement. However, it is not yet clear to what extent the surrogate method can be applied to infer radiative capture cross sections. We performed an experiment to determine the gamma decay probabilities for 176 Lu and 173 Yb by using the surrogate reactions 174 Yb( 3 He,pγ) 176 Lu * and 174 Yb( 3 He,αγ) 173 Yb * , respectively, and compare them with the well-known corresponding probabilities obtained in the 175 Lu(n,γ) and 172 Yb(n,γ) reactions. This experiment provides answers to understand why, in the case of gamma-decay, the surrogate method gives significant deviations compared to the corresponding neutron-induced reaction. In this work, we have also assessed whether the surrogate method can be applied to extract capture probabilities in the actinide region. Previous experiments on fission have also been reinterpreted. Thus, this work provides new insights into the surrogate method. This work is organised in the following way: in chapter 1, the theoretical aspects related to the surrogate method will be introduced. The validity of the surrogate method will be investigated by means of statistical model calculations. In chapter 2, a review on

  10. Nuclear fission and neutron-induced fission cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    James, G.D.; Lynn, J.E.; Michaudon, A.; Rowlands, J.; de Saussure, G.

    1981-01-01

    A general presentation of current knowledge of the fission process is given with emphasis on the low energy fission of actinide nuclei and neutron induced fission. The need for and the required accuracy of fission cross section data in nuclear energy programs are discussed. A summary is given of the steps involved in fission cross section measurement and the range of available techniques. Methods of fission detection are described with emphasis on energy dependent changed and detector efficiency. Examples of cross section measurements are given and data reduction is discussed. The calculation of fission cross sections is discussed and relevant nuclear theory including the formation and decay of compound nuclei and energy level density is introduced. A description of a practical computation of fission cross sections is given.

  11. 238U subthreshold neutron induced fission cross section

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  12. Actinide neutron-induced fission cross section measurements at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

    Tovesson, Fredrik K [Los Alamos National Laboratory; Laptev, Alexander B [Los Alamos National Laboratory; Hill, Tony S [INL

    2010-01-01

    Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications in a wide energy range from sub-thermal energies up to 200 MeV. A parallel-plate ionization chamber are used to measure fission cross sections ratios relative to the {sup 235}U standard while incident neutron energies are determined using the time-of-flight method. Recent measurements include the {sup 233,238}U, {sup 239-242}Pu and {sup 243}Am neutron-induced fission cross sections. Obtained data are presented in comparison with ex isting evaluations and previous data.

  13. Model calculations of excitation functions of neutron-induced reactions on Rh

    International Nuclear Information System (INIS)

    Strohmaier, Brigitte

    1995-01-01

    Cross sections of neutron-induced reactions on 103 Rh have been calculated by means of the statistical model and the coupled-channels optical model for incident-neutron energies up to 30 MeV. The incentive for this study was a new measurement of the 103 Rh(n, n') 103m Rh cross section which will - together with the present calculations -enter into a dosimetry-reaction evaluation. The validation of the model parameters relied on nuclear-structure data as far as possible. (author)

  14. Nuclear fission and neutron-induced fission cross-sections

    CERN Document Server

    James, G D; Michaudon, A; Michaudon, A; Cierjacks, S W; Chrien, R E

    2013-01-01

    Nuclear Fission and Neutron-Induced Fission Cross-Sections is the first volume in a series on Neutron Physics and Nuclear Data in Science and Technology. This volume serves the purpose of providing a thorough description of the many facets of neutron physics in different fields of nuclear applications. This book also attempts to bridge the communication gap between experts involved in the experimental and theoretical studies of nuclear properties and those involved in the technological applications of nuclear data. This publication will be invaluable to those interested in studying nuclear fis

  15. New calculation for the neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV

    International Nuclear Information System (INIS)

    Mesa, J.; Deppman, A.; Likhachev, V.P.; Arruda-Neto, J.D.T.; Manso, M.V.; Garcia, C.E.; Rodriguez, O.; Guzman, F.; Garcia, F.

    2003-01-01

    The 233 Pa(n,f) cross section, a key ingredient for fast reactors and accelerators driven systems, was measured recently with relatively good accuracy [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. The results are at strong variance with accepted evaluations and an existing indirect experiment. This circumstance led us to perform a quite detailed and complete evaluation of the 233 Pa(n,f) cross section between 1.0 and 3.0 MeV, where use of our newly developed routines for the parametrization of the nuclear surface and the calculation of deformation parameters and level densities (including low-energy discrete levels) were made. The results show good quantitative and excellent qualitative agreement with the experimental direct data obtained by Tovesson et al. [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. Additionally, our methodology opens new possibilities for the analysis of subthreshold fission and above threshold second-chance fission for both 233 Pa and its decay product 233 U, as well as other strategically important fissionable nuclides

  16. Reaction Cross Section Calculations in Neutron Induced Reactions and GEANT4 Simulation of Hadronic Interactions for the Reactor Moderator Material BeO

    Directory of Open Access Journals (Sweden)

    Veli ÇAPALI

    2016-05-01

    Full Text Available BeO is one of the most common moderator material for neutron moderation; due to its high density, neutron capture cross section and physical-chemical properties that provides usage at elevated temperatures. As it’s known, for various applications in the field of reactor design and neutron capture, reaction cross–section data are required. The cross–sections of (n,α, (n,2n, (n,t, (n,EL and (n,TOT reactions for 9Be and 16O nuclei have been calculated by using TALYS 1.6 Two Component Exciton model and EMPIRE 3.2 Exciton model in this study. Hadronic interactions of low energetic neutrons and generated isotopes–particles have been investigated for a situation in which BeO was used as a neutron moderator by using GEANT4, which is a powerful simulation software. In addition, energy deposition along BeO material has been obtained. Results from performed calculations were compared with the experimental nuclear reaction data exist in EXFOR.

  17. 7Li neutron-induced elastic scattering cross section measurement using a slowing-down spectrometer

    Directory of Open Access Journals (Sweden)

    Heusch M.

    2010-10-01

    Full Text Available A new integral measurement of the 7Li neutron induced elastic scattering cross section was determined in a wide neutron energy range. The measurement was performed on the LPSC-PEREN experimental facility using a heterogeneous graphite-LiF slowing-down time spectrometer coupled with an intense pulsed neutron generator (GENEPI-2. This method allows the measurement of the integral elastic scattering cross section in a slowing-down neutron spectrum. A Bayesian approach coupled to Monte Carlo calculations was applied to extract naturalC, 19F and 7Li elastic scattering cross sections.

  18. Determination of the neutron-induced fission cross section of 242Pu

    International Nuclear Information System (INIS)

    Koegler, Toni Joerg

    2016-01-01

    Neutron induced fission cross sections of actinides like the Pu-isotopes are of relevance for the development of nuclear transmutation technologies. For 242 Pu, current uncertainties are of around 21%. Sensitivity studies show that the total uncertainty has to be reduced to below 5% to allow for reliable neutron physics simulations. This challenging task was performed at the neutron time-of-flight facility of the new German National Center for High Power Radiation Sources at HZDR, Dresden. Within the TRAKULA project, thin, large and homogeneous deposits of 235 U and 242 Pu have been produced successfully. Using two consecutively placed fission chambers allowed the determination of the neutron induced fission cross section of 242 Pu relative to 235 U. The areal density of the Plutonium targets was calculated using the measured spontaneous fission rate. Experimental results of the fast neutron induced fission of 242 Pu acquired at nELBE will be presented and compared to recent experiments and evaluated data. Corrections addressing the neutron scattering are discussed by using results of different neutron transport simulations (Geant 4, MCNP 6 and FLUKA).

  19. Evaluated neutron-induced cross sections for 40Ca from 20 to 40 MeV

    International Nuclear Information System (INIS)

    Hetrick, D.M.; Fu, C.Y.; Larson, D.C.

    1982-09-01

    Nuclear model codes were used to compute cross sections for neutron-induced reactions on 40 Ca for incident energies from 20 to 40 MeV. The input parameters for the model codes were determined through analysis of experimental data in this energy region. Computed cross sections along with emission spectra for each product were combined into an Evaluated Nuclear Data File (ENDF) using the proposed format for charged-particle reactions. Discussion of the models used, the resulting calculations, and the final evaluated data file are presented

  20. Use of shell model calculations in R-matrix studies of neutron-induced reactions

    International Nuclear Information System (INIS)

    Knox, H.D.

    1986-01-01

    R-matrix analyses of neutron-induced reactions for many of the lightest p-shell nuclei are difficult due to a lack of distinct resonance structure in the reaction cross sections. Initial values for the required R-matrix parameters, E,sub(lambda) and γsub(lambdac) for states in the compound system, can be obtained from shell model calculations. In the present work, the results of recent shell model calculations for the lithium isotopes have been used in R-matrix analyses of 6 Li+n and 7 Li+n reactions for E sub(n) 7 Li and 8 Li on the 6 Li+n and 7 Li+n reaction mechanisms and cross sections are discussed. (author)

  1. Evaluation of 28,29,30Si neutron induced cross sections for ENDF/B-VI

    International Nuclear Information System (INIS)

    Hetrick, D.M.; Larson, D.C.; Larson, N.M.; Leal, L.C.; Epperson, S.J.

    1997-04-01

    Separate evaluations have been done for the three stable isotopes of silicon for ENDF/B-VI. The evaluations are based on analysis of experimental data, supplemented by results of nuclear model calculations. The computational methods and the parameters required as input to the nuclear model codes are reviewed. Discussion of the evaluated data given for resonance parameters, neutron induced reaction cross sections, associated angular and energy distributions, and gamma-ray production cross sections is included. Extensive comparisons of the evaluated cross sections to measured data are shown in this report. The evaluations include all necessary data to allow KERMA (Kinetic Energy Released in MAterials) and displacement cross sections to be calculated directly. These quantities are fundamental to studies of neutron heating and radiation damage

  2. Neutron-induced capture cross sections via the surrogate reaction method

    International Nuclear Information System (INIS)

    Boutoux, G.; Jurado, B.; Aiche, M.; Barreau, G.; Capellan, N.; Companis, I.; Czajkowski, S.; Dassie, D.; Haas, B.; Mathieu, L.; Meot, V.; Bail, A.; Bauge, E.; Daugas, J. M.; Faul, T.; Gaudefroy, L.; Morel, P.; Pillet, N.; Roig, O.; Romain, P.; Taieb, J.; Theroine, C.; Burke, J.T.; Companis, I.; Derkx, X.; Gunsing, F.; Matea, I.; Tassan-Got, L.; Porquet, M.G.; Serot, O.

    2011-01-01

    The surrogate reaction method is an indirect way of determining cross sections for nuclear reactions that proceed through a compound nucleus. This technique enables neutron-induced cross sections to be extracted for nuclear reactions on short-lived unstable nuclei that otherwise can not be measured. This technique has been successfully applied to determine the neutron-induced fission cross sections of several short-lived nuclei. In this work, we investigate whether this powerful technique can also be used to determine of neutron-induced capture cross sections. For this purpose we use the surrogate reaction 174 Yb( 3 He, pγ) 176 Lu to infer the well known 175 Lu(n, γ) cross section and compare the results with the directly measured neutron-induced data. This surrogate experiment has been performed in March 2010. The experimental technique used and the first preliminary results will be presented. (authors)

  3. Neutron-induced cross sections of short-lived nuclei via the surrogate reaction method

    Directory of Open Access Journals (Sweden)

    Morel P.

    2011-10-01

    Full Text Available The measurement of neutron-induced cross sections of short-lived nuclei is extremely difficult due to the radioactivity of the samples. The surrogate reaction method is an indirect way of determining cross sections for nuclear reactions that proceed through a compound nucleus. This method presents the advantage that the target material can be stable or less radioactive than the material required for a neutron-induced measurement. We have successfully used the surrogate reaction method to extract neutron-induced fission cross sections of various short-lived actinides. In this work, we investigate whether this technique can be used to determine neutron-induced capture cross sections in the rare-earth region.

  4. Neutron-induced cross sections of short-lived nuclei via the surrogate reaction method

    Directory of Open Access Journals (Sweden)

    Tassan-Got L.

    2012-02-01

    Full Text Available The measurement of neutron-induced cross sections of short-lived nuclei is extremely difficult due to the radioactivity of the samples. The surrogate reaction method is an indirect way of determining cross sections for nuclear reactions that proceed through a compound nucleus. This method presents the advantage that the target material can be stable or less radioactive than the material required for a neutron-induced measurement. We have successfully used the surrogate reaction method to extract neutron-induced fission cross sections of various short-lived actinides. In this work, we investigate whether this technique can be used to determine neutron-induced capture cross sections in the rare-earth region.

  5. High-energy Neutron-induced Fission Cross Sections of Natural Lead and Bismuth-209

    CERN Document Server

    Tarrio, D; Carrapico, C; Eleftheriadis, C; Leeb, H; Calvino, F; Herrera-Martinez, A; Savvidis, I; Vlachoudis, V; Haas, B; Koehler, P; Vannini, G; Oshima, M; Le Naour, C; Gramegna, F; Wiescher, M; Pigni, M T; Audouin, L; Mengoni, A; Quesada, J; Becvar, F; Plag, R; Cennini, P; Mosconi, M; Rauscher, T; Couture, A; Capote, R; Sarchiapone, L; Vlastou, R; Domingo-Pardo, C; Dillmann, I; Pavlopoulos, P; Karamanis, D; Krticka, M; Jericha, E; Ferrari, A; Martinez, T; Trubert, D; Oberhummer, H; Karadimos, D; Plompen, A; Isaev, S; Terlizzi, R; Cortes, G; Cox, J; Cano-Ott, D; Pretel, C; Colonna, N; Berthoumieux, E; Vaz, P; Heil, M; Lopes, I; Lampoudis, C; Walter, S; Calviani, M; Gonzalez-Romero, E; Embid-Segura, M; Stephan, C; Igashira, M; Papachristodoulou, C; Aerts, G; Tavora, L; Berthier, B; Rudolf, G; Andrzejewski, J; Villamarin, D; Ferreira-Marques, R; Tain, J L; O'Brien, S; Reifarth, R; Kadi, Y; Neves, F; Poch, A; Kerveno, M; Rubbia, C; Lazano, M; Dahlfors, M; Wisshak, K; Salgado, J; Dridi, W; Ventura, A; Andriamonje, S; Assimakopoulos, P; Santos, C; Voss, F; Ferrant, L; Patronis, N; Chiaveri, E; Guerrero, C; Perrot, L; Vicente, M C; Lindote, A; Praena, J; Baumann, P; Kappeler, F; Rullhusen, P; Furman, W; David, S; Marrone, S; Tassan-Got, L; Gunsig, F; Alvarez-Velarde, F; Massimi, C; Mastinu, P; Pancin, J; Papadopoulos, C; Tagliente, G; Haight, R; Chepel, V; Kossionides, E; Badurek, G; Marganiec, J; Lukic, S; Pavlik, A; Goncalves, I; Duran, I; Alvarez, H; Abbondanno, U; Fujii, K; Milazzo, P M; Moreau, C

    2011-01-01

    The CERN Neutron Time-Of-Flight (n\\_TOF) facility is well suited to measure small neutron-induced fission cross sections, as those of subactinides. The cross section ratios of (nat)Pb and (209)Bi relative to (235)U and (238)U were measured using PPAC detectors. The fragment coincidence method allows to unambiguously identify the fission events. The present experiment provides the first results for neutron-induced fission up to 1 GeV for (nat)Pb and (209)Bi. A good agreement with previous experimental data below 200 MeV is shown. The comparison with proton-induced fission indicates that the limiting regime where neutron-induced and proton-induced fission reach equal cross section is close to 1 GeV.

  6. Methods and procedures for evaluation of neutron-induced activation cross sections

    International Nuclear Information System (INIS)

    Gardner, M.A.

    1981-09-01

    One cannot expect measurements alone to supply all of the neutron-induced activation cross-section data required by the fission reactor, fusion reactor, and nuclear weapons development communities, given the wide ranges of incident neutron energies, the great variety of possible reaction types leading to activation, and targets both stable and unstable. Therefore, the evaluator must look to nuclear model calculations and systematics to aid in fulfilling these cross-section data needs. This review presents some of the recent developments and improvements in the prediction of neutron activation cross sections, with specific emphasis on the use of empirical and semiempirical methods. Since such systematics require much less nuclear informaion as input and much less computational time than do the multistep Hauser-Feshbach codes, they can often provide certain cross-section data at a sufficient level of accuracy within a minimum amount of time. The cross-section information that these systematics can and cannot provide and those cases in which they can be used most reliably are discussed

  7. Measurement of fast neutron induced fission cross section of minor-actinide

    International Nuclear Information System (INIS)

    Hirakawa, Naohiro

    1997-03-01

    In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am, Cm). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA is measured using Dynamitron Accelerator in Tohoku University. The experimental method and the samples, which were developed or introduced during the last year, were improved in this fiscal year: (1) Development of a sealed fission chamber, (2) Intensification of Li neutron target, (3) Improvement of time-resolution of Time-of-Flight (TOF) electronic circuit, (4) Introduction of Np237 samples with large sample mass and (5) Introduction of a U235 sample with high purity. Using these improved tools and samples, the fission cross section ratio of Np237 relative to U235 was measured between 5 to 100 keV, and the fission cross section of Np237 was deduced. On the other hand, samples of Am241 and Am243 were obtained from Japan Atomic Energy Research Institute (JAERI) after investigating fission cross section of two americium isotopes (Am241 and Am 243) which are important for core physics calculation of fast reactors. (author)

  8. Methods and procedures for evaluation of neutron-induced activation cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, M.A.

    1981-09-01

    One cannot expect measurements alone to supply all of the neutron-induced activation cross-section data required by the fission reactor, fusion reactor, and nuclear weapons development communities, given the wide ranges of incident neutron energies, the great variety of possible reaction types leading to activation, and targets both stable and unstable. Therefore, the evaluator must look to nuclear model calculations and systematics to aid in fulfilling these cross-section data needs. This review presents some of the recent developments and improvements in the prediction of neutron activation cross sections, with specific emphasis on the use of empirical and semiempirical methods. Since such systematics require much less nuclear informaion as input and much less computational time than do the multistep Hauser-Feshbach codes, they can often provide certain cross-section data at a sufficient level of accuracy within a minimum amount of time. The cross-section information that these systematics can and cannot provide and those cases in which they can be used most reliably are discussed.

  9. Integrated system for production of neutronics and photonics calculational constants. Neutron-induced interactions: index of experimental data

    International Nuclear Information System (INIS)

    MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.

    1976-01-01

    Indexes to the neutron-induced interaction data in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has two arrangements: isotope (ZA) order and reaction-number order

  10. Integrated system for production of neutronics and photonics calculational constants. Neutron-induced interactions: index of experimental data

    Energy Technology Data Exchange (ETDEWEB)

    MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.

    1976-07-04

    Indexes to the neutron-induced interaction data in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has two arrangements: isotope (ZA) order and reaction-number order.

  11. Measurement of activation cross sections for quasi-monoenergetic neutron induced reactions of {sup 89}Y

    Energy Technology Data Exchange (ETDEWEB)

    Zaman, Muhammad; Kim, Guinyun; Kim, Kwangsoo; Nadeem, Muhammad [Kyungpook National University, Department of Physics and Center for High Energy Physics, Daegu (Korea, Republic of); Naik, Haladhara [Kyungpook National University, Department of Physics and Center for High Energy Physics, Daegu (Korea, Republic of); Bhabha Atomic Research Centre, Radiochemistry Division, Mumbai (India); Lee, Manwoo [Dongnam Inst. of Radiological and Medical Sciences, Research Center, Busan (Korea, Republic of)

    2017-09-15

    The neutron induced cross sections of the {sup 89}Y(n, 2n){sup 88}Y, {sup 89}Y(n, 3n){sup 87}Y and {sup 89}Y(n, 4n){sup 86}Y reactions were measured in the neutron energy range of 15.2 to 37.2 MeV by using an activation and off-line γ-ray spectrometric technique. The quasi-monoenergetic neutrons used for the above reactions are based on a {sup 9}Be(p, n) reaction. Simulations of the neutron spectra from the Be target were done using the MCNPX 2.6.0 program. Theoretical calculations were performed for the {sup 89}Y(n, 2n){sup 88}Y, {sup 89}Y(n, 3n){sup 87}Y and {sup 89}Y(n, 4n){sup 86}Y reaction cross sections using nuclear model code Talys 1.8. The measured and calculated cross sections were compared with the literature data given in EXFOR and the TENDL-2015 data libraries. The present data of the {sup 89}Y(n, xn) reaction were also compared with the similar data of the {sup 89}Y(γ, xn) reaction to examine the effect of the entrance channel parameters as well as the role of projectiles and ejectiles. (orig.)

  12. Determination of the fast-neutron-induced fission cross-section of 242Pu at nELBE

    Science.gov (United States)

    Kögler, Toni; Beyer, Roland; Junghans, Arnd R.; Schwengner, Ronald; Wagner, Andreas

    2018-03-01

    The fast-neutron-induced fission cross section of 242Pu was determined in the energy range of 0.5 MeV to 10MeV at the neutron time-of-flight facility nELBE. Using a parallel-plate fission ionization chamber this quantity was measured relative to 235U(n,f). The number of target nuclei was thereby calculated by means of measuring the spontaneous fission rate of 242Pu. An MCNP 6 neutron transport simulation was used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

  13. Determination of the fast-neutron-induced fission cross-section of 242Pu at nELBE

    Directory of Open Access Journals (Sweden)

    Kögler Toni

    2018-01-01

    Full Text Available The fast-neutron-induced fission cross section of 242Pu was determined in the energy range of 0.5 MeV to 10MeV at the neutron time-of-flight facility nELBE. Using a parallel-plate fission ionization chamber this quantity was measured relative to 235U(n,f. The number of target nuclei was thereby calculated by means of measuring the spontaneous fission rate of 242Pu. An MCNP 6 neutron transport simulation was used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

  14. Measurement of the neutron-induced fission cross-section of 240,242Pu

    International Nuclear Information System (INIS)

    Salvador-Castineira, P.; Hambsch, F.J.; Brys, T.; Oberstedt, S.; Vidali, M.; Pretel, C.

    2014-01-01

    Fast spectrum neutron-induced fission cross-section data for transuranic isotopes are in high demand in the nuclear data community. In particular, highly accurate data are needed for the new Generation-IV nuclear applications. The aim is to obtain precise neutron-induced fission cross-sections for 240 Pu and 242 Pu. In this context accurate data on spontaneous fission half-lives have also been measured. To minimise the total uncertainty on the fission cross-sections the detector efficiency has been studied in detail. Both isotopes have been measured using a twin Frisch-grid ionisation chamber (TFGIC) due to its superiority compared to other detector systems in view of radiation hardness, 2 x 2π solid angle coverage and very good energy resolution. (authors)

  15. Neutron-induced cross sections of actinides via the surrogate-reaction method

    Directory of Open Access Journals (Sweden)

    Ducasse Q.

    2013-12-01

    Full Text Available The surrogate-reaction method is an indirect way of determining cross sections for reactions that proceed through a compound nucleus. This technique may enable neutron-induced cross sections to be extracted for short-lived nuclei that otherwise cannot be measured. However, the validity of the surrogate method has to be investigated. In particular, the absence of a compound nucleus formation and the Jπ dependence of the decay probabilities may question the method. In this work we study the reactions 238U(d,p239U, 238U(3He,t238Np, 238U(3He,4He237U as surrogates for neutron-induced reactions on 238U, 237Np and 236U, respectively, for which good quality data exist. The experimental set-up enabled the measurement of fission and gamma-decay probabilities. The first results are hereby presented.

  16. Above-threshold structure in {sup 244}Cm neutron-induced fission cross section

    Energy Technology Data Exchange (ETDEWEB)

    Maslov, V.M. [Radiation Physics and Chemistry Problems Inst., Minsk-Sosny (Belarus)

    1997-03-01

    The quasi-resonance structure appearing above the fission threshold in neutron-induced fission cross section of {sup 244}Cm(n,f) is interpreted. It is shown to be due to excitation of few-quasiparticle states in fissioning {sup 245}Cm and residual {sup 244}Cm nuclides. The estimate of quasiparticle excitation thresholds in fissioning nuclide {sup 245}Cm is consistent with pairing gap and fission barrier parameters. (author)

  17. Modeled Neutron Induced Nuclear Reaction Cross Sections for Radiochemsitry in the region of Thulium, Lutetium, and Tantalum I. Results of Built in Spherical Symmetry in a Deformed Region

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, R. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-09-06

    We have developed a set of modeled nuclear reaction cross sections for use in radiochemical diagnostics. Systematics for the input parameters required by the Hauser-Feshbach statistical model were developed and used to calculate neutron induced nuclear reaction cross sections for targets ranging from Terbium (Z = 65) to Rhenium (Z = 75). Of particular interest are the cross sections on Tm, Lu, and Ta including reactions on isomeric targets.

  18. ZZ RRDF-98, Cross-sections and covariance matrices for 22 neutron induced dosimetry reactions

    International Nuclear Information System (INIS)

    Zolotarev, K.I.; Ignatyuk, A.V.; Mahokhin, V.N.; Pashchenko, A.B.

    2005-01-01

    1 - Description of program or function: Format: ENDF-6 format; Number of groups: Continuous energy; Dosimetry reactions: 6-C-12(n,2n), 8-O-16(n,2n), 9-F-19(n,2n), 12-Mg-24(n,p), 22-Ti-46(n,2n), 22-Ti-46(n,p), 22-Ti-47(n,x), 22-Ti-48(n,p), 22-Ti-48(n,x), 22-Ti-49(n,x), 23-V-51(n,alpha), 26-Fe-54(n,2n), 26-Fe-54(n,alpha), 26-Fe-56(n,p), 27-Co-59(n,alpha), 29-Cu-63(n,alpha), 33-As-75(n,2n), 41-Nb-93(n,2n), 41-Nb-93(n,n'), 45-Rh-103(n,n'), 49-In-115(n,n'), 59-Pr-141(n,2n); Origin: Russian Federation; Weighting spectrum: None. RRDF-98 contains original evaluations of cross section data performed at the Institute of Physics and Power Engineering, Obninsk, for 22 neutron induced dosimetry reactions. The dataset also contains the corresponding covariance matrices. 2 - Methods: The evaluation of excitation functions was performed on the basis of statistical analysis of corrected experimental data in the framework of generalized least squares method and taking into account the results of optical-statistical STAPRE and GNASH calculations. The experimental cross section data including the most recent results were critically reviewed and processed in this study. If necessary, the data were normalized in order to make adjustments in relevant cross sections and decay schemes. The covariance matrices were prepared and the evaluated cross section data are presented in ENDF-6 format (Files 3, 33). For estimation of correlations between experimental data the total uncertainties of measured cross sections have been separated into statistical and systematic parts and correlation coefficients between components of systematic parts were assigned according to information given in the original publications and EXFOR library. Then the correlation matrix of cross sections measured within one experiment was calculated and approximated by matrix with a constant (average) correlation coefficient. The overall correlation matrix was composed of such sub-matrices in the assumption that the cross

  19. Linear cascade calculations of matrix due to neutron-induced nuclear reactions

    International Nuclear Information System (INIS)

    Avila, Ricardo E

    2000-01-01

    A method is developed to calculate the total number of displacements created by energetic particles resulting from neutron-induced nuclear reactions. The method is specifically conceived to calculate the damage in lithium ceramics by the 6L i(n, α)T reaction. The damage created by any particle is related to that caused by atoms from the matrix recoiling after collision with the primary particle. An integral equation for that self-damage is solved by interactions, using the magic stopping powers of Ziegler, Biersack and Littmark. A projectile-substrate dependent Kinchin-Pease model is proposed, giving and analytic approximation to the total damage as a function of the initial particle energy (au)

  20. Measurements of neutron-induced fission cross sections of Pb and Bi at intermediate energies

    International Nuclear Information System (INIS)

    Ryzhov, Igor; Tutin, Gennady; Eismont, Vilen; Mitryukhin, Andrey; Oplavin, Valery; Soloviev, Sergey; Conde, Henri; Olsson, Nils; Renberg, Per-Ulf

    2002-01-01

    Neutron-induced fission cross sections of nat Pb and 209 Bi have been measured relative to the 238 U(n.f) cross section at energies 96 MeV for lead and 133 MeV for bismuth. The measurements were performed at the quasi-mono-energetic neutron beam facility of The Svedberg Laboratory in Uppsala using Frisch-gridded ionization chamber. The results obtained are compared with other experimental data. The present state of the Bi standard recommended by IAEA is discussed. (author)

  1. Measurement of the neutron-induced deuteron breakup reaction cross-section between 5 and 25 MeV

    International Nuclear Information System (INIS)

    Laborie, J.M.; Ledoux, X.; Varignon, C.; Lazauskas, R.; Morillon, B.; Belier, G.

    2012-01-01

    This article presents a full program devoted to the calculation and the measurement of the neutron-induced deuteron break-up reaction cross-section between 5 and 10 MeV, and between 20 and 25 MeV. Measurements are compared with theoretical calculations based on the solution of the Faddeev equations for a realistic nuclear Hamiltonian. The experiments were performed at the Tandem 7 MV accelerator at CEA, DAM, DIF, France. The measurements were carried out with a C 6 D 6 detector as active deuterium target located at the center of a 4π neutron counter (see C. Varignon et al., Nucl. Instrum. Methods B 248, 329 (2006)) which allows to count the two neutrons emitted in the 2 H(n, 2n)p reaction. Comparisons of the new data and calculations are made with the existing data as well as the CENDL2, JENDL3.3 and ENDF/B-VII evaluations. (orig.)

  2. Evaluation of cross sections for neutron-induced reactions in sodium

    International Nuclear Information System (INIS)

    Larson, D.C.

    1980-09-01

    An evaluation of the neutron-induced cross sections of 23 Na has been done for the energy range from 10 -5 eV to 20 MeV. All significant cross sections are given, including differential cross sections for production of gamma rays. The recommended values are based on experimental data where available, and use results of a consistent model code analysis of available data to predict cross sections where there are no experimental data. This report describes the evaluation that was submitted to the Cross Section Evaluation Working Group (CSEWG) for consideration as a part of the Evaluated Nuclear Data File, Version V, and subsequently issued as MAT 1311. 126 references, 130 figures, 14 tables

  3. Analysis of reaction cross-section production in neutron induced fission reactions on uranium isotope using computer code COMPLET.

    Science.gov (United States)

    Asres, Yihunie Hibstie; Mathuthu, Manny; Birhane, Marelgn Derso

    2018-04-22

    This study provides current evidence about cross-section production processes in the theoretical and experimental results of neutron induced reaction of uranium isotope on projectile energy range of 1-100 MeV in order to improve the reliability of nuclear stimulation. In such fission reactions of 235 U within nuclear reactors, much amount of energy would be released as a product that able to satisfy the needs of energy to the world wide without polluting processes as compared to other sources. The main objective of this work is to transform a related knowledge in the neutron-induced fission reactions on 235 U through describing, analyzing and interpreting the theoretical results of the cross sections obtained from computer code COMPLET by comparing with the experimental data obtained from EXFOR. The cross section value of 235 U(n,2n) 234 U, 235 U(n,3n) 233 U, 235 U(n,γ) 236 U, 235 U(n,f) are obtained using computer code COMPLET and the corresponding experimental values were browsed by EXFOR, IAEA. The theoretical results are compared with the experimental data taken from EXFOR Data Bank. Computer code COMPLET has been used for the analysis with the same set of input parameters and the graphs were plotted by the help of spreadsheet & Origin-8 software. The quantification of uncertainties stemming from both experimental data and computer code calculation plays a significant role in the final evaluated results. The calculated results for total cross sections were compared with the experimental data taken from EXFOR in the literature, and good agreement was found between the experimental and theoretical data. This comparison of the calculated data was analyzed and interpreted with tabulation and graphical descriptions, and the results were briefly discussed within the text of this research work. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

  4. Neutron-induced cross sections of actinides via the surrogate-reaction method

    Directory of Open Access Journals (Sweden)

    Tveten G. M.

    2013-03-01

    Full Text Available The surrogate-reaction method is an indirect way of determining cross sections for reactions that proceed through a compound nucleus. This technique may enable neutron-induced cross sections to be extracted for short-lived nuclei that otherwise cannot be measured. However, the validity of the surrogate method for extracting capture cross sections has to be investigated. In this work we study the reactions 238U(d,p239U, 238U(3He,t238Np, 238U(3He,4He237U as surrogates for neutroninduced reactions on 238U, 237Np and 236U, respectively, for which good quality data exist. The experimental set-up enabled the measurement of fission and gamma-decay probabilities. First results are presented and discussed.

  5. Study of the surrogate-reaction method applied to neutron-induced capture cross sections

    International Nuclear Information System (INIS)

    Boutoux, G.; Jurado, B.; Méot, V.; Roig, O.; Mathieu, L.; Aïche, M.; Barreau, G.; Capellan, N.; Companis, I.; Czajkowski, S.; Schmidt, K.-H.; Burke, J.T.; Bail, A.; Daugas, J.M.; Faul, T.; Morel, P.; Pillet, N.; Théroine, C.; Derkx, X.; Sérot, O.

    2012-01-01

    Gamma-decay probabilities of 173 Yb and 176 Lu have been measured using the surrogate reactions 174 Yb( 3 He,αγ) 173 Yb* and 174 Yb( 3 He,pγ) 176 Lu*, respectively. For the first time, the gamma-decay probabilities have been obtained with two independent experimental methods based on the use of C 6 D 6 scintillators and Germanium detectors. Our results for the radiative-capture cross sections are several times higher than the corresponding neutron-induced data. To explain these differences, we have used our gamma-decay probabilities to extract rather direct information on the spin distributions populated in the transfer reactions used. They are about two times wider and the mean values are 3 to 4 ℏ higher than the ones populated in the neutron-induced reactions. As a consequence, in the transfer reactions neutron emission to the ground and first excited states of the residual nucleus is strongly suppressed and gamma-decay is considerably enhanced.

  6. Measurement of fast neutron induced fission cross section of minor-actinide

    International Nuclear Information System (INIS)

    Hirakawa, Naohiro

    2000-06-01

    In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am, Cm). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA is measured using Dynamitron accelerator in Tohoku University. The followings were performed in this fiscal year; (1) Research of nuclear data of MA, (2) Sample preparation and sample mass assay, (3) Investigation of neutron sources with the energy of several 10 keV, (4) Preliminary measurement of fission cross section using Dynamitron accelerator. As the result, four 237 Np samples were prepared and the sample mass were measured using alpha-spectrometry with the accuracy of 1.2%. Then, it was confirmed that a neutron source via 7 Li(p,n) 7 Be reaction using a Li-thick target is suitable for measuring fission cross section of MA in the energy region of several 10 keV. Furthermore, it was verified by the preliminary measurement that the measurement of fission cross section of MA is available using a fission chamber and electronics developed in this study. (author)

  7. Measurement of fast neutron induced fission cross sections of 232Th, 238U, 237Np and 243Am

    International Nuclear Information System (INIS)

    Kanda, Kazutaka; Sato, Osamu; Yoshida, Kazuo; Imaruoka, Hiromitsu; Terayama, Hiromichi; Yoshida, Masashi; Hirakawa, Naohiro

    1984-01-01

    Neutron induced fission cross sections of 232 Th, 238 U, 237 Np and 243 Am relative to 235 U were measured in the energy range from 1.5 to 6.6 MeV. The present results are compared with experimental results of others and evaluated data in JENDL-2 and ENDF/B-IV. (author)

  8. Determination of the neutron-induced fission cross section of {sup 242}Pu; Bestimmung des neutroneninduzierten Spaltquerschnitts von {sup 242}Pu

    Energy Technology Data Exchange (ETDEWEB)

    Koegler, Toni Joerg

    2016-04-26

    Neutron induced fission cross sections of actinides like the Pu-isotopes are of relevance for the development of nuclear transmutation technologies. For {sup 242}Pu, current uncertainties are of around 21%. Sensitivity studies show that the total uncertainty has to be reduced to below 5% to allow for reliable neutron physics simulations. This challenging task was performed at the neutron time-of-flight facility of the new German National Center for High Power Radiation Sources at HZDR, Dresden. Within the TRAKULA project, thin, large and homogeneous deposits of {sup 235}U and {sup 242}Pu have been produced successfully. Using two consecutively placed fission chambers allowed the determination of the neutron induced fission cross section of {sup 242}Pu relative to {sup 235}U. The areal density of the Plutonium targets was calculated using the measured spontaneous fission rate. Experimental results of the fast neutron induced fission of {sup 242}Pu acquired at nELBE will be presented and compared to recent experiments and evaluated data. Corrections addressing the neutron scattering are discussed by using results of different neutron transport simulations (Geant 4, MCNP 6 and FLUKA).

  9. Measurement of the neutron-induced deuteron breakup reaction cross-section between 5 and 25 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Laborie, J.M.; Ledoux, X.; Varignon, C.; Lazauskas, R.; Morillon, B.; Belier, G. [CEA, DAM, DIF, Arpajon (France)

    2012-06-15

    This article presents a full program devoted to the calculation and the measurement of the neutron-induced deuteron break-up reaction cross-section between 5 and 10 MeV, and between 20 and 25 MeV. Measurements are compared with theoretical calculations based on the solution of the Faddeev equations for a realistic nuclear Hamiltonian. The experiments were performed at the Tandem 7 MV accelerator at CEA, DAM, DIF, France. The measurements were carried out with a C{sub 6}D{sub 6} detector as active deuterium target located at the center of a 4{pi} neutron counter (see C. Varignon et al., Nucl. Instrum. Methods B 248, 329 (2006)) which allows to count the two neutrons emitted in the {sup 2}H(n, 2n)p reaction. Comparisons of the new data and calculations are made with the existing data as well as the CENDL2, JENDL3.3 and ENDF/B-VII evaluations. (orig.)

  10. Neutron-induced Fission Cross Sections of Am and Cm isotopes (Final Report of Research Contract 14485). Resonance and Fast Neutron Induced Fission Cross Sections of Americium and Curium Nuclides (Third-year Progress Report of Research Contract 14485)

    International Nuclear Information System (INIS)

    Alekseev, A.A.; Bergman, A.A.; Berlev, A.I.; Koptelov, E.A.; Egorov, A.S.; Samylin, B.F.; Trufanov, A.M.; Fursov, B.I.; Shorin, V.S.

    2012-01-01

    The neutron induced fission cross sections of Am and Cm isotopes were measured relative to 239 Pu in the neutron energy range from 1 eV to 20 keV at the INR RAS lead slowing down spectrometer LSDS-100. The fission resonance integrals were also estimated using the measured cross section data. The results have been compared with the available experimental and evaluated data. This analysis has shown the present status of the measured fission cross sections and the necessity to revise the evaluated cross sections libraries for the minor actinides. (author)

  11. Nuclear structure effects on calculated fast neutron reaction cross sections

    International Nuclear Information System (INIS)

    Avrigeanu, V.

    1992-01-01

    The importance of accurate low-lying level schemes for reaction cross section calculation and need for microscopically calculated levels are proved with reference to fast neutron induced reactions in the A = 50 atomic mass range. The uses of the discrete levels both for normalization of phenomenological level density approaches and within Hauser-Feshbach calculations are discussed in this respect. (Author)

  12. Application of evaporation model to the calculation of energy spectrum and angular distribution of recoil nuclei from neutron induced reaction

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Y; Sugimoto, M; Sugiyama, K [Tohoku Univ., Sendai (Japan). Faculty of Engineering

    1978-12-01

    Calculated angular distributions and energy spectra from 14.8 MeV neutron induced (n,2n) reactions based on a simple evaporation model were obtained by means of the Monte Carlo method. It was ascertained that the effects on the spectra of the method of determining the nuclear temperature and the value of the level density parameter are much smaller than those of the reaction Q-value and the nuclear mass. As a check on the calculational procedure, results of similar calculations were compared with the experimental recoil escape efficiency for /sup 27/Al(n,..cap alpha..)/sup 24/Na reaction. Distortions of the energy spectra in thick target materials were also obtained. These results suggest that this model is fully applicable to the calculation of primary knock-on atoms distributions from various nuclear reactions.

  13. New approach to neutron-induced transmutation, radioactivity and afterheat calculations and its application to fusion reactors

    International Nuclear Information System (INIS)

    Fukumoto, Hideshi

    1986-01-01

    A new method and an accompanying computer code CINAC have been developed for the calculation of neutron-induced transmutation, radioactivity and afterheat. In the method, the generation and depletion of nuclides during and after reactor operation are described in a matrix form in which the arrangement of nuclides is determined systematically. The solutions are obtained by an eigenvalue analysis of the matrix without any time steps or iterative schemes which would increase the computational time. The method can treat any type of activation chains equally, and it gives analytical solutions for linear chains. The CINAC code, coupled with the radiation transport codes ANISN and DOT3.5, can also calculate the dose distribution at any time after shutdown in a one- or two-dimensional geometry of fusion reactors. Two calculations were carried out using CINAC to confirm its validity. The results were compared to those calculated by the THIDA code system which is based on a matrix exponential method. The new method was 50 times faster than the latter, while the discrepancy between them was 4 % at the most. (author)

  14. Cross Sections for Neutron-induced Reactions on Actinide Targets Extracted from Surrogate Experiments: A Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Escher, J E; Burke, J T; Dietrich, F S; Lesher, S R; Scielzo, N D; Thompson, I J; Younes, W

    2009-10-01

    The Surrogate nuclear reactions method, an indirect approach for determining cross sections for compound-nuclear reactions involving difficult-to-measure targets, is reviewed. Focusing on cross sections for neutron-induced reactions on actinides, we review the successes of past and present applications of the method and assess its uncertainties and limitations. The approximations used in the analyses of most experiments work reasonably well for (n,f) cross sections for neutron energies above 1-2 MeV, but lead to discrepancies for low-energy (n,f) reactions, as well as for (n,{gamma}) applications. Correcting for some of the effects neglected in the approximate analyses leads to improved (n,f) results. We outline steps that will further improve the accuracy and reliability of the Surrogate method and extend its applicability to reactions that cannot be approached with the present implementation of the method.

  15. Cross Sections for Neutron-induced Reactions on Actinide Targets Extracted from Surrogate Experiments: A Status Report

    International Nuclear Information System (INIS)

    Escher, J.E.; Burke, J.T.; Dietrich, F.S.; Lesher, S.R.; Scielzo, N.D.; Thompson, I.J.; Younes, W.

    2009-01-01

    The Surrogate nuclear reactions method, an indirect approach for determining cross sections for compound-nuclear reactions involving difficult-to-measure targets, is reviewed. Focusing on cross sections for neutron-induced reactions on actinides, we review the successes of past and present applications of the method and assess its uncertainties and limitations. The approximations used in the analyses of most experiments work reasonably well for (n,f) cross sections for neutron energies above 1-2 MeV, but lead to discrepancies for low-energy (n,f) reactions, as well as for (n,γ) applications. Correcting for some of the effects neglected in the approximate analyses leads to improved (n,f) results. We outline steps that will further improve the accuracy and reliability of the Surrogate method and extend its applicability to reactions that cannot be approached with the present implementation of the method.

  16. Resonance analysis and evaluation of the 235U neutron induced cross sections

    International Nuclear Information System (INIS)

    Leal, L.C.

    1990-06-01

    Neutron cross sections of fissile nuclei are of considerable interest for the understanding of parameters such as resonance absorption, resonance escape probability, resonance self-shielding,and the dependence of the reactivity on temperature. In the present study, new techniques for the evaluation of the 235 U neutron cross sections are described. The Reich-Moore formalism of the Bayesian computer code SAMMY was used to perform consistent R-matrix multilevel analyses of the selected neutron cross-section data. The Δ 3 -statistics of Dyson and Mehta, along with high-resolution data and the spin-separated fission cross-section data, have provided the possibility of developing a new methodology for the analysis and evaluation of neutron-nucleus cross sections. The results of the analysis consists of a set of resonance parameters which describe the 235 U neutron cross sections up to 500 eV. The set of resonance parameters obtained through a R-matrix analysis are expected to satisfy statistical properties which lead to information on the nuclear structure. The resonance parameters were tested and showed good agreement with the theory. It is expected that the parametrization of the 235 U neutron cross sections obtained in this dissertation represents the current state of art in data as well as in theory and, therefore, can be of direct use in reactor calculations. 44 refs., 21 figs., 8 tabs

  17. Neutron induced reaction cross-sections on 115In at around 14 MeV

    International Nuclear Information System (INIS)

    Csikai, J.; Lantos, Z.; Buczko, C.M.; Sudar, S.

    1990-01-01

    A systematic investigation was carried out on 115 In isotope to determine the contribution of different reactions to the total non-elastic cross-section in the 13.43 and 14.84 MeV range. All the major component cross-sections of σ NE were measured with exception of the σ g (n,n'). In the knowledge of σ NE , the energy dependence of σ g (n,n') could be deduced. The isomeric cross section ratios both for (n,2n) and (n,n') processes were also determined in the given energy range. The present experiment proves the dependence of σ m /(σ g +σ m ) ratio on the spin value (I m ) of the isomeric state in (n,2n) reaction. Excitation functions of (n,2n), (n,n') and (n,ch) reactions were compared with results calculated by STAPRE code. (author). 37 refs, 5 figs, 4 tabs

  18. The Determination of Neutron-Induced Reaction Cross Section Data on Even-Even, Magic- Number Nuclide Chromium 52 Using EXIFON Code

    International Nuclear Information System (INIS)

    Jonah, S.A.

    2013-01-01

    The EXIFON code version 2.0 is a calculational tool, which is based on both many-body theory and random matrix physics. In this work, it has been used to calculate neutron induced reaction cross section data from 0 to 20 MeV on an even-even, magic number nuclide 52 Cr with neutron number, N=28. Specifically, the (n,p), (n,α) and (n,2n) reaction cross section data were calculated as functions of incident energy of neutrons. Data obtained from the experimental data in the IAEA, EXFOR data Library and recommended data libraries around the globe, JENDL, ENDF and JEFF were used to validate the calculated data. The data indicate that the calculated data without shell corrections are in good agreement with experimental data as well as the recommended data from the evaluated data libraries. The calculated results could provide useful insight into the choice of some input parameters near closed shells using the EXIFON code.

  19. Evaluation of Neutron-induced Cross Sections and their Related Covariances with Physical Constraints

    Science.gov (United States)

    De Saint Jean, C.; Archier, P.; Privas, E.; Noguère, G.; Habert, B.; Tamagno, P.

    2018-02-01

    Nuclear data, along with numerical methods and the associated calculation schemes, continue to play a key role in reactor design, reactor core operating parameters calculations, fuel cycle management and criticality safety calculations. Due to the intensive use of Monte-Carlo calculations reducing numerical biases, the final accuracy of neutronic calculations increasingly depends on the quality of nuclear data used. This paper gives a broad picture of all ingredients treated by nuclear data evaluators during their analyses. After giving an introduction to nuclear data evaluation, we present implications of using the Bayesian inference to obtain evaluated cross sections and related uncertainties. In particular, a focus is made on systematic uncertainties appearing in the analysis of differential measurements as well as advantages and drawbacks one may encounter by analyzing integral experiments. The evaluation work is in general done independently in the resonance and in the continuum energy ranges giving rise to inconsistencies in evaluated files. For future evaluations on the whole energy range, we call attention to two innovative methods used to analyze several nuclear reaction models and impose constraints. Finally, we discuss suggestions for possible improvements in the evaluation process to master the quantification of uncertainties. These are associated with experiments (microscopic and integral), nuclear reaction theories and the Bayesian inference.

  20. Neutron-induced photon production in MCNP

    International Nuclear Information System (INIS)

    Little, R.C.; Seamon, R.E.

    1983-01-01

    An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems

  1. Neutron induced fission cross sections for 232Th, 235,238U, 237Np, and 239Pu

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Ullmann, J.L.; Balestrini, S.J.; Hill, N.W.; Carlson, A.D.; Wasson, O.A.

    1989-01-01

    Neutron-induced fission cross section ratios for samples of 232 Th, 235,238 U, 237 Np and 239 Pu have been measured from 1 to 400 MeV. The fission reaction rate was determined for all samples simultaneously using a fast parallel plate ionization chamber at a 20-m flight path. A well characterized annular proton recoil telescope was used to measure the neutron fluence from 3 to 30 MeV. Those data provided the shape of the 235 U(n,f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for 235 U(n,f) at 14.178 MeV. From 30 to 400 MeV cross section values were determined using the neutron fluence measured with a plastic scintillator. Cross section values of 232 Th, 235,238 U, 237 Np and 239 Pu were computed from the ratio data using the authors' values for 235 U(n,f). In addition to providing new results at high neutron energies, these data highlight several areas of deficiency in the evaluated nuclear data files and provide new information for the 235 U(n,f) standard

  2. Measurement and analysis of 14 MeV neutron-induced double-differential neutron emission cross sections needed for fission and fusion reactor technology

    International Nuclear Information System (INIS)

    Wang Dahai.

    1990-10-01

    The main objectives of this IAEA Co-ordinated Research Programme are to improve the data on 14 MeV neutron-induced double-differential neutron emission cross sections for materials needed for fission and fusion reactor technology. This report summarizes the conclusions and recommendations which were agreed by all participants during the Second Research Co-ordination Meeting

  3. Isomeric cross sections of neutron induced reactions on Ge and Ir isotopes

    International Nuclear Information System (INIS)

    Vlastou, R.; Papadopoulos, C.T.; Kokkoris, M.; Perdikakis, G.; Galanopoulos, S.; Patronis, N.; Serris, M.; Perdikakis, G.; Harissopulos, S.; Demetriou, P.

    2008-01-01

    The 72 Ge(n,α) 69m Zn, 74 Ge(n,α) 71m Zn, 76 Ge(n,2n) 75g+m Ge and 191 Ir(n,2n) 190 Ir g+m1 and 191 Ir(n,2n) 190 Ir m2 reaction cross sections were measured from 9.6 to 11.4 MeV relative to the 27 Al(n,α) 24 Na reference reaction via the activation method. The quasi-monoenergetic neutron beams were produced via the 2 H(d,n) 3 He reaction at the 5 MV VdG Tandem T11/25 accelerator of NCSR 'Demokritos'. Statistical model calculations using the codes STAPRE-F and EMPIRE (version 2.19) and taking into account pre-equilibrium emission were performed on the data measured in this work as well as on data reported in literature. (authors)

  4. Test of RIPL-2 cross section calculations

    International Nuclear Information System (INIS)

    Herman, M.

    2002-01-01

    The new levels and optical segments and microscopic HF-BCS level densities (part of the density segment) were tested in practical calculations of cross sections for neutron induced reactions on 22 targets (40-Ca, 47-Ti, 52-Cr, 55-Mn, 58-Ni, 63-Cu, 71-Ga, 80-Se, 92-Mo, 93-Nb, 100-Mo, 109-Ag, 114-Cd, 124-Sn, 127-I, 133-Cs, 140-Ce, 153-Eu, 169-Tm, 186-W, 197-Au, 208-Pb). For each target all reactions involving up to 3 neutron, 1 proton and 1 α-particle emissions (subject to actual reaction thresholds) were considered in the incident energy range from 1 keV up to 20 MeV (in some cases up to 27 MeV). In addition, total, elastic, and neutron capture cross sections were calculated

  5. The shell structure effects in neutron cross section calculation by a ...

    African Journals Online (AJOL)

    The role of the shell structure properties of the nucleus in the calculation of neutron-induced reaction cross-section data based on nuclear reaction theory has been investigated. In this investigation, measured, evaluated and calculated (n.p) reaction cross-section data on la spherical nucleus (i.e. 112Sn) and a deformed ...

  6. Measurements of D-T neutron induced radioactivity in plasma-facing materials and their role in qualification of activation cross-section libraries and codes

    International Nuclear Information System (INIS)

    Kumar, A.; Abdou, M.A.; Kosako, K.; Oyama, Y.; Nakamura, T.; Maekawa, H.

    1995-01-01

    The D-T neutron-induced radioactivity constitutes one of the foremost issues in fusion reactor design. The validation of activation cross-sections and decay data libraries is one of the important requirements for validating ITER design from safety and waste disposal viewpoints. An elaborate, experimental program was initiated in 1988, under USDOE-JAERI collaborative program, to validate the radioactivity codes/libraries. The measurements of decay-γ spectra from irradiated, high purity samples of Al, Si, Ti, V, Cr, Mn-Cu alloy, Fe, Co, Ni, Cu, stainless steel 316 (AISI 316), Zn, Zr, Nb, Mo, In, Sn, Ta, W, and Pb, among others, were conducted under D-T neutron fluences varying from 1.6 x 10 10 ncm -2 to 6.1 x 10 13 ncm -2 . As many as 14 neutron energy spectra were covered for a number of materials. The analysis of isotopic activities of the irradiated materials using activation cross-section libraries of four leading radioactivity codes, i.e. ACT4/THIDA-2, REAC-3, DKR-ICF, and RACC, has shown large discrepancies among the calculations, on the one hand, and between the calculations and the measurements, on the other. A discussion is also presented on definition and obtention of safety cum quality factors for various activation libraries. (orig.)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  8. Using a Time Projection Chamber to Measure High Precision Neutron-Induced Fission Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Manning, Brett [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-06

    2014 LANSCE run cycle data will provide a preliminary 239Pu(n,f) cross section and will quantify uncertainties: PID and Target/beam non-uniformities. Continued running during the 2015 LANSCE run cycle: Thin targets to see both fission fragments and 239Pu(n,f) cross section and fully quantified uncertainties

  9. Binary and tertiary neutron induced reaction cross sections of chromium and iron

    International Nuclear Information System (INIS)

    Garg, S.B.

    1989-01-01

    Investigation has been carried out for the following binary and tertiary reaction cross-sections of Cr-52 and Fe-56: (n,p), (n,pn), (n,np), (n,α), (n, nα), (n, 2n) and (n, 3n), energy spectra of the emitted neutron, proton, α-particle and γ-rays, angle-energy correlated double differential cross-sections for the secondary emitted neutrons and total production cross-sections for neutron, hydrogen, helium and gamma-rays. 12 refs, 20 figs, 1 tab

  10. Integrated system for production of neutronics and photonics calculational constants. Supplemental neutron-induced interactions (Z less than or equal to 35): graphical, experimental data

    International Nuclear Information System (INIS)

    Cullen, D.E.; Howerton, R.J.; MacGregor, M.H.; Perkins, S.T.

    1976-01-01

    This report (Vol. 8) presents graphs of supplemental neutron-induced cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists of interactions where more than one data set is needed to show cross-section behavior. In contrast, Vol. 7 of this UCRL-50400 series consists primarily of interactions where a single data set contains enough points to show cross-section behavior. In Vol. 7 can be found the total, elastic, capture, and fission cross sections (along with the parameters anti ν, α, and eta). Volume 8 contains all other reactions. Data are plotted with associated cross-section error bars (when given) and compared with the Evaluated Nuclear Data Library (ENDL) as of July 4, 1976. The plots are arranged in ascending order of atomic number (Z) and atomic weight (A). Part A contains the plots for Z = 1 to 35; Part B contains the plots for Z greater than 35

  11. Integrated system for production of neutronics and photonics calculational constants. Major neutron-induced interactions (Z less than or equal to 55): graphical, experimental data

    International Nuclear Information System (INIS)

    Cullen, D.E.; Howerton, R.J.; MacGregor, M.H.; Perkins, S.T.

    1976-01-01

    This report (vol. 7) presents graphs of major neutron-induced interaction cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists primarily of interactions where a single data set contains enough points to show cross-section behavior. In contrast, vol. 8 of this UCRL-50400 series consists of interactions where more than one data set is needed to show cross section behavior. Thus, you can find the total, elastic, capture, and fission cross sections (along with the parameters anti ν, α, and eta) in vol. 7 and all other reactions in vol. 8. Data are plotted with associated cross section error bars (when given) and compared with the Evaluated Nuclear Data Library (ENDL) as of July 4, 1976. The plots are arranged in ascending order of atomic number (Z) and atomic weight (A). Part A contains the plots for Z = 1 to 55; Part B contains the plots for Z greater than 55

  12. Integrated system for production of neutronics and photonics calculational constants. Major neutron-induced interactions (Z less than or equal to 55): graphical, experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D.E.; Howerton, R.J.; MacGregor, M.H.; Perkins, S.T.

    1976-07-04

    This report (vol. 7) presents graphs of major neutron-induced interaction cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists primarily of interactions where a single data set contains enough points to show cross-section behavior. In contrast, vol. 8 of this UCRL-50400 series consists of interactions where more than one data set is needed to show cross section behavior. Thus, you can find the total, elastic, capture, and fission cross sections (along with the parameters anti ..nu.., ..cap alpha.., and eta) in vol. 7 and all other reactions in vol. 8. Data are plotted with associated cross section error bars (when given) and compared with the Evaluated Nuclear Data Library (ENDL) as of July 4, 1976. The plots are arranged in ascending order of atomic number (Z) and atomic weight (A). Part A contains the plots for Z = 1 to 55; Part B contains the plots for Z greater than 55.

  13. Neutron induced fission cross section ratios for 232Th, 235,238U, 237Np and 239Pu from 1 to 400 MeV

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Ullmann, J.L.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1988-01-01

    Time-of-flight measurements of neutron induced fission cross section ratios for 232 Th, 235,238 U, 237 Np, and 239 Pu, were performed using the WNR high intensity spallation neutron source located at Los Alamos National Laboratory. A multiple-plate gas ionization chamber located at a 20-m flight path was used to simultaneously measure the fission rate for all samples over the energy range from 1 to 400 MeV. Because the measurements were made with nearly identical neutron fluxes, we were able to cancel many systematic uncertainties present in previous measurements. This allows us to resolve discrepancies among different data sets. In addition, these are the first neutron-induced fission cross section values for most of the nuclei at energies above 30 MeV. (author)

  14. Neutron induced fission cross section ratios for 232Th, /sup 235,238/U, 237Np, and 239Pu from 1 to 400 MeV

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Ullmann, J.L.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1988-01-01

    Time-of-flight measurements of neutron induced fission cross section ratios for 232 Th, /sup 235,238/U, 237 Np, and 239 Pu, were performed using the WNR high intensity spallation neutron source located at Los Alamos National Laboratory. A multiple-plate gas ionization chamber located at a 20-m flight path was used to simultaneously measure the fission rate for all samples over the energy range from 1 to 400 MeV. Because the measurements were made with nearly identical neutron fluxes, we were able to cancel many systematic uncertainties present in previous measurements. This allows us to resolve discrepancies among different data sets. In addition, these are the first neutron-induced fission cross section values for most of the nuclei at energies above 30 MeV. 8 refs., 3 figs

  15. 14.2 MeV neutron induced U-235 fission cross section measurement

    International Nuclear Information System (INIS)

    Li Jingwen; Shen Guanren; Ye Zongyuan; Li Anli; Zhou Shuhua; Sun Zhongfan; Wu Jingxia; Huang Tanzi

    1986-01-01

    The cross section of U-235 fission induced by 14.2 MeV neutrons was measured by the time correlated associated particle method. The result obtained is (2.078+-0.040) barn. Comparison with other author's is also given. (author)

  16. Neutron-induced fission cross sections of uraniums up to 40 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Maslov, V.M. [Radiation Physics and Chemistry Problems Inst., Minsk-Sosny (Belarus); Hasegawa, A.

    1998-11-01

    Statistical theory of nuclear reactions, well-proved below 20 MeV, is applied for {sup 235}U and {sup 238}U fission data analysis up to {approx}40 MeV. It is shown that measured data could be reproduced. Chance structure of measured fission cross section is provided, it`s validity is supported by description of data for competing (n,xn)-reactions. Role of fissility of target nucleus is addressed. It seems that gap in incident neutron energy interval of 20 MeV - 50 MeV, below which evaluation approaches are well-developed, and above which simplified statistical approaches are valid, could be covered. (author)

  17. Evidence of pair correlations in actinide neutron-induced fission cross sections

    International Nuclear Information System (INIS)

    Maslov, V.M.

    2000-01-01

    It is shown that irregularities in fission cross sections in MeV incident neutron energy region could be attributed to the interplay of few-quasiparticle excitations in the level density of the fissioning and residual nuclei. It is suggested the intrinsic quasiparticle state density modelling approach both at stable and saddle-point deformations. The experimental manifestation of few-quasiparticle irregularities in the level density depends on the fission barrier structure and internal excitation energy at the saddle point, corresponding to the higher barrier hump. The explicit evidence is observed in case of fissile and non-fissile target nuclides [ru

  18. Actinide neutron induced cross section measurements using the oscillation technique in the Minerve reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, B.; Leconte, P.; Gruel, A.; Antony, M.; Di-Salvo, J.; Hudelot, J.P.; Pepino, A.; Lecluze, A. [CEA Cadarache, DEN/CAD/DER/SPRC/LEPh, 13 - Saint-Paul-lez-Durance (France)

    2009-07-01

    CEA is deeply involved research programs concerning nuclear fuel advanced studies (actinides, plutonium), waste management, the scientific and technical support of French PWR reactors and EPR reactor, and innovative systems. In this framework, specific neutron integral experiments have been carried out in the critical ZPR (zero power reactor) facilities of the CEA at Cadarache such as MINERVE, EOLE and MASURCA. This paper deals with MINERVE Pool Reactor experiments. MINERVE is mainly devoted to neutronics studies of different reactor core types. The aim is to improve the knowledge of the integral absorption cross sections of actinides (OSMOSE program), of new absorbers (OCEAN program) and also for fission Products (CBU program) in thermal, epithermal and fast neutron spectra. (authors)

  19. Systematics of neutron-induced fission cross sections over the energy range 0.1 through 15 MeV, and at 0.0253 eV

    International Nuclear Information System (INIS)

    Behrens, J.W.

    1977-01-01

    Recent studies have shown straightforward systematic behavior as a function of constant proton and neutron number for neutron-induced fission cross sections of the actinide elements in the incident-neutron energy range 3 to 5 MeV. In this report, the second in a series, fission cross-section values are studied over the MeV incident-neutron energy range, and at 0.0253 eV. Fission-barrier heights and neutron-binding energies are correlated by constant proton and neutron number; however, these systematic behaviors alone do not explain the trends observed in the fission cross-section values

  20. Improvement of measurements, theoretical computations and evaluations of neutron induced helium production cross sections. Summary report on the third and final research co-ordination meeting

    International Nuclear Information System (INIS)

    Pashchenko, A.B.

    1996-09-01

    The present report contains the Summary of the Third and Final IAEA Research Co-ordination Meeting (RCM) on ''Improvement of Measurements, Theoretical Computations and Evaluations of Neutron Induced Helium Production Cross Sections'' which was hosted by the Tohoku University and held in Sendai, Japan, from 25 to 29 September 1995. This RCM was organized by the IAEA Nuclear Data Section (NDS), with the co-operation and assistance of local organizers from Tohoku University. Summarized are the proceedings and results of the meeting. The List of Participants and meeting Agenda are included. (author)

  1. Measurement and analysis of 14 MeV neutron-induced double-differential neutron emission cross sections needed for fission and fusion reactor technology

    International Nuclear Information System (INIS)

    Wang Dahai; Mehta, M.K.

    1988-07-01

    The main objectives of this IAEA Co-ordinated Research Programme are to improve the current status of data for 14 MeV neutron-induced double-differential neutron emission cross sections for V, Cr, Fe, Nb, Ta and 238 U. The principal objectives of this first meeting were to report on the status of participants' work, to exchange experience in experimental work and to establish the future work. Considering the unsatisfactory status of the data for 6 Li, 7 Li, 9 Be, Mo, W and Bi and their importance in fusion reactor technology participants agreed to include these isotopes in the programme

  2. Calculation of the intermediate energy activation cross section

    Energy Technology Data Exchange (ETDEWEB)

    Furihata, Shiori; Yoshizawa, Nobuaki [Mitsubishi Research Inst., Inc., Tokyo (Japan)

    1997-03-01

    We discussed the activation cross section in order to predict accurately the activation of soil around an accelerator with high energy and strong intensity beam. For the assessment of the accuracy of activation cross sections estimated by a numerical model, we compared the calculated cross section with various experimental data, for Si(p,x){sup 22}Na, Al(p,x){sup 22}Na, Fe(p,x){sup 22}Na, Si(p,x){sup 7}Be, O(p,x){sup 3}H, Al(p,x){sup 3}H and Si(p,x){sup 3}H reactions. We used three computational codes, i.e., quantum molecular dynamics (QMD) plus statistical decay model (SDM), HETC-3STEP and the semiempirical method developed by Silberberg et.al. It is observed that the codes are accurate above 1GeV, except for {sup 7}Be production. We also discussed the difference between the activation cross sections of proton- and neutron-induced reaction. For the incident energy at 40MeV, it is found that {sup 3}H production cross sections of neutron-induced reaction are ten times as large as those of proton-induced reaction. It is also observed that the choice of the activation cross sections seriously affects to the estimate of saturated radioactivity, if the maximum energy of neutron flux is below 100MeV. (author)

  3. Resonant neutron-induced atomic displacements

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-01

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

  4. A new empirical formula for 14-15 MeV neutron-induced (n, p) reaction cross sections

    International Nuclear Information System (INIS)

    Tel, E; Sarer, B; Okuducu, S; Aydin, A; Tanir, G

    2003-01-01

    In this study, we have suggested a new empirical formula to reproduce the cross sections of the (n, p) reactions at 14-15 MeV for the neutron incident energy. This formula obtained using the asymmetry parameters represents a modification to the original formula of Levkovskii. The resulting modified formulae yielded cross sections, representing smaller χ 2 deviations from experimental values, and values much closer to unity as compared with those calculated using Levkovskii's original formula. The results obtained have been discussed and compared with the existing formulae, and found to be well in agreement, when used to correlate the available experimental σ(n, p) data of different nuclei

  5. Integrated system for production of neutronics and photonics calculational constants. Neutron-induced interactions: bibliography of experimental data

    International Nuclear Information System (INIS)

    MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.

    1976-01-01

    The bibliographic citations in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has three arrangements: alphabetically by author, alphabetically by publication, and numerically by reference number

  6. Integrated system for production of neutronics and photonics calculational constants. Neutron-induced interactions: bibliography of experimental data

    Energy Technology Data Exchange (ETDEWEB)

    MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.

    1976-07-04

    The bibliographic citations in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has three arrangements: alphabetically by author, alphabetically by publication, and numerically by reference number.

  7. Neutron-Induced Fission Cross Section of Uranium, Americium and Curium Isotopes. Progress report - Research Contract 14485, Coordinated Research Project on Minor Actinide Neutron Reaction Data (MANREAD)

    International Nuclear Information System (INIS)

    Alekseev, A.A.; Bergman, A.A.; Berlev, A.I.; Koptelov, E.A.; Samylin, B.F.; Trufanov, A.M.; Fursov, B.I.; Shorin, V.S.

    2009-12-01

    This report contains brief description of the Lead Slowing Down Spectrometer and results of measurements of neutron-induced fission cross sections for 236 U, 242m Am, 243 Cm, 244 Cm, 245 Cm and 246 Cm done at this spectrometer. The work was partially supported through the IAEA research contract RC-14485-RD in the framework of the IAEA Coordinated Research Project 'Minor Actinide Neutron Reaction Data (MANREAD)'. The detailed description of the experimental set up, measurements procedure and data treatment can be found in the JIA-1182 (2007) and JIA-1212 (2009) reports from the Institute of Nuclear Research of the Russian Academy of Science published in Russian. Part 1 contains the first year report of the research contract and part 2 the second year report. (author)

  8. Preliminary neutron shielding calculations of the electronics in the EAST BES systems focusing on neutron induced displacement damage

    Energy Technology Data Exchange (ETDEWEB)

    Náfrádi, Gábor, E-mail: nafradi@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Kovácsik, Ákos, E-mail: kovacsik.akos@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Németh, József, E-mail: nemeth.jozsef@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary); Pór, Gábor, E-mail: por@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Zoletnik, Sándor, E-mail: zoletnik.sandor@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary)

    2016-11-15

    Monte Carlo N-Particle (MCNP) calculations were carried out to compare neutron shielding capabilities of three frequently used neutron shielding materials: polyethylene without neutron absorbers, polyethylene with boron absorbers and polyethylene with lithium absorbers, according to Non Ionizing Energy Loss (NIEL). The results of 1D shielding calculations showed that simple neutron moderating materials can provide sufficient and cheap shielding against 2.45 MeV and 14.1 MeV fusion neutrons, in terms of 1 MeV neutron equivalent flux, in silicon targets, which is the most commonly used material of electronic components. Based on these results a new shielding concept is proposed which can be taken into consideration where the reduction of displacement damage is the main goal and the free space available for shielding is limited. Based on this shielding concept detailed 3D calculations were carried out to describe the properties of the neutron shielding of the Beam Emission Spectroscopy (BES) system installed at the EAST tokamak.

  9. 238U neutron-induced fission cross section for incident neutron energies between 5 eV and 3.5 MeV

    International Nuclear Information System (INIS)

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

    1979-01-01

    A measurement of the 238 U neutron-induced fission cross section was performed at the ORELA Linac facility in the neutron energy range between 5 eV and 3.5 MeV. The favorable signal-to-background ratio and high resolution of this experiment resulted in the identificaion of 85 subthreshold fission resonances or clusters of resonances in the neutron energy region between 5 eV and 200 keV. The fission data below 100 keV are characteristic of a weak coupling situation between Class I and Class II levels. The structure of the fission levels at the 720 eV and 1210 eV fission clusters is discussed. There is an apparent enhancement of the fission cross section at the opening of the 2 + neutron inelastic channel in 238 U at 45 keV. An enhancement of the subthreshold fission cross section between 100 keV and 200 keV is tentatively interpreted in terms of the presence of a Class II, partially damped vibrational level. There is a marked structure in the fission cross section above 200 keV up to and including the plateau between 2 and 3.5 MeV. 11 figures and 6 tables

  10. Neutron induced fission cross sections for /sup 232/Th, /sup 235,238/U, /sup 237/Np and /sup 239/Pu from 1 to 400 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, P.W.; Ullmann, J.L.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1988-01-01

    Neutron-induced fission cross section ratios for samples of /sup 232/Th, /sup 235,238/U, /sup 237/Np and /sup 239/Pu have been measured from 1 to 400 MeV. The fission reaction rate was determined for all samples simultaneously using a fast parallel plate ionization chamber at a 20-m flight path. A well characterized annular proton recoil telescope was used to measure the neutron fluence up to 30 MeV. These data provided the shape of the /sup 235/U(n,f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known values were determined using the neutron fluence measured with a second proton recoil telescope. Cross section values for /sup 232/Th, /sup 238/U, /sup 237/Np, and /sup 239/Pu were computed from the ratio data using our values for /sup 235/U(n,f). In addition to providing new results at high neutron energies, these data resolve long standing discrepancies among different data sets. 1 ref., 1 fig.

  11. Measurements of neutron-induced capture and fission reactions on $^{235}$ U: cross sections and ${\\alpha}$ ratios, photon strength functions and prompt ${\\gamma}$-ray from fission

    CERN Multimedia

    We propose to measure the neutron-induced capture cross section of the fissile isotope $^{235}$U using a fission tagging set-up. This new set-up has been tested successfully in 2010 and combines the n_TOF 4${\\pi}$ Total Absorption Calorimeter (TAC) with MicroMegas (MGAS) fission detectors. It has been proven that such a combination of detectors allows distinguishing with very good reliability the electromagnetic cascades from the capture reactions from dominant ${\\gamma}$-ray background coming from the fission reactions. The accurate discrimination of the fission background is the main challenge in the neutron capture cross section measurements of fissile isotopes. The main results from the measurement will be the associated capture cross section and ${\\alpha}$ ratio in the resolved (0.3-2250 eV) and unresolved (2.25-30 keV) resonance regions. According to the international benchmarks and as it is mentioned in the NEA High Priority Request List (HPRL), the 235U(n,${\\gamma}$) cross section is of utmost impo...

  12. Actinide neutron induced cross-sections; analysis of the OSMOSE LWR-UO{sub 2} experiment in MINERVE

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, D.; Litaize, O.; Santamarina, A.; Antony, M.; Hudelot, J. P. [Commissariat a l' Energie Atomique, Cadarache, DEN/DER, 13108 Saint-Paul-Lez-Durance (France)

    2006-07-01

    This paper describes the interpretation of the first phase of the OSMOSE experimental program. The OSMOSE experiment began in 2005 in the MINERVE French facility and will continue until 2008. It consists in reactivity worth measurements of separated actinides by an oscillation technique. First results are obtained in a standard LWR neutron spectrum (UO{sub 2} lattice). The present study focuses on the following isotopes: {sup 234,236}U, {sup 237}Np, {sup 239,242}Pu. The comparison between APOLLO2 accurate deterministic calculations and experiments shows the reliability of the latest JEFF-3.1 European nuclear data library for all oscillated isotopes, except {sup 237}Np. The obtained (C/E-1){+-}({delta}E/E) values are the following: {sup 234}U: -5%{+-}2% {sup 237}Np: -11%{+-}2% {sup 239}Pu: +1%{+-}2% {sup 242}Pu: +2%{+-}2% An energetic decomposition of the reactivity worth is carried out using Standard Perturbation Theory that underlines the underestimation of the {sup 237}Np(n, {gamma}) thermal and resonant capture cross-section. (authors)

  13. Simultaneous measurement of neutron-induced fission and capture cross sections for {sup 241}Am at neutron energies below fission threshold

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, K., E-mail: hirose.kentaro@jaea.go.jp [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nishio, K.; Makii, H.; Nishinaka, I.; Ota, S. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nagayama, T. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Engineering, Ibaraki University, Mito 310-0056 (Japan); Tamura, N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Goto, S. [Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Andreyev, A.N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Vermeulen, M.J. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Gillespie, S.; Barton, C. [Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Kimura, A.; Harada, H. [Nuclear Science and Engineering Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Meigo, S. [J-PARC Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Chiba, S. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo 152-8550 (Japan); Ohtsuki, T. [Research Reactor Institute, Kyoto University, Kumatori-cho S' ennangun,Osaka 590-0494 (Japan)

    2017-06-01

    Fission and capture reactions were simultaneously measured in the neutron-induced reactions of {sup 241}Am at the spallation neutron facility of the Japan Proton Accelerator Research Complex (J-PARC). Data for the neutron energy range of E{sub n}=0.1–20 eV were taken with the TOF method. The fission events were observed by detecting prompt neutrons accompanied by fission using liquid organic scintillators. The capture reaction was measured by detecting γ rays emitted in the deexcitation of the compound nuclei using the same detectors, where the prompt fission neutrons and capture γ rays were separated by a pulse shape analysis. The cross sections were obtained by normalizing the relative yields at the first resonance to evaluations or other experimental data. The ratio of the fission to capture cross sections at each resonance is compared with those from an evaluated nuclear data library and other experimental data. Some differences were found between the present values and the library/literature values at several resonances.

  14. Neutron induced current pulses in fission chambers

    International Nuclear Information System (INIS)

    Taboas, A.L.; Buck, W.L.

    1978-01-01

    The mechanism of neutron induced current pulse generation in fission chambers is discussed. By application of the calculated detector transfer function to proposed detector current pulse shapes, and by comparison with actually observed detector output voltage pulses, a credible, semi-empirical, trapezoidal pulse shape of chamber current is obtained

  15. Measurements of the neutron-induced fission cross sections of 240Pu and 242Pu relative to 235U

    International Nuclear Information System (INIS)

    Behrens, J.W.; Browne, J.C.; Carlson, G.W.

    1976-01-01

    A continuation is given of the fission-cross-section ratio measurements in progress at the Lawrence Livermore Laboratory. Preliminary results are provided for the 240 Pu/ 235 U and 242 Pu/ 235 U ratios from 0.02 to 30 MeV and 0.1 to 30 MeV, respectively. Using the threshold-cross-section method, the ratios were normalized to the values 1.368 +- 0.030 and 1.116 +- 0.025, respectively, from 1.75 to 4.00 MeV

  16. Neutron induced fission cross section ratios for /sup 232/Th, /sup 235,238/U, /sup 237/Np, and /sup 239/Pu from 1 to 400 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, P.W.; Ullmann, J.L.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1988-01-01

    Time-of-flight measurements of neutron induced fission cross section ratios for /sup 232/Th, /sup 235,238/U, /sup 237/Np, and /sup 239/Pu, were performed using the WNR high intensity spallation neutron source located at Los Alamos National Laboratory. A multiple-plate gas ionization chamber located at a 20-m flight path was used to simultaneously measure the fission rate for all samples over the energy range from 1 to 400 MeV. Because the measurements were made with nearly identical neutron fluxes, we were able to cancel many systematic uncertainties present in previous measurements. This allows us to resolve discrepancies among different data sets. In addition, these are the first neutron-induced fission cross section values for most of the nuclei at energies above 30 MeV. 8 refs., 3 figs.

  17. Calculation of the Reaction Cross Section for Several Actinides

    International Nuclear Information System (INIS)

    Hambsch, Franz-Josef; Oberstedt, Stephan; Vladuca, Gheorghita; Tudora, Anabella; Filipescu, Dan

    2005-01-01

    New, self-consistent, neutron-induced reaction cross-section calculations for 235,238U, 237Np, and 231,232,233Pa have been performed. The statistical model code STATIS was extended to take into account the multi-modality of the fission process. The three most dominant fission modes, the two asymmetric standard I (S1) and standard II (S2) modes, and the symmetric superlong (SL) mode have been taken into account. De-convoluted fission cross sections for these modes in 235,238U(n,f) and 237Np(n,f) based on experimental branching ratios, were calculated for the first time up to the second chance fission threshold. For 235U(n,f) and 233Pa(n,f), the calculations being made up to 50 MeV and 20 MeV incident neutron energy, respectively, higher fission chances have been considered. This implied the need for additional calculations for the neighbouring isotopes.As a side product also mass yield distributions could be calculated at energies hitherto not accessible by experiment. Experimental validation of the predictions is being envisaged

  18. Model calculations as one means of satisfying the neutron cross-section requirements of the CTR program

    International Nuclear Information System (INIS)

    Gardner, D.G.

    1975-01-01

    A large amount of cross section and spectral information for neutron-induced reactions will be required for the CTR design program. To undertake to provide the required data through a purely experimental measurement program alone may not be the most efficient way of attacking the problem. It is suggested that a preliminary theoretical calculation be made of all relevant reactions on the dozen or so elements that now seem to comprise the inventory of possible construction materials to find out which are actually important, and over what energy ranges they are important. A number of computer codes for calculating cross sections for neutron induced reactions have been evaluated and extended. These will be described and examples will be given of various types of calculations of interest to the CTR program. (U.S.)

  19. Measurement of gamma-ray production cross sections in neutron-induced reactions for Al and Pb

    International Nuclear Information System (INIS)

    Pavlik, A.; Vonach, H.; Hitzenberger, H.

    1995-01-01

    The prompt gamma-radiation from the interaction of fast neutrons with aluminum and lead was measured using the white neutron beam of the WNR facility at the Los Alamos National Laboratory. The samples (Al and isotopically enriched 207 Pb and 208 Pb) were positioned at about 20 m or 41 m distance from the neutron production target. The spectra of the emitted gamma-rays were measured with a high-resolution HPGe detector. The incident neutron energy was determined by the time-of-flight method and the neutron fluence was measured with a U fission chamber. From the aluminum gamma-ray spectra excitation functions for prominent gamma-transitions in various residual nuclei (in the range from O to Al) were derived for neutron energies from 3 MeV to 400 MeV. For lead (n,xnγ) reactions were studied for neutron energies up to 200 MeV by analyzing prominent gamma-transitions in the residual nuclei 200,202,204,206,207,208 Pb. The experimental results were compared with nuclear model calculations using the code GNASH. A good overall agreement was obtained without special parameter adjustments

  20. Actinide neutron-induced fission up to 20 MeV

    International Nuclear Information System (INIS)

    Maslov, V.M.

    2001-01-01

    Fission and total level densities modelling along with double-humped fission barrier parameters allow to describe available actinide neutron-induced fission cross section data in the incident neutron energy range of ∼ 10 keV - 20 MeV. Saddle asymmetries relevant to shell correction model calculations influence fission barriers, extracted by cross section data analysis. The inner barrier was assumed axially symmetric in case of U, Np and Pu neutron-deficient nuclei. It is shown that observed irregularities in neutron-induced fission cross section data in the energy range of 0.5-3 MeV could be attributed to the interplay of few-quasiparticle excitations in the level density of fissioning and residual nuclei. Estimates of first-chance fission cross section and secondary neutron spectrum model were validated by 238 U fission, (n,2n) and (n,3n) data description up to 20 MeV. (author)

  1. Actinide neutron-induced fission up to 20 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Maslov, V M [Radiation Physics and Chemistry Problems Institute, Minsk-Sosny (Belarus)

    2001-12-15

    Fission and total level densities modelling along with double-humped fission barrier parameters allow to describe available actinide neutron-induced fission cross section data in the incident neutron energy range of {approx} 10 keV - 20 MeV. Saddle asymmetries relevant to shell correction model calculations influence fission barriers, extracted by cross section data analysis. The inner barrier was assumed axially symmetric in case of U, Np and Pu neutron-deficient nuclei. It is shown that observed irregularities in neutron-induced fission cross section data in the energy range of 0.5-3 MeV could be attributed to the interplay of few-quasiparticle excitations in the level density of fissioning and residual nuclei. Estimates of first-chance fission cross section and secondary neutron spectrum model were validated by {sup 238}U fission, (n,2n) and (n,3n) data description up to 20 MeV. (author)

  2. PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo; Sugi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iijima, Shungo; Nishigori, Takeo

    1999-06-01

    The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,{gamma}), (n,n`), (n,p), (n,{alpha}), (n,d), (n,t), (n,{sup 3}He), (n,2n), (n,n`p), (n,n`{alpha}), (n,n`d), (n,n`t), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)

  3. Evaluation of neutron-induced reactions in 48Ti and 238U

    International Nuclear Information System (INIS)

    Carlson, B.V.; Fiorentino, J.; Frederico, T.; Isidro Filho, M.P.; Mastroleo, R.C.; Rego, R.A.

    1984-05-01

    Preliminary results of the evaluation of neutron-induced reactions in 48 Ti and 238 U are presented. Calculated cross sections for the reactions (n,γ), (n,n'), (n, 2n) and (n,p) as well as for (n,f) in 238 U are given. Comparisons with available experimental data are made and possible changes in the parameters are discussed. (Author) [pt

  4. Neutron induced electron radiography

    International Nuclear Information System (INIS)

    Andrade, Marcos Leandro Garcia

    2008-01-01

    In the present paper a new radiography technique, the 'Neutron Induced Electron Radiography' - NIER, to inspect low thickness samples on the order of micra, has been developed. This technique makes use of low energy electrons as penetrating radiation generated from metallic gadolinium screens when irradiated by thermal neutrons. The conditions to obtain the best image for the conventional X-ray film Kodak-AA were determined by using a digital system to quantify the darkening level of the film. The irradiations have been performed at a radiography equipment installed at the beam-hole no. 8 of the 5 MW IEA-R1 nuclear research reactor of IPEN-CNEN/SP. The irradiation time to obtain the best radiography was 100 seconds and for such condition the technique was able to discern 1 μm in 24 μm of aluminum at a resolution of 32 μm. By visual comparison the images obtained by the NIER shown a higher quality when compared with the ones from other usual techniques the make use of electrons a penetrating radiation and films for image registration. Furthermore the use of the digital system has provided a smaller time for data acquisition and data analysis as well as an improvement in the image visualization. (author)

  5. Measurement of neutron-induced fission cross-sections of Th232, U238, U233 and Np237 relative to U235 from 1 MeV to 200 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Shcherbakov, O.A.; Laptev, A.B.; Petrov, G.A. [Petersburg Nuclear Physics Inst., Gatchina, Leningrad district (Russian Federation); Fomichev, A.V.; Donets, A.Y.; Osetrov, O.I.

    1998-11-01

    The measurements of neutron-induced cross-section ratios for Th232, U238, U233 and Np237 relative to U235 have been carried out in the energy range from 1 MeV up to 200 MeV using the neutron time-of-flight spectrometer GNEIS based on 1 GeV proton synchrocyclotron. Below 20 MeV, the results of present measurements are roughly in agreement with evaluated data though there are some discrepances to be resolved. (author)

  6. The IAEA co-ordinated research programme on improvement of measurements, theoretical computations and evaluations of neutron induced helium production cross sections. Status report. Prepared at the final CRP meeting in Sendai, Japan 25-29 September 1995

    International Nuclear Information System (INIS)

    Pashchenko, A.B.

    1996-12-01

    The present report describes the results of the IAEA Co-ordinated Research Programme (CRP) on ''Improvements of Measurements, Theoretical Computation and Evaluations of Neutron Induced Helium Production Cross Sections''. Summarized is the progress achieved under the CRP in the following areas: measurements of α-production cross sections for structural materials, theoretical computations at (nα) cross sections; measurements of activation cross sections; and improvement of experimental methods for (n,α) investigations. The status report gives also short summaries on the work of each laboratory which contributed to the results of the CRP. Attached is the list of program members and participants of CRP meetings. (author). Refs, 2 figs, 1 tab

  7. Calculation of the resonance cross section functions

    International Nuclear Information System (INIS)

    Slipicevic, K.F.

    1967-11-01

    This paper includes the procedure for calculating the Doppler broadened line shape functions ψ and χ which are needed for calculation of resonance cross section functions. The obtained values are given in tables

  8. Calculation of the resonance cross section functions

    Energy Technology Data Exchange (ETDEWEB)

    Slipicevic, K F [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-11-15

    This paper includes the procedure for calculating the Doppler broadened line shape functions {psi} and {chi} which are needed for calculation of resonance cross section functions. The obtained values are given in tables.

  9. Systematics of neutron-induced fission yields

    International Nuclear Information System (INIS)

    Blachot, J.; Brissot, R.

    1983-10-01

    The main characteristics of the mass and charge distributions for thermal neutron induced fission of actinides are reviewed. We show that these distributions can be reasonably reproduced with only 24 data as input. We use a representation where the element yields together with the most probable mass Ap(Z) play the dominant role. The ability of this model to calculate mass yields for the fission of not yet measured actinides is also shown. The influence of the excitation energy of the fissile system on charge and mass distribution is also discussed

  10. Fast-neutron-induced fission of 242Pu at nELBE

    Directory of Open Access Journals (Sweden)

    Kögler Toni

    2017-01-01

    Full Text Available The fast neutron-induced fission cross section of 242Pu was determined in the range of 0.5 MeV to 10 MeV relative to 235U(n,f at the neutron time-of-flight facility nELBE. The number of target nuclei was calculated by means of measuring the spontaneous fission rate of 242Pu. Neutron transport simulations with Geant4 and MCNP6 are used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

  11. Neutron-induced fission cross-section of 233U, 241Am and 243Am in the energy range 0.5 MeV ≤ En ≤ 20 MeV

    International Nuclear Information System (INIS)

    Belloni, F.; Milazzo, P.M.; Calviani, M.

    2011-01-01

    Neutron-induced fission cross-sections of 233 U, 241 Am and 243 Am relative to 235 U have been measured in a wide energy range at the neutron time of flight facility n-TOF in Geneva to address the present discrepancies in evaluated and experimental databases for reactions and isotopes relevant for transmutation and new generation fast reactors. A dedicated fast ionization chamber was used. Each isotope was mounted in a different cell of the modular detector. The measurements took advantage of the characteristics of the n-TOF installation. Its intrinsically low background, coupled to its high instantaneous neutron flux, results in high accuracy data. Its wide energy neutron spectrum helps to reduce systematic uncertainties due to energy-domain matching problems while the 185 m flight path and a 6 ns pulse width assure an excellent energy resolution. This paper presents results obtained between 500 keV and 20 MeV neutron energy. (authors)

  12. Integrated system for production of neutronics and photonics calculational constants. Volume 15, Part C. The LLL Evaluated Nuclear Data Library (ENDL): translation of ENDL neutron-induced interaction data into the ENDF/B format

    International Nuclear Information System (INIS)

    Howerton, R.J.

    1976-01-01

    The LLL evaluated nuclear data library (ENDL) has been translated into the evaluated neutron data file/version B (ENDF/B) format. This translation is for the convenience of those who wish to use ENDL data but who are more familiar with ENDF/B formats and procedures. Only that portion of ENDL dealing with neutron-induced interactions (including photon production from neutron-induced reactions) has been translated

  13. Light charged particle production in fast neutron-induced reactions on carbon (En=40 to 75 MeV) (II). Tritons and alpha particles

    International Nuclear Information System (INIS)

    Dufauquez, C.; Slypen, I.; Benck, S.; Meulders, J.P.; Corcalciuc, V.

    2000-01-01

    Double-differential cross sections for fast neutron-induced triton and alpha-particle production on carbon are reported at six incident neutron energies between 40 and 75 MeV. Angular distributions were measured at laboratory angles between 20 deg. and 160 deg. . Energy-differential, angle-differential and total cross sections are also reported. Experimental cross sections are compared to existing experimental data and to theoretical model calculations

  14. Model cross section calculations using LAHET

    International Nuclear Information System (INIS)

    Prael, R.E.

    1992-01-01

    The current status of LAHET is discussed. The effect of a multistage preequilibrium exciton model following the INC is examined for neutron emission benchmark calculations, as is the use of a Fermi breakup model for light nuclei rather than an evaporation model. Comparisons are made also for recent fission cross section experiments, and a discussion of helium production cross sections is presented

  15. Statistical Verification and Validation of the EXFOR database: (n,n'), (n,2n), (n,p), (n,α) and other neutron-induced threshold reaction cross-sections

    International Nuclear Information System (INIS)

    Koning, Arjan

    2014-01-01

    The NEA Data Bank operates as an international centre of reference for its member countries with respect to basic nuclear tools, such as computer codes and nuclear data, used for the analysis and prediction of phenomena in the nuclear field. It provides a direct service to its users by acquiring, developing, improving and validating these tools and making them available as requested. In the continuity of WPEC Subgroup 30 work on improving the accessibility and quality of the EXFOR database, the NEA Data Bank initiated a number of activities to further improve and validate its nuclear databases. In particular, it was proposed to perform a comprehensive review of cross-section data in the EXFOR database. This report describes the development of an efficient review system and its application to more than 10,000 cross-section data sets of neutron-induced threshold and activation reactions. The current report is set up as follows. In Section 2 it is outlined how the original EXFOR data collection is transformed into a database that can be subjected to a statistical analysis. In Section 3, a classification for scoring of EXFOR data sets is proposed. Next, in Section 4 the goodness-of-fit estimators that lead to the scoring are defined. Finally a graphical comparison is presented of all experimental data, together with the available major nuclear data libraries, covered in this paper. (author)

  16. Neutron induced activity in fuel element components

    International Nuclear Information System (INIS)

    Kjellbert, N.

    1978-03-01

    A thorough investigation of the importance of various nuclides in neutron-induced radioactivity from fuel element construction materials has been carried out for both BWR and PWR fuel assemblies. The calculations were performed with the ORIGEN computer code. The investigation was directed towards the final storage of the assembly components and special emphasis was put to the examination of the sources of carbon-14, cobalt-60, nickel-59, nickel-63 and zirconium-93/niobium-93m. It is demonstrated that the nuclides nickel-59, in Inconel and stainless steel, and zirconium-93/niobium-93m, in Zircaloy, are the ones which constitute the very long term radiotoxic hazard of the irradiated materials. (author)

  17. Methods for calculating anisotropic transfer cross sections

    International Nuclear Information System (INIS)

    Cai, Shaohui; Zhang, Yixin.

    1985-01-01

    The Legendre moments of the group transfer cross section, which are widely used in the numerical solution of the transport calculation can be efficiently and accurately constructed from low-order (K = 1--2) successive partial range moments. This is convenient for the generation of group constants. In addition, a technique to obtain group-angle correlation transfer cross section without Legendre expansion is presented. (author)

  18. Measurement of the neutron-induced fission cross section of 237Np relative to 235U from 0.02 to 30 MeV

    International Nuclear Information System (INIS)

    Behrens, J.W.; Magana, J.W.; Browne, J.C.

    1977-01-01

    The 237 Np/ 235 U fission cross section ratio has been measured from 0.02 to 30 MeV. Using the threshold method, a value of 1.294 +- 0.019 is obtained for the average cross section ratio in the interval from 1.75 to 4.00 MeV

  19. Measurement of the neutron-induced fission cross section of 232Th relative to 235U from 0.7 to 30 MeV

    International Nuclear Information System (INIS)

    Behzens, T.W.; Ables, E.; Browne, T.C.

    1982-01-01

    The authors have measured the fission cross-section ratio 232 Th: 235 U as a function of neutron energy from 0.7 to 30 MeV using ionization fission chambers, the threshold cross-section method, and the time-of-flight technique at the Lawrence Livermore National Laboratory 100-MeV electron linear accelerator. The measured cross-section ratio, averaged over the neutron energy interval from 1.75 to 4.00 MeV, was 0.1086 + 0.0024

  20. Fission cross-section calculations and the multi-modal fission model

    International Nuclear Information System (INIS)

    Hambsch, F.J.

    2004-01-01

    New, self consistent, neutron-induced reaction cross section calculations for 235,238 U, 237 Np have been performed. The statistical model code STATIS was improved to take into account the multimodality of the fission process. The three most dominant fission modes, the two asymmetric standards I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode have been taken into account. De-convoluted fission cross sections for those modes for 235,238 U(n,f) and 237 Np(n,f) based on experimental branching ratios, were calculated for the first time up to the second chance fission threshold. For 235 U(n,f), the calculations being made up to 28 MeV incident neutron energy, higher fission chances have been considered. This implied the need for additional calculations for the neighbouring isotopes. As a side product also mass yield distributions could be calculated at energies hitherto not accessible by experiment. Experimental validation of the predictions is being envisaged

  1. Integral cross-section measurements for investigating the emission of complex particles in 14 MeV neutron-induced nuclear reactions

    International Nuclear Information System (INIS)

    Qaim, S.M.

    1981-01-01

    Some of the off-line techniques used for the determination of integral cross-section data are reviewed and, as a critical check, some typical data sets are compared. The systematic trends reported in the cross-section data for (n,d), (n,t), (n, 3 He) and (n,α) reactions are discussed. A brief discussion of the possible reaction mechanisms is given. Some of the applications of the data are outlined. (author)

  2. Neutron-induced fission cross-section measurement of 234U with quasi-monoenergetic beams in the keV and MeV range using micromegas detectors

    Science.gov (United States)

    Tsinganis, A.; Kokkoris, M.; Vlastou, R.; Kalamara, A.; Stamatopoulos, A.; Kanellakopoulos, A.; Lagoyannis, A.; Axiotis, M.

    2017-09-01

    Accurate data on neutron-induced fission cross-sections of actinides are essential for the design of advanced nuclear reactors based either on fast neutron spectra or alternative fuel cycles, as well as for the reduction of safety margins of existing and future conventional facilities. The fission cross-section of 234U was measured at incident neutron energies of 560 and 660 keV and 7.5 MeV with a setup based on `microbulk' Micromegas detectors and the same samples previously used for the measurement performed at the CERN n_TOF facility (Karadimos et al., 2014). The 235U fission cross-section was used as reference. The (quasi-)monoenergetic neutron beams were produced via the 7Li(p,n) and the 2H(d,n) reactions at the neutron beam facility of the Institute of Nuclear and Particle Physics at the `Demokritos' National Centre for Scientific Research. A detailed study of the neutron spectra produced in the targets and intercepted by the samples was performed coupling the NeuSDesc and MCNPX codes, taking into account the energy spread, energy loss and angular straggling of the beam ions in the target assemblies, as well as contributions from competing reactions and neutron scattering in the experimental setup. Auxiliary Monte-Carlo simulations were performed with the FLUKA code to study the behaviour of the detectors, focusing particularly on the reproduction of the pulse height spectra of α-particles and fission fragments (using distributions produced with the GEF code) for the evaluation of the detector efficiency. An overview of the developed methodology and preliminary results are presented.

  3. Calculation of 235U(n,n') cross sections for ENDF/B-VI

    International Nuclear Information System (INIS)

    Young, P.G.; Arthur, E.D.

    1988-01-01

    Cross sections for neutron-induced reactions on 235 U between 0.01 and 20 MeV have been calculated in a preliminary analysis for the ENDF/B-VI evaluation with particular emphasis on neutron inelastic scattering. A deformed optical model potential that fits total, elastic, inelastic, and low-energy average resonance data is used to calculate direct (n,n') cross sections and transmission coefficients for a Hauser-Feshbach statistical theory analysis using a multiple fission barrier representation. Direct cross sections for higher-lying vibrational states are provided from DWBA calculations, normalized using B(E/ital l/) values determined from (d,d') and Coulomb excitation data. Initial fission barrier parameters and transition state density enhancements appropriate to the compound systems involved were obtained from previous analyses, especially fits to charged-particle fission probability data. Further modifications to fit 235 U(n,f) data were small, and the final fission parameters are generally consistent with published values. The results from this preliminary analysis are compared with the ENDF/B-V evaluation as well as with experimental data. 26 refs., 5 figs., 3 tabs

  4. Neutron induced 238U subthreshold fission cross section for neutron energies between 5 eV and 3.5 MeV

    International Nuclear Information System (INIS)

    Perez, R.B.; Difilippo, F.C.; Saussure, G. de; Ingle, R.W.

    1978-01-01

    A measurement of the 238 U fission cross section between 5 eV and 3.5 MeV was performed. Included is the identification of 85 resonances or clusters of resonances below 200 keV. Also the fission widths for the 27 resolved class I levels were computed from their fission areas, and a neutron width of 0.005 MeV was estimated for the quasi-class II level in the 721 eV fission cluster. The fission level spacing and cross sections are discussed. 9 references

  5. Ab initio calculations of cross luminescence materials

    International Nuclear Information System (INIS)

    Kanchana, V.

    2016-01-01

    Abintio calculations have been performed to study the structural, electronic, and optical properties of ABX 3 (A=alkali, B=alkaline-earth, and X=halide) compounds. The ground state properties are calculated using the pseudopotential method with the inclusion of van der Waals interaction, which we find inevitable in reproducing the experimental structure properties in alkali iodides because of its layered structure. All calculations were performed using the Full-Potential Linearized Augmented Plane Wave method. The band structures are plotted with various functionals and we find the newly developed Tran and Blaha modified Becke-Johnson potential to improve the band gap significantly. The optical properties such as complex dielectric function, refractive index, and absorption spectra are calculated which clearly reveal the optically isotropic nature of these materials though being structurally anisotropic, which is the key requirement for ceramic scintillators. Cross luminescence materials are very interesting because of its fast decay. One of the major criteria for the cross luminescence to happen is the energy difference between valence band and next deeper core valence band being lesser when compared to energy gap of the compound, so that radiative electronic transition may occur between valence band and core valence band. We found this criteria to be satisfied in all the studied compounds leading to cross luminescence except for KSrI 3 , RbSrI 3 . The present study suggest that among the six compounds studied, CsSrI 3 , CsMgCl 3 , CsCaCl 3 , and CsSrCl 3 compounds are cross luminescence materials, which is well explained from the band structure, optical properties calculations. Chlorides are better scintillators that iodides and CsMgCl 3 is found to be promising one among the studied compounds. Apart from these materials we have also discussed electronic structure and optical properties of other scintillator compounds. (author)

  6. Recent cross-section measurements of neutron-induced reactions of importance for background estimates in 0νββ searches

    Science.gov (United States)

    Tornow, W.; Bhike, M.; Finch, S. W.; Krishichayan

    2017-09-01

    We report on cross-section measurements for the reactions 76Ge(n,2n)75Ge, 76Ge(n,n'γ)76Ge, 126,127,128Te(n,γ)127,129,131Te, and 136Xe(n,n'γ)136Xe in the neutron energy range between 0.5 MeV and 15 MeV.

  7. Recent cross-section measurements of neutron-induced reactions of importance for background estimates in 0νββ searches

    Directory of Open Access Journals (Sweden)

    Tornow W.

    2017-01-01

    Full Text Available We report on cross-section measurements for the reactions 76Ge(n,2n75Ge, 76Ge(n,n′γ76Ge, 126,127,128Te(n,γ127,129,131Te, and 136Xe(n,n′γ136Xe in the neutron energy range between 0.5 MeV and 15 MeV.

  8. Neutron-Induced Charged Particle Studies at LANSCE

    Science.gov (United States)

    Lee, Hye Young; Haight, Robert C.

    2014-09-01

    Direct measurements on neutron-induced charged particle reactions are of interest for nuclear astrophysics and applied nuclear energy. LANSCE (Los Alamos Neutron Science Center) produces neutrons in energy of thermal to several hundreds MeV. There has been an effort at LANSCE to upgrade neutron-induced charged particle detection technique, which follows on (n,z) measurements made previously here and will have improved capabilities including larger solid angles, higher efficiency, and better signal to background ratios. For studying cross sections of low-energy neutron induced alpha reactions, Frisch-gridded ionization chamber is designed with segmented anodes for improving signal-to-noise ratio near reaction thresholds. Since double-differential cross sections on (n,p) and (n,a) reactions up to tens of MeV provide important information on deducing nuclear level density, the ionization chamber will be coupled with silicon strip detectors (DSSD) in order to stop energetic charged particles. In this paper, we will present the status of this development including the progress on detector design, calibrations and Monte Carlo simulations. This work is funded by the US Department of Energy - Los Alamos National Security, LLC under Contract DE-AC52-06NA25396.

  9. Proton and deuteron production in neutron-induced reactions on carbon at En=42.5, 62.7, and 72.8 MeV

    International Nuclear Information System (INIS)

    Slypen, I.; Corcalciuc, V.; Meulders, J.P.

    1995-01-01

    Double-differential cross sections for proton and deuteron production in fast neutron induced reactions on carbon are reported for three incident neutron energies: 42.5, 62.7, and 72.8 MeV. Angular distributions were measured at laboratory angles between 20 degree and 160 degree. Procedures for data taking and data reduction are presented. Energy-differential cross sections and total cross sections are also reported. Experimental cross sections are compared with existing data and with theoretical calculations in the frame of the intranuclear cascade model

  10. 232Th and 238U neutron emission cross section calculations and analysis of experimental data

    International Nuclear Information System (INIS)

    Tel, E.

    2004-01-01

    In this study, pre-equilibrium neutron-emission spectra produced by (n,xn) reactions on nuclei 2 32Th and 2 38U have been calculated. Angle-integrated cross sections in neutron induced reactions on targets 2 32Th and 2 38U have been calculated at the bombarding energies up to 18 MeV. We have investigated multiple pre-equilibrium matrix element constant from internal transition for 2 32Th (n,xn) neutron emission spectra. In the calculations, the geometry dependent hybrid model and the cascade exciton model including the effects of pre-equilibrium have been used. In addition, we have described how multiple pre-equilibrium emissions can be included in the Feshbach-Kerman-Koonin (FKK) fully quantum-mechanical theory. By analyzing (n,xn) reaction on 232 T h and 2 38U, with the incident energy from 2 Me V to 18 Me V, the importance of multiple pre-equilibrium emission can be seen cleady. All calculated results have been compared with experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other

  11. Angular distribution of fragments from neutron-induced fission of 238U in the intermediate energy region

    International Nuclear Information System (INIS)

    Carlsson, Magnus

    2004-06-01

    Areas ranging from nuclear structure models to accelerator-driven systems benefit from improved neutron-induced fission data in the intermediate energy region. In this Master's degree thesis, the fragment angular distribution from fission of 238 U, induced by 21-MeV neutrons, has been analysed from an experiment performed with the Medley/DIFFICILE setup at the The Svedberg Laboratory in Uppsala. The data have been corrected for low energy neutrons in the beam. The results agree with other experiments, as well as with model calculations. The data should be a starting point for further analysis with a goal to deduce the fission cross-section of 238 U

  12. Neutron-induced activation measurements and EXFOR compilations in the energy range up to 20 MeV

    International Nuclear Information System (INIS)

    Semkova, V.; Otuka, N.

    2015-01-01

    Accurate neutron-induced activation cross-section data are of interest in many fields of science and applications. Such data are needed for calculations and analysis of neutron transport, activation of materials, gas production and radiation damage, dose rates etc. Experimental data provide bases for the parameterization of reaction cross section calculations, and for the assessment of nuclear models and evaluated data libraries. Activation technique in combination with gamma spectrometry is well known and widely used method for neutron-induced reaction cross-section measurements. However, in some cases considerable differences exist between the results from different experiments. A careful consideration of the all factors that may affect each particular measurement is needed in order to obtain reliable data. Measured data are of little value until they are made conveniently available for users and evaluators. The International Network of Nuclear Reaction Data Centres (NRDC) collaborates in collection, compilation and dissemination of experimental nuclear reaction data in the EXFOR data library. In the present work some aspects of the 58 Ni(n,p) 58 Co activation cross-section measurements at two different experimental facilities and EXFOR compilation files will be presented

  13. Detection of boron in metal alloys with solid state nuclear track detector by neutron induced autoradiography

    International Nuclear Information System (INIS)

    Ali Nabipour; Hosseini, A.; Afarideh, H.

    2002-01-01

    Neutron induced autoradiography is very useful technique for detection as well as measurement of Boron densities in metal alloys. The method is relatively simple and quite sensitive in comparison with other techniques with resolution in the range of PPM. Using this technique with it is also possible to investigate microscopic scattering of Boron in metal alloys. In comparison with most techniques neutron induced autoradiography has its own difficulties and limitations. In this research measurement of Boron densities and investigation of that diffusion in metal alloys has been carried out. A flat nicely polished Boron doped metal samples is covered with a track detecting plastic (CR-39 solid state nuclear track detector) and exposed to thermal neutron dose. After irradiation the plastic detector have been removed and put in an etching solution. Since the diffusion rate of corrosive solution in those area, which heavy ions have been, produces as the result of nuclear reaction with thermal neutron are more than the other areas, some cavities are formed. The diameter of cavities or tracks cross section are increased with increasing the etching time, to some extent that it is possible to observe the cavities with optical microscopes. The density of tracks on the detector surface is directly related to the Boron concentration in the sample and thermal neutron dose. So by measuring the number of tracks on surface of the detector it would possible to calculate the concentration of Boron in metal samples. (Author)

  14. New resonance cross section calculational algorithms

    International Nuclear Information System (INIS)

    Mathews, D.R.

    1978-01-01

    Improved resonance cross section calculational algorithms were developed and tested for inclusion in a fast reactor version of the MICROX code. The resonance energy portion of the MICROX code solves the neutron slowing-down equations for a two-region lattice cell on a very detailed energy grid (about 14,500 energies). In the MICROX algorithms, the exact P 0 elastic scattering kernels are replaced by synthetic (approximate) elastic scattering kernels which permit the use of an efficient and numerically stable recursion relation solution of the slowing-down equation. In the work described here, the MICROX algorithms were modified as follows: an additional delta function term was included in the P 0 synthetic scattering kernel. The additional delta function term allows one more moments of the exact elastic scattering kernel to be preserved without much extra computational effort. With the improved synthetic scattering kernel, the flux returns more closely to the exact flux below a resonance than with the original MICROX kernel. The slowing-down calculation was extended to a true B 1 hyperfine energy grid calculatn in each region by using P 1 synthetic scattering kernels and tranport-corrected P 0 collision probabilities to couple the two regions. 1 figure, 6 tables

  15. distributions for the thermal neutron induced fission of 234U

    Directory of Open Access Journals (Sweden)

    Al-Adili A.

    2016-01-01

    In addition, the analysis of thermal neutron induced fission of 234U(n,f will be discussed. Currently analysis of data is ongoing, originally taken at the ILL reactor. The experiment is of particular interest since no measurement exist of the mass and energy distributions for this system at thermal energies. One main problem encountered during analysis was the huge background of 235U(nth,f. Despite the negligible isotopic traces in the sample, the cross section difference is enormous. Solution to this parasitic background will be highlighted.

  16. Calculations and Evaluations of Cross Sections for n + 204,206,207,208,natPb Reactions in the En ≤ 250 MeV Energy Range

    International Nuclear Information System (INIS)

    Han Yinlu; Shen Qingbiao; Zhang Zhengjun; Cai Chonghai

    2005-01-01

    The quality and reliability of the computational simulation of a macroscopic nuclear device are directly related to the quality of the underlying basic nuclear data. To meet these needs, according to advanced nuclear models that account for details of nuclear structure and the quantum nature of nuclear reaction and the experimental data of total, nonelastic, and elastic scattering cross sections, and elastic scattering angular distributions of Pb and its isotopes, all cross sections of neutron-induced reaction, angular distributions, energy spectra, especially the double-differential cross sections for neutron, proton, deuteron, triton, helium, and alpha emissions are calculated and analyzed for n + 204,206,207,208,nat Pb at incident neutron energies below 20 MeV by using the UNF nuclear model code. At neutron incident energies 20 n ≤ 250 MeV, MEND codes are used. Theoretical calculations are compared with existing experimental data and other evaluated data from ENDF/B-VI and JENDL-3

  17. Parity violation in neutron induced reactions

    International Nuclear Information System (INIS)

    Gudkov, V.P.

    1991-06-01

    The theory of parity violation in neutron induced reactions is discussed. Special attention is paid to the energy dependence and enhancement factors for the various types of nuclear reactions and the information which might be obtained from P-violating effects in nuclei. (author)

  18. Neutron-induced fission of uranium isotopes up to 100 MeV

    International Nuclear Information System (INIS)

    Lestone, J.P.; Gavron, A.

    1994-01-01

    The statistical-model description of the neutron-induced fission of U isotopes has been developed using densities of intrinsic states and spin cutoff parameters obtained directly from appropriate Nilsson model single-particle levels. The first-chance fission cross sections are reproduced well when the rotational contributions to the nuclear level densities are taken into account. In order to fit the U(n,f) cross sections above the threshold of second-chance fission, we must: (1) assume that the triaxial level-density enhancement is washed out at an excitation energy of approximately 7 MeV above the triaxial barriers with a width of approximately 1 MeV, implying a γ deformation for the first barriers where 10<γ<20 degree, and (2) include preequilibrium particle emission in the calculations. Above an incoming-neutron kinetic energy of approximately 17 MeV, our statistical model U(n,f) of cross sections increasingly overestimates the experimental data. This is not surprising since, at these high energies, little data exist on the scattering of neutrons to help guide the choice of optical-model parameters. A satisfactory reproduction of all of the available U(n,f) cross sections above 17 MeV is obtained by scaling our calculated compound-nucleus formation cross sections. This scaling factor falls from 1.0 at 17 MeV to 0.82 at 100 MeV

  19. Development of the multistep compound process calculation code

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan)

    1998-03-01

    A program `cmc` has been developed to calculate the multistep compound (MSC) process by Feshback-Kerman-Koonin. A radial overlap integral in the transition matrix element is calculated microscopically, and comparisons are made for neutron induced {sup 93}Nb reactions. Strengths of the two-body interaction V{sub 0} are estimated from the total MSC cross sections. (author)

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

    Science.gov (United States)

    Pritychenko, B.; Mughabghab, S. F.

    2012-12-01

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

  1. Neutron induced permanent damage in Josephson junctions

    International Nuclear Information System (INIS)

    Mueller, G.P.; Rosen, M.

    1982-01-01

    14 MeV neutron induced permanent changes in the critical current density of Josephson junctions due to displacement damage in the junction barrier are estimated using a worst case model and the binary collision simulation code MARLOWE. No likelihood of single event hard upsets is found in this model. It is estimated that a fluence of 10 18 -10 19 neutrons/cm 2 are required to change the critical current density by 5%

  2. Neutron induced bystander effect among zebrafish embryos

    Science.gov (United States)

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

    2015-12-01

    The present paper reported the first-ever observation of neutron induced bystander effect (NIBE) using zebrafish (Danio rerio) embryos as the in vivo model. The neutron exposure in the present work was provided by the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility at the National Institute of Radiological Sciences (NIRS), Chiba, Japan. Two different strategies were employed to induce NIBE, namely, through directly partnering and through medium transfer. Both results agreed with a neutron-dose window (20-50 mGy) which could induce NIBE. The lower dose limit corresponded to the threshold amount of neutron-induced damages to trigger significant bystander signals, while the upper limit corresponded to the onset of gamma-ray hormesis which could mitigate the neutron-induced damages and thereby suppress the bystander signals. Failures to observe NIBE in previous studies were due to using neutron doses outside the dose-window. Strategies to enhance the chance of observing NIBE included (1) use of a mono-energetic high-energy (e.g., between 100 keV and 2 MeV) neutron source, and (2) use of a neutron source with a small gamma-ray contamination. It appeared that the NASBEE facility used in the present study fulfilled both conditions, and was thus ideal for triggering NIBE.

  3. Practical calculations of quantum breakup cross sections

    International Nuclear Information System (INIS)

    McCurdy, C. W.; Rescigno, T. N.

    2000-01-01

    The Schroedinger equation is solved numerically using the method of exterior complex scaling for several models of the breakup of an atom by electron impact. Using the accurate wave functions thereby obtained for these model problems, several well-known integral expressions for quantum-mechanical breakup amplitudes are tested. It is shown that some formally correct integral expressions for the breakup amplitudes can yield numerically unstable or poorly convergent results. Calculations are presented for a case with simple exponential potentials and a case in which a metastable state of the target, analogous to an autoionizing state, can decay into the breakup channel. For cases involving only short-range (non-Coulomb) interactions, alternative expressions can be found that are stable in calculations of practical scale. (c) 2000 The American Physical Society

  4. Discussion of electron cross sections for transport calculations

    International Nuclear Information System (INIS)

    Berger, M.J.

    1983-01-01

    This paper deals with selected aspects of the cross sections needed as input for transport calculations and for the modeling of radiation effects in biological materials. Attention is centered mainly on the cross sections for inelastic interactions between electrons and water molecules and the use of these cross sections for the calculation of energy degradation spectra and of ionization and excitation yields. 40 references, 3 figures, 1 table

  5. Neutron induced fission of U isotopes up to 100 MeV

    International Nuclear Information System (INIS)

    Lestone, J.P.; Gavron, A.

    1993-01-01

    We have developed a statistical model description of the neutron induced fission of U isotopes using densities of intrinsic states and spin cut off parameters obtained directly from appropriate Nilsson model single particle levels. The first chance fission cross sections are well reproduced when the rotational contributions to the nuclear level densities are taken into account. In order to fit the U(n,f) cross sections above the threshold of second chance fission, we need to: (1) assume that the triaxial level density enhancement is washed out at an excitation energy of ∼7 MeV above the triaxial barriers with a width of ∼1 MeV, implying a γ deformation for the first barriers of 10 degree < γ < 20 degree; and (2) include pre-equilibrium particle emission in the calculations. Above an incoming neutron kinetic energy of ∼17 MeV our statistical model U(n,f) cross sections increasingly overestimate the experimental data when so called ''good'' optical model potentials are used to calculate the compound nucleus formation cross sections. This is not surprising since at these high energies little data exists on the scattering of neutrons to help guide the choice of optical model parameters. A satisfactory reproduction of all the available U(n,f) cross sections above 17 MeV is obtained by a simple scaling of our calculated compound nucleus formation cross sections. This scaling factor falls from 1.0 at 17 MeV to 0.82 at 100 MeV

  6. Neutron-induced helium implantation in GCFR cladding

    International Nuclear Information System (INIS)

    Yamada, H.; Poeppel, R.B.; Sevy, R.H.

    1980-10-01

    The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10 10 He/cm 2 .s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 μm, more than 99% of helium particles are implanted in the first 2-μm-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding

  7. Development of a Nuclear Reaction Database on Silicon for Simulation of Neutron-Induced Single-Event Upsets in Microelectronics and its Application

    International Nuclear Information System (INIS)

    Watanabe, Yukinobu; Kodama, Akihiro; Tukamoto, Yasuyuki; Nakashima, Hideki

    2005-01-01

    We have developed a cross-section database for neutron-induced reactions on 28Si in the energy range between 2 MeV and 3 GeV in order to analyze single-event upsets (SEUs) phenomena induced by cosmic-ray neutrons in microelectronic devices. A simplified spherical device model is proposed for simulation of the initial processes of SEUs. The model is applied to SEU cross-section calculations for semiconductor memory devices. The calculated results are compared with measured SEU cross sections and the other simulation result. The dependence of SEU cross sections on incident neutron energy and secondary ions having the most important effects on SEUs are discussed

  8. Use of nuclear reaction models in cross section calculations

    International Nuclear Information System (INIS)

    Grimes, S.M.

    1975-03-01

    The design of fusion reactors will require information about a large number of neutron cross sections in the MeV region. Because of the obvious experimental difficulties, it is probable that not all of the cross sections of interest will be measured. Current direct and pre-equilibrium models can be used to calculate non-statistical contributions to neutron cross sections from information available from charged particle reaction studies; these are added to the calculated statistical contribution. Estimates of the reliability of such calculations can be derived from comparisons with the available data. (3 tables, 12 figures) (U.S.)

  9. Evaluation of fusion-evaporation cross-section calculations

    Science.gov (United States)

    Blank, B.; Canchel, G.; Seis, F.; Delahaye, P.

    2018-02-01

    Calculated fusion-evaporation cross sections from five different codes are compared to experimental data. The present comparison extents over a large range of nuclei and isotopic chains to investigate the evolution of experimental and calculated cross sections. All models more or less overestimate the experimental cross sections. We found reasonable agreement by using the geometrical average of the five model calculations and dividing the average by a factor of 11.2. More refined analyses are made for example for the 100Sn region.

  10. Partial Cross Sections of Neutron-Induced Reactions on nCu at En = 6, 8, 10, 12, 14, and 16 MeV for 0νββ Background Studies

    Science.gov (United States)

    Gooden, M. E.; Fallin, B. A.; Finch, S. W.; Kelley, J. H.; Howell, C. R.; Rusev, G.; Tonchev, A. P.; Tornow, W.; Stanislav, V.

    2014-05-01

    Partial cross-section measurements of (n,n'γ) reactions on natCu were carried out at TUNL using monoenergetic neutrons at six energies of En = 6, 8, 10, 12, 14, 16 MeV. These studies were performed to provide accurate cross-section data on materials abundant in experimental setups involving HPGe detectors used to search for rare events, like the neutrino-less double-beta decay of 76Ge. Spallation and (α,n) neutrons are expected to cause the largest source of external background in the energy region of interest. At TUNL pulsed neutron beams were produced via the 2H(d,n)3He reaction and the deexcitation γ rays from the reaction natCu(n,xγ) were detected with clover HPGe detectors. Cross-section results for the strongest transtions in 63Cu and 65Cu will be reported, and will compared to model calculations and to data recently obtained at LANL with a white neutron beam.

  11. Neutron induced alpha production from carbon between 18 and 22 MeV

    International Nuclear Information System (INIS)

    Stevens, A.P.

    1976-10-01

    Cross sections for neutron induced alpha production in carbon were measured at seventeen energies between 18 and 22 MeV, using a deuterated anthracene crystal as both target and detector. Pulse shape discrimination was employed to separate the alphas and elastically scattered deuterons from the other reaction products. Published (n,d) elastic scattering data were used as a standard to obtain the alpha production cross sections. Comparison with available measurements shows good agreement

  12. An Ada environment for relativistic cross section calculations

    International Nuclear Information System (INIS)

    Nilsson, E.

    1990-01-01

    We have developed an Ada environment adapted to relativistic cross section calculations. Objects such as four-vectors, γ- matrices and propagators are defined as well as operations between these objects. In this environment matrix elements can be expressed in a compact and readable way as Ada code. Unpolarized cross sections are calculated numerically by explicitly summing and averaging over spins and polarizations. A short presentation of the technique is given

  13. Calculation of atom displacement cross section for structure material

    International Nuclear Information System (INIS)

    Liu Ping; Xu Yiping

    2015-01-01

    The neutron radiation damage in material is an important consideration of the reactor design. The radiation damage of materials mainly comes from atom displacements of crystal structure materials. The reaction cross sections of charged particles, cross sections of displacements per atom (DPA) and KERMA are the basis of radiation damage calculation. In order to study the differences of DPA cross sections with different codes and different evaluated nuclear data libraries, the DPA cross sections for structure materials were calculated with UNF and NJOY codes, and the comparisons of results were given. The DPA cross sections from different evaluated nuclear data libraries were compared. And the comparison of DPA cross sections between NJOY and Monte Carlo codes was also done. The results show that the differences among these evaluated nuclear data libraries exist. (authors)

  14. Parametric equations for calculation of macroscopic cross sections

    International Nuclear Information System (INIS)

    Botelho, Mario Hugo; Carvalho, Fernando

    2015-01-01

    Neutronic calculations of the core of a nuclear reactor is one thing necessary and important for the design and management of a nuclear reactor in order to prevent accidents and control the reactor efficiently as possible. To perform these calculations a library of nuclear data, including cross sections is required. Currently, to obtain a cross section computer codes are used, which require a large amount of processing time and computer memory. This paper proposes the calculation of macroscopic cross section through the development of parametric equations. The paper illustrates the proposal for the case of macroscopic cross sections of absorption (Σa), which was chosen due to its greater complexity among other cross sections. Parametric equations created enable, quick and dynamic way, the determination of absorption cross sections, enabling the use of them in calculations of reactors. The results show efficient when compared with the absorption cross sections obtained by the ALPHA 8.8.1 code. The differences between the cross sections are less than 2% for group 2 and less than 0.60% for group 1. (author)

  15. Angular distribution of fragments from neutron-induced fission of {sup 238}U in the intermediate energy region

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Magnus

    2004-06-01

    Areas ranging from nuclear structure models to accelerator-driven systems benefit from improved neutron-induced fission data in the intermediate energy region. In this Master's degree thesis, the fragment angular distribution from fission of {sup 238}U, induced by 21-MeV neutrons, has been analysed from an experiment performed with the Medley/DIFFICILE setup at the The Svedberg Laboratory in Uppsala. The data have been corrected for low energy neutrons in the beam. The results agree with other experiments, as well as with model calculations. The data should be a starting point for further analysis with a goal to deduce the fission cross-section of {sup 238}U.

  16. Tritium production in neutron induced reactions

    International Nuclear Information System (INIS)

    Krasa, A.; Andreotti, E.; Hult, M.; Marissens, G.; Plompen, A.; Angelone, M.; Pillon, M.

    2011-01-01

    We present an overview of the present knowledge of (n,t) reaction excitation functions in the 14-21 MeV energy range for Cd, Cr, Fe, Mg, Mo, Ni, Pb, Pd, Ru, Sn, Ti, Zr. Experimental data are compared with evaluated data libraries, cross-section systematics, and TALYS calculations. The new values for the "5"0Cr(n,t)"4"8V cross-section measured using γ-spectrometry at 15, 16, 17.3 MeV are presented. The trend of the results confirms that while early experimental data at 14.6 MeV are strongly overestimated, the calculations performed with the default version of TALYS strongly underestimate the excitation curve in the measured energy region

  17. Recent developments in nuclear reaction theories and calculations

    International Nuclear Information System (INIS)

    Gardner, D.G.

    1980-01-01

    A brief review is given of some recent developments in the fields of optical model potentials; level densities; and statistical model, precompound, and direct reaction codes and calculations. Significant developments have occurred in all of these fields since the 1977 Conference on Neutron Cross Sections, which will greatly enhance the ability to calculate high-energy neutron-induced reaction cross sections in the next few years. 11 figures, 3 tables

  18. Measurement of double differential cross sections for light charged particles production in neutron induced reaction at 62.7 MeV on lead target; Mesures des sections efficaces doublement differentielles de production de particules chargees legeres lors de reactions induites par neutrons de 62.7 MeV sur cible de plomb

    Energy Technology Data Exchange (ETDEWEB)

    Kerveno, M

    2000-09-27

    In order to develop new options for nuclear waste management, studies are carrying out on the perfecting of hybrid systems (sub-critical reactor driven by accelerator). This thesis work takes place more precisely in the framework of nuclear data linked to hybrid systems development. Increasing the upper limit energy value (from 20 to 150 MeV) of data bases supposes that theoretical codes could have sufficient predictive power in this energy range. Thus it's necessary to measure new cross sections to constrain these codes. The experiment, performed at Louvain-la-Neuve Cyclotron, aims to determine the double differential cross sections for light charged particles production in neutron induced reactions at 62.7 MeV on natural lead target. The detection device consists of 6 NE102-CsI telescopes. Time of flight measurements are used to reconstruct the neutron energy spectra. The general framework (hybrid systems and associated nuclear data problematic) in which this work takes place is presented in a first part. The experimental set up used for our measurements is described in a second part. The three following parts are dedicated to the data analysis and double differential cross sections extraction. The particle discrimination, the energy calibration of detectors as the different corrections applied to the experimental spectra are related in details. And finally a comparative study between our experimental results and some theoretical predictions is presented. (author)

  19. Measurement of double differential cross sections for light charged particles production in neutron induced reaction at 62.7 MeV on lead target; Mesures des sections efficaces doublement differentielles de production de particules chargees legeres lors de reactions induites par neutrons de 62.7 MeV sur cible de plomb

    Energy Technology Data Exchange (ETDEWEB)

    Kerveno, M

    2000-09-27

    In order to develop new options for nuclear waste management, studies are carrying out on the perfecting of hybrid systems (sub-critical reactor driven by accelerator). This thesis work takes place more precisely in the framework of nuclear data linked to hybrid systems development. Increasing the upper limit energy value (from 20 to 150 MeV) of data bases supposes that theoretical codes could have sufficient predictive power in this energy range. Thus it's necessary to measure new cross sections to constrain these codes. The experiment, performed at Louvain-la-Neuve Cyclotron, aims to determine the double differential cross sections for light charged particles production in neutron induced reactions at 62.7 MeV on natural lead target. The detection device consists of 6 NE102-CsI telescopes. Time of flight measurements are used to reconstruct the neutron energy spectra. The general framework (hybrid systems and associated nuclear data problematic) in which this work takes place is presented in a first part. The experimental set up used for our measurements is described in a second part. The three following parts are dedicated to the data analysis and double differential cross sections extraction. The particle discrimination, the energy calibration of detectors as the different corrections applied to the experimental spectra are related in details. And finally a comparative study between our experimental results and some theoretical predictions is presented. (author)

  20. Reaction cross section calculation of some alkaline earth elements

    Science.gov (United States)

    Tel, Eyyup; Kavun, Yusuf; Sarpün, Ismail Hakki

    2017-09-01

    Reaction cross section knowledge is crucial to application nuclear physics such as medical imaging, radiation shielding and material evaluations. Nuclear reaction codes can be used if the experimental data are unavailable or are improbably to be produced because of the experimental trouble. In this study, there action cross sections of some target alkaline earth elements have been calculated by using pre-equilibrium and equilibrium nuclear reaction models for nucleon induced reactions. While these calculations, the Hybrid Model, the Geometry Dependent Hybrid Model, the Full Exciton Model, the Cascade Exciton Model for pre-equilibrium reactions and the Weisskopf-Ewing Model for equilibrium reactions have been used. The calculated cross sections have been discussed and compared with the experimental data taken from Experimental Nuclear Reaction Data library.

  1. Reaction cross section calculation of some alkaline earth elements

    Directory of Open Access Journals (Sweden)

    Tel Eyyup

    2017-01-01

    Full Text Available Reaction cross section knowledge is crucial to application nuclear physics such as medical imaging, radiation shielding and material evaluations. Nuclear reaction codes can be used if the experimental data are unavailable or are improbably to be produced because of the experimental trouble. In this study, there action cross sections of some target alkaline earth elements have been calculated by using pre-equilibrium and equilibrium nuclear reaction models for nucleon induced reactions. While these calculations, the Hybrid Model, the Geometry Dependent Hybrid Model, the Full Exciton Model, the Cascade Exciton Model for pre-equilibrium reactions and the Weisskopf-Ewing Model for equilibrium reactions have been used. The calculated cross sections have been discussed and compared with the experimental data taken from Experimental Nuclear Reaction Data library.

  2. Calculation of deflection for cross laminated timber floor panel

    Directory of Open Access Journals (Sweden)

    Kozarić Ljiljana M.

    2016-01-01

    Full Text Available In this paper analytically calculated values of effective flexural stiffness and deflections of five-layer CLT panels height 14 cm due to the payload defined in Eurocode 1 for floors in residential buildings are compared. Effective flexural stiffness was calculated using Gamma method, K-method and Kreuzinger's analogy. Three floor panels with identical height but with different combinations of lamination thicknesses in cross-layers were analyzed. The panels are 4.5 meters long and 1 meter wide. Lamination thicknesses in cross-sections of panels are 33,4 cm+21,9 cm, then 33 cm+22,5 cm and 52,8 cm.

  3. Group constant preparation for the estimate of neutron induced damage in structural materials

    International Nuclear Information System (INIS)

    Panini, G.C.

    1996-01-01

    Neutron heating (kerma), displacement per atom cross sections (DPA), gas and γ-ray production are important parameters for the estimate of the damage produced by neutron induced nuclear reactions in the structural materials. The NJOY System for Nuclear Data Processing has been extensively used in order to compute the above quantities; here the theory, the algorithms and the connected problems are described. (author). 6 refs, 3 tabs

  4. SCALP: Scintillating ionization chamber for ALPha particle production in neutron induced reactions

    Science.gov (United States)

    Galhaut, B.; Durand, D.; Lecolley, F. R.; Ledoux, X.; Lehaut, G.; Manduci, L.; Mary, P.

    2017-09-01

    The SCALP collaboration has the ambition to build a scintillating ionization chamber in order to study and measure the cross section of the α-particle production in neutron induced reactions. More specifically on 16O and 19F targets. Using the deposited energy (ionization) and the time of flight measurement (scintillation) with a great accuracy, all the nuclear reaction taking part on this project will be identify.

  5. Terrestrial neutron-induced soft errors in advanced memory devices

    CERN Document Server

    Nakamura, Takashi; Ibe, Eishi; Yahagi, Yasuo; Kameyama, Hideaki

    2008-01-01

    Terrestrial neutron-induced soft errors in semiconductor memory devices are currently a major concern in reliability issues. Understanding the mechanism and quantifying soft-error rates are primarily crucial for the design and quality assurance of semiconductor memory devices. This book covers the relevant up-to-date topics in terrestrial neutron-induced soft errors, and aims to provide succinct knowledge on neutron-induced soft errors to the readers by presenting several valuable and unique features. Sample Chapter(s). Chapter 1: Introduction (238 KB). Table A.30 mentioned in Appendix A.6 on

  6. Measurement, calculation and evaluation of photon production cross-sections

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1990-11-01

    The meeting proceedings were divided into three sessions devoted to the following topics: Experimental measurement and techniques (3 papers), calculation of photon cross-sections (9 papers), and evaluation (2 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  7. RZ calculations for self shielded multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

    2006-07-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  8. RZ calculations for self shielded multigroup cross sections

    International Nuclear Information System (INIS)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

    2006-01-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  9. Detection of buried land mines using back scattered neutron induced γ-ray analysis

    International Nuclear Information System (INIS)

    Aziz, M.; Megahd, R.

    2003-01-01

    The application of nuclear technique to detection buried land mine is examined. MCNP code was used to design a computer model that calculate the back scattered neutron induced γ rays from buried simulate explosive materials. The characteristic γ rays for each isotopes were used to distinguish materials. The advantage of the nuclear technique was discussed. The results were compared with experimental measurements which show good agreement

  10. First and second chance fission calculations for actinides and related topics

    International Nuclear Information System (INIS)

    Maino, G.; Menapace, E.; Motta, M.; Ventura, A.

    1980-01-01

    First and second chance contributions to neutron induced fission cross sections in an energy range of interest for reactor applications (E/sub n/less than or equal to 13 MeV) were obtained by extensive and consistent calculations for 241 Am; moreover, a simplified semiempirical approach was applied to 235 U and 239 Pu

  11. Low-energy neutron-induced single-event upsets in static random access memory

    International Nuclear Information System (INIS)

    Guo Xiaoqiang; Guo Hongxia; Wang Guizhen; Ling Dongsheng; Chen Wei; Bai Xiaoyan; Yang Shanchao; Liu Yan

    2009-01-01

    The visual analysis method of data process was provided for neutron-induced single-event upset(SEU) in static random access memory(SRAM). The SEU effects of six CMOS SRAMs with different feature size(from 0.13 μm to 1.50 μm) were studied. The SEU experiments were performed using the neutron radiation environment at Xi'an pulsed reactor. And the dependence of low-energy neutron-induced SEU cross section on SRAM's feature size was given. The results indicate that the decreased critical charge is the dominant factor for the increase of single event effect sensitivity of SRAM devices with decreased feature size. Small-sized SRAM devices are more sensitive than large-sized ones to single event effect induced by low-energy neutrons. (authors)

  12. Neutron induced fission of 237Np – status, challenges and opportunities

    Directory of Open Access Journals (Sweden)

    Ruskov Ivan

    2018-01-01

    Full Text Available Nowadays, there is an increased interest in a complete study of the neutron-induced fission of 237Np. This is due to the need of accurate and reliable nuclear data for nuclear science and technology. 237Np is generated (and accumulated in the nuclear reactor core during reactor operation. As one of the most abundant long-lived isotopes in spent fuel (“waste”, the incineration of 237Np becomes an important issue. One scenario for burning of 237Np and other radio-toxic minor actinides suggests they are to be mixed into the fuel of future fast-neutron reactors, employing the so-called transmutation and partitioning technology. For testing present fission models, which are at the basis of new generation nuclear reactor developments, highly accurate and detailed neutron-induced nuclear reaction data is needed. However, the EXFOR nuclear database for 237Np on neutron-induced capture cross-section, σγ, and fission cross-section, σf, as well as on the characteristics of capture and fission resonance parameters (Γγ, Γf, σoΓf, fragments mass-energy yield distributions, multiplicities of neutrons vn and γ-rays vγ, has not been updated for decades.

  13. Average cross sections calculated in various neutron fields

    International Nuclear Information System (INIS)

    Shibata, Keiichi

    2002-01-01

    Average cross sections have been calculated for the reactions contained in the dosimetry files, JENDL/D-99, IRDF-90V2, and RRDF-98 in order to select the best data for the new library IRDF-2002. The neutron spectra used in the calculations are as follows: 1) 252 Cf spontaneous fission spectrum (NBS evaluation), 2) 235 U thermal fission spectrum (NBS evaluation), 3) Intermediate-energy Standard Neutron Field (ISNF), 4) Coupled Fast Reactivity Measurement Facility (CFRMF), 5) Coupled thermal/fast uranium and boron carbide spherical assembly (ΣΣ), 6) Fast neutron source reactor (YAYOI), 7) Experimental fast reactor (JOYO), 8) Japan Material Testing Reactor (JMTR), 9) d-Li neutron spectrum with a 2-MeV deuteron beam. The items 3)-7) represent fast neutron spectra, while JMTR is a light water reactor. The Q-value for the d-Li reaction mentioned above is 15.02 MeV. Therefore, neutrons with energies up to 17 MeV can be produced in the d-Li reaction. The calculated average cross sections were compared with the measurements. Figures 1-9 show the ratios of the calculations to the experimental data which are given. It is found from these figures that the 58 Fe(n, γ) cross section in JENDL/D-99 reproduces the measurements in the thermal and fast reactor spectra better than that in IRDF-90V2. (author)

  14. Hydrogen isotope double differential production cross sections induced by 62.7 MeV neutrons on a lead target

    International Nuclear Information System (INIS)

    Kerveno, M.; Haddad, F.; Eudes, Ph.; Kirchner, T.; Lebrun, C.; Slypen, I.; Meulders, J.P.; Le Brun, C.; Lecolley, F.R.; Lecolley, J.F.; Louvel, M.; Lefebvres, F.; Hilaire, S.; Koning, A.J.

    2002-01-01

    Double differential hydrogen isotope production cross sections have been extracted in 62.7 MeV neutron induced reactions on a lead target. The angular distribution was measured at eight angles from 20 deg. to 160 deg. allowing the extraction of angle-differential, energy differential, and total production cross sections. A first set of comparisons with several theoretical calculations is also presented

  15. Measurement, calculation and evaluation of photon production cross-sections

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1990-03-01

    The IAEA Specialists' Meeting on Measurement, Calculation and Evaluation of Photon Production Cross-Sections was held in Smolenice, Czechoslovakia, 5-7 February 1990. The meeting was hosted by the Institute of Physics of the Electro-Physical Research Centre, Slovak Academy of Sciences, Bratislava. This report contains the conclusions and recommendations of this meeting. The papers which the participants have presented at the meeting will be published as an INDC Report. (author)

  16. Hauser*5, a computer code to calculate nuclear cross sections

    International Nuclear Information System (INIS)

    Mann, F.M.

    1979-07-01

    HAUSER*5 is a computer code that uses the statistical (Hauser-Feshbach) model, the pre-equilibrium model, and a statistical model of direct reactions to predict nuclear cross sections. The code is unrestricted as to particle type, includes fission and capture, makes width-fluctuation corrections, and performs three-body calculations - all in minimum computer time. Transmission coefficients can be generated internally or supplied externally. This report describes equations used, necessary input, and resulting output. 2 figures, 4 tables

  17. Computer Package for Graphite Total Cross-Section Calculations

    International Nuclear Information System (INIS)

    Adib, M.; Fathalla, M.

    2008-01-01

    An additive formula is given which allows calculating the contribution of the total neut.>neutron transmission through crystalline graphite. The formula takes into account the graphite form of poly or pyrolytic crystals and its parameters. Computer package Graphite has been designed in order to provide the required calculations in the neutron energy range from 0.1 MeV to 10 eV. The package includes three codes: PCG (Polycrystalline Graphite), PG (Pyrolytic Graphite) and HOPG (Highly Oriented Pyrolytic Graphite) for calculating neutron transmission through fine graphite powder (polycrystalline), neutron transmission and removal coefficient of PG crystal in terms of its mosaic spread for neutrons incident along its c-axis and the transmission of neutrons incident on HOPG crystal at different angles, respectively. For comparison of the experimental neutron transmission data with the calculated values, the program takes into consideration the effect of both wavelength and neutron beam divergence in either 2 constant wavelength spread mode (δλ=constant) or constant wavelength resolution mode (δλ/λ=constant). In order to check the validity for application of computer package Graphite in cross-section calculations, a comparison between calculated values with the available experimental data were carried out. An overall agreement is indicated with an accuracy sufficient for determine the neutron transmission characteristics

  18. Evaluation and calculation of neutron transactinide cross-sections

    International Nuclear Information System (INIS)

    Konshin, V.A.

    1980-01-01

    This paper reviews the state of the art of nuclear theory and its application to the evaluation and calculation of neutron reaction cross sections of transactinium isotopes. In particular, the paper describes the current evaluation of the total files of neutron reaction data for 240 Pu and 241 Pu in the energy range between 10 -5 eV and 15 MeV based on a thorough analysis of available experimental data and on the use of modern theoretical concepts, and the work in progress on the evaluation of the total neutron reaction data file for 242 Pu and 241 Am. (author)

  19. Optical model calculation of neutron-nucleus scattering cross sections

    International Nuclear Information System (INIS)

    Smith, M.E.; Camarda, H.S.

    1980-01-01

    A program to calculate the total, elastic, reaction, and differential cross section of a neutron interacting with a nucleus is described. The interaction between the neutron and the nucleus is represented by a spherically symmetric complex potential that includes spin-orbit coupling. This optical model problem is solved numerically, and is treated with the partial-wave formalism of scattering theory. The necessary scattering theory required to solve this problem is briefly stated. Then, the numerical methods used to integrate the Schroedinger equation, calculate derivatives, etc., are described, and the results of various programming tests performed are presented. Finally, the program is discussed from a user's point of view, and it is pointed out how and where the program (OPTICAL) can be changed to satisfy particular needs

  20. ACDOS2: a code for neutron-induced activities and dose rates

    International Nuclear Information System (INIS)

    Ruby, L.; Keney, G.S.; Lagache, J.C.

    1981-10-01

    In order to anticipate problems from the radioactivation of neutral beam sources as a result of testing, a code has been developed which calculates both the radioactivities produced and the dose rates resulting therefrom. The code ACDOS2 requires neutron source strength and spectral distribution as input, or alternately, the source strength can be calculated internally from an input of neutral beam source parameters. A variety of simple geometries can be specified, and up to 12 times of interest following the shutdown of the neutron source. Radiation attenuating and daughter radioactivities are treated accurately. ACDOS2 is also of use for neutron-induced radioactivation problems involving accelerators, fusion reactors, or fission reactors

  1. Trojan Horse Method for neutrons-induced reaction studies

    Science.gov (United States)

    Gulino, M.; Asfin Collaboration

    2017-09-01

    Neutron-induced reactions play an important role in nuclear astrophysics in several scenario, such as primordial Big Bang Nucleosynthesis, Inhomogeneous Big Bang Nucleosynthesis, heavy-element production during the weak component of the s-process, explosive stellar nucleosynthesis. To overcome the experimental problems arising from the production of a neutron beam, the possibility to use the Trojan Horse Method to study neutron-induced reactions has been investigated. The application is of particular interest for reactions involving radioactive nuclei having short lifetime.

  2. Status of experimental data for neutron induced reactions

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Mamoru [Tohoku Univ., Sendai (Japan)

    1998-11-01

    A short review is presented on the status of experimental data for neutron induced reactions above 20 MeV based on the EXFOR data base and journals. Experimental data which were obtained in a systematic manner and/or by plural authors are surveyed and tabulated for the nuclear data evaluation and the benchmark test of the evaluated data. (author). 61 refs.

  3. The PSIMECX medium-energy neutron activation cross-section library. Part III: Calculational methods for heavy nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F.

    1998-09-01

    The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: {sup 12}C, {sup 13}C, {sup 16}O, {sup 17}O, {sup 18}O, {sup 23}Na, {sup 24}Mg, {sup 25}Mg, {sup 26}Mg, {sup 27}Al, {sup 28}Si, {sup 29}Si, {sup 30}Si, {sup 31}P, {sup 32}S, {sup 33}S, {sup 34}S, {sup 36}S, {sup 35}Cl, {sup 37}Cl, {sup 39}K, {sup 40}K, {sup 41}K, {sup 40}Ca, {sup 42}Ca, {sup 43}Ca, {sup 44}Ca, {sup 46}Ca, {sup 48}Ca, {sup 46}Ti, {sup 47}Ti, {sup 48}Ti, {sup 49}Ti, {sup 50}Ti, {sup 50}V, {sup 51}V, {sup 50}Cr, {sup 52}Cr, {sup 53}Cr, {sup 54}Cr, {sup 55}Mn, {sup 54}Fe, {sup 56}Fe, {sup 57}Fe, {sup 58}Fe, {sup 58}Ni, {sup 60}Ni, {sup 61}Ni, {sup 62}Ni, {sup 64}Ni, {sup 63}Cu, {sup 65}Cu, {sup 64}Zn, {sup 66}Zn, {sup 67}Zn, {sup 68}Zn, {sup 70}Zn, {sup 92}Mo, {sup 94}Mo, {sup 95}Mo, {sup 96}Mo, {sup 97}Mo, {sup 98}Mo, {sup 100}Mo, {sup 121}Sb, {sup 123}Sb, {sup 204}Pb, {sup 206}Pb, {sup 207}Pb, {sup 208}Pb, {sup 232}Th and {sup 238}U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are main constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This third report describes and discusses the calculational methods used for the heavy nuclei. The library itself has been described in the first report of this series and the treatment for the medium and light mass nuclei is given in the second. (author)

  4. Simplified polynomial representation of cross sections for reactor calculation

    International Nuclear Information System (INIS)

    Dias, A.M.; Sakai, M.

    1985-01-01

    It is shown a simplified representation of a cross section library generated by transport theory using the cell model of Wigner-Seitz for typical PWR fuel elements. The effect of burnup evolution through tables of reference cross sections and the effect of the variation of the reactor operation parameters considered by adjusted polynomials are presented. (M.C.K.) [pt

  5. Measurements of fast neutron-induced fission data of Np-237

    Energy Technology Data Exchange (ETDEWEB)

    Win, Than; Saito, Keiichiro; Baba, Mamoru; Iwasaki, Tomohiko; Ibaraki, Masanobu; Miura, Takako; Sanami, Toshiya; Nauchi, Yasushi; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan). Faculty of Engineering

    1998-03-01

    We have performed the following measurements for {sup 237}Np using the 4.5 MV Dynamitron accelerator of Tohoku University as the pulsed neutron source: (1) Prompt fission neutron spectrum for 0.62 MeV incident neutrons, and (2) Neutron-Induced fission cross-section between 10 and 100 keV. The prompt fission neutron spectrum was measured using TOF method with a heavily shielded NE213 scintillation detector. The Maxwellian temperature T{sub m} derived is 1.28 MeV, which is lower than that of 1.38 MeV in JENDL-3.2. The fission cross sections were measured between 10 - 100 keV. The results are between JENDL-3.2 and ENDF/B-VI. (author)

  6. Photoionization cross sections and Auger rates calculated by many-body perturbation theory

    International Nuclear Information System (INIS)

    Kelly, H.P.

    1976-01-01

    Methods for applying the many body perturbation theory to atomic calculations are discussed with particular emphasis on calculation of photoionization cross sections and Auger rates. Topics covered include: Rayleigh--Schroedinger theory; many body perturbation theory; calculations of photoionization cross sections; and Auger rates

  7. 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.

  8. 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

  9. Radiation damage calculations for compound materials

    International Nuclear Information System (INIS)

    Greenwood, L.R.

    1990-01-01

    This paper reports on the SPECOMP computer code, developed to calculate neutron-induced displacement damage cross sections for compound materials such as alloys, insulators, and ceramic tritium breeders for fusion reactors. These new calculations rely on recoil atom energy distributions previously computed with the DISCS computer code, the results of which are stored in SPECTER computer code master libraries. All reaction channels were considered in the DISCS calculations and the neutron cross sections were taken from ENDF/B-V. Compound damage calculations with SPECOMP thus do not need to perform any recoil atom calculations and consequently need no access to ENDF or other neutron data bases. The calculations proceed by determining secondary displacements for each combination of recoil atom and matrix atom using the Lindhard partition of the recoil energy to establish the damage energy

  10. Calculated Cross Sections for the Electron Impact Ionization of Molecular Ions

    Science.gov (United States)

    Deutsch, H.; Becker, K.; Defrance, P.; Onthong, U.; Parajuli, R.; Probst, M.; Matt-Leubner, S.; Maerk, T.

    2002-10-01

    We report the results of the application of the semi- classical Deutsch-Märk (DM) formalism to the calculation of the absolute electron-impact ionization cross section of the molecular ions H2+, N2+, O2+, CD+, CO+, CO2+, H3O+, and CH4+ for which experimental data have been reported . Where available, we also compare our calculated cross sections with calculated cross sections using the BEB method of Kim and co-workers. The level of agreement between the experimentally determined and calculated cross section is satisfactory in some cases. In all cases, the calculated cross sections exceed the measured cross sections which is not surprising in view of the experimental complications in measuring ionization cross sections of molecular ions due to the presence of competing channels such as ionization dissociative ionization, and dissociative excitation. Work supported in part by FWF, OEAW, and NASA.

  11. Calculation of the photoionization cross section of the 4d10 subshell of the La atom

    International Nuclear Information System (INIS)

    Amusia, M.Ya.; Sheftel, S.I.

    1976-01-01

    The photoionization cross section of 4d 10 subshell of La atom is calculated. The cross section curve near its threshold is strongly modified by rearrangement of outer shells in the process of photoionization. (Auth.)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  13. Effect of core polarizability on photoionization cross-section calculations.

    Science.gov (United States)

    Kirkpatrick, R. C.

    1972-01-01

    Demonstration of the importance of core polarizability in a case where cancellation is only moderate, with suggestion of an improvement to the scaled Thomas-Fermi (STF) wave functions of Stewart and Rotenberg (1965). The inclusion of dipole polarizability of the core for argon is shown to substantially improve the agreement between the theoretical and experimental photoionization cross sections for the ground-state configuration.

  14. Consistent interpretation of neutron-induced charged-particle emission in silicon

    International Nuclear Information System (INIS)

    Hermsdorf, D.

    1982-06-01

    Users requesting gas production cross sections for Silicon will be confronted with serious discrepancies taking evaluated data as well as experimental ones. To clarify the accuracies achieved at present in experiments and evaluations in this paper an intercomparison of different evaluated nuclear data files has been carried out resulting in recommendations for improvements of these files. The analysis of the experimental data base also shows contradictory measurements or in most cases a lack of data. So an interpretation of reliable measured data in terms of nuclear reaction theories has been done using statistical and direct reaction mechanism models. This study results in a consistent and comprehensive evaluated data set for neutron-induced charged-particle production in Silicon which will be incorporated in file 2015 of the SOKRATOR library. (author)

  15. Charged-particle magnetic-quadrupole spectrometer for neutron induced reactions

    International Nuclear Information System (INIS)

    Haight, R.C.; Grimes, S.M.; Tuckey, B.J.; Anderson, J.D.

    1975-01-01

    A spectrometer has been developed for measuring the charged particle production cross sections and spectra in neutron-induced reactions. The spectrometer consists of a magnetic quadrupole doublet which focuses the charged particles onto a silicon surface barrier detector telescope which is 2 meters or more from the irradiated sample. Collimators, shielding, and the large source-to-detector distance reduce the background enough to use the spectrometer with a 14-MeV neutron source producing 4 . 10 12 n/s. The spectrometer has been used in investigations of proton, deuteron, and alpha particle production by 14-MeV neutrons incident on various materials. Protons with energies as low as 1.1 MeV have been measured. The good resolution of the detectors has also made possible an improved measurement of the neutron- neutron scattering length from the 0 0 proton spectrum from deuteron breakup by 14-MeV neutrons

  16. Comparison of Neutron Cross-Sections Using IAEA Nuclear Codes ''ABAREX'' and ''SCAT2''

    International Nuclear Information System (INIS)

    Myint Myint Moe; Win Sin; Sein Htoon

    2004-05-01

    Moel calculations can be used to provide nuclear data for applications in science and technology. The energy averaged neutron induced nuclear reaction cross-sections particular for Al-27, Mg-24, Cr-52, Mn-55, Zn-64 and U-238 with neutrons of energy (0.005 to 10 MeV) are calculated using IAEA nuclear codes ''ABAREX'' and ''SCAT2''. The results are compared with those given in ENDF- 3 nuclear data

  17. ACDOS2: an improved neutron-induced dose rate code

    International Nuclear Information System (INIS)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere

  18. ACDOS2: an improved neutron-induced dose rate code

    Energy Technology Data Exchange (ETDEWEB)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.

  19. Calculation of inelastic cross-sections for: H++Cs→H(n=2)+Cs+

    International Nuclear Information System (INIS)

    Valance, A.; Spiess, G.

    1975-01-01

    The cross sections for the processes H + + Cs→H(2p and 2s) + Cs + were calculated in the center of mass energy range 250-2400eV using a pseudo-potential formalism for the potential curves and coupling matrix elements and a perturbated stationary state formulation for the calculation of the cross sections [fr

  20. Process of cross section generation for radiation shielding calculations, using the NJOY code

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1986-10-01

    The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt

  1. Neutron-induced mutation experiments. Comprehensive report, March 1, 1977-August 31, 1980

    International Nuclear Information System (INIS)

    Abrahamson, S.

    1981-02-01

    Neutron-induced X-linked lethal mutations were induced in Drosophila melanogaster oogonia at energies of .43, .66, 2, and 6 MeV. The 37 irradiations were carried out at the RARAF facility at Brookhaven National Laboratory. RBE's (relative to x-ray data similarly collected) were calculated to be .43 MeV to 4.8; .66 MeV to 4.0; 2 MeV to 3.2; and 6 MeV to 2.9. The dose/frequency response curves for all energies best fit a linear rather than a linear-quadratic model following regression analyses. Control data for specific locus mutations (420,000 tests) were gathered. This data, combined with other data (both X-linked lethal and specific locus) has been used to estimate the number of loci on the X-chromosome of Drosophila which can mutate to recessive lethals

  2. An evaporation-based model of thermal neutron induced ternary fission of plutonium

    International Nuclear Information System (INIS)

    Lestone, J.P.

    2008-01-01

    Ternary fission probabilities for thermal neutron induced fission of plutonium are analyzed within the framework of an evaporation-based model where the complexity of time-varying potentials, associated with the neck collapse, are included in a simplistic fashion. If the nuclear temperature at scission and the fission-neck-collapse time are assumed to be ~ 1.2 MeV and ~ 10 -22 s, respectively, then calculated relative probabilities of ternary-fission light-charged-particle emission follow the trends seen in the experimental data. The ability of this model to reproduce ternary fission probabilities spanning seven orders of magnitude for a wide range of light-particle charges and masses implies that ternary fission is caused by the coupling of an evaporation-like process with the rapid re-arrangement of the nuclear fluid following scission. (author)

  3. The PSIMECX medium-energy neutron activation cross-section library. Part II: Calculational methods for light to medium mass nuclei

    International Nuclear Information System (INIS)

    Atchison, F.

    1998-09-01

    The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: 12 C, 13 C, 16 O, 17 O, 18 O, 23 Na, 24 Mg, 25 Mg, 26 Mg, 27 Al, 28 Si, 29 Si, 30 Si, 31 P, 32 S, 33 S, 34 S, 36 S, 35 Cl, 37 Cl, 39 K, 40 K, 41 K, 40 Ca, 42 Ca, 43 Ca, 44 Ca, 46 Ca, 48 Ca, 46 Ti, 47 Ti, 48 Ti, 49 Ti, 50 Ti, 50 V, 51 V, 50 Cr, 52 Cr, 53 Cr, 54 Cr, 55 Mn, 54 Fe, 56 Fe, 57 Fe, 58 Fe, 58 Ni, 60 Ni, 61 Ni, 62 Ni, 64 Ni, 63 Cu, 65 Cu, 64 Zn, 66 Zn, 67 Zn, 68 Zn, 70 Zn, 92 Mo, 94 Mo, 95 Mo, 96 Mo, 97 Mo, 98 Mo, 100 Mo, 121 Sb, 123 Sb, 204 Pb, 206 Pb, 207 Pb, 208 Pb, 232 Th and 238 U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are principal constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This second report, of a series of three, describes and discusses the calculational methods used for the stable isotopes up to and including 123 Sb. The library itself has been described in the first report of the series and the treatment for the heavy nuclei is given in the third. (author)

  4. To calculation of electron-optical characteristics of crossed lenses

    International Nuclear Information System (INIS)

    Afanas'ev, V.P.; Gritsyuk, N.P.; Lachashvili, R.A.; Yavor, S.Ya.

    1979-01-01

    Two approaches are used for theoretical study of crossed lenses (CL), which are formed by plates with slits turned by 90 deg: 1) aberration determination according to axial determination of potential and 2) trajectory analysis of CL. While studying CL of definite configuration it is necessary to take into account aberrations of the highest order. The following conclusions are drawn: the classical method is applied for fast determination of the main characteristics, of CL by means of average power computers (the ''M-220'' or the ''Minsk-32'' computers); the trajectory analysis should be performed by means of power computer (the BESM-6) when it is necessary to obtain more detailed information of the CL, including the trajectory deflection along the system axis, the dimensions of the point source image taking into account the aberrations of the highest order

  5. Progress on calculation of direct inelastic scattering cross section of neutron

    Energy Technology Data Exchange (ETDEWEB)

    Zhenpeng, Chen [Qinghua Univ., Beijing, BJ (China). Dept. of Physics

    1996-06-01

    For n+ {sup 238}U inelastic scattering cross, there exist discrepancies among the available evaluations in various libraries. This is partly duo to the difference of direct inelastic scattering cross section calculated with coupled channel optical model (CCOM). The research on the level frame used in CCOM calculation, the research on used parameters of spherical optical model in CCOM calculation and the research on the amplitude of octupole phonon {beta}{sub 3} were presented. (2 figs.).

  6. Determination of equivalent cross sections for representation of control rod regions in diffusion calculations

    International Nuclear Information System (INIS)

    Scherer, W.; Neef, H.J.

    1976-07-01

    The representation of control rod regions in reactor calculations requires a combination of transport and diffusion theory calculations. A method is described which produces equivalent cross sections for a rodded region. These cross sections used in a diffusion theory calcualtion yield the same rod efficiency and reaction rate distribution as the transport theory calculation for the explicit heterogeneous control rod. The description of the method is complemented by sample problems. (orig.) [de

  7. Brief note on the statistical calculation of final continuum reaction cross sections of light nuclides

    International Nuclear Information System (INIS)

    Murata, Toru

    2003-01-01

    The level density parameters are determined to reproduce level structure and/or resonance level spacing of the nucleus. In the statistical compound nucleus model, cross sections to discrete levels decrease abruptly, and continuum level cross section increase strongly above the energy point where the continuum levels switched on. In the present study, for the nucleus which level scheme were well determined up to higher excitation energy more than 10 MeV, discrete level cross sections were calculated and summed up and compared with the cross section to the assumed continuum level corresponding to the discrete levels above several MeV excitation energy. Calculation of the (n, n') cross sections were made with CASTHY code of Moldauer model option using level density parameters determined with former method. It is shown that the assumed continuum cross section is fairly large compared with the summed up cross section. Origins of the discrepancy were discussed. (J.P.N.)

  8. Evaluation of the neutron induced reactions on 235U from 2.25 keV up to 30 MeV

    Science.gov (United States)

    Trkov, Andrej; Capote, Roberto; Pigni, Marco T.; Pronyaev, Vladimir G.; Sin, Mihaela; Soukhovitskii, Efrem S.

    2017-09-01

    An evaluation of fast neutron induced reactions on 235U is performed in the 2.25 keV-30 MeV incident energy range with the code EMPIRE-3.2 Malta, combined with selected experimental data. The reaction model includes a dispersive optical model potential (RIPL 2408) that couples seven levels of the ground-state rotational band and a triple-humped fission barrier with absorption in the wells described within the optical model for fission. EGSM nuclear level densities are used in Hauser-Feshbach calculations of the compound-nuclear decay. The starting values for the model parameters are retrieved from the RIPL-3 data-base. Excellent agreement is achieved with available experimental data for neutron emission, neutron capture and fission, which gives confidence that the quantities for which there is no experimental information are also predicted accurately. In the fast neutron region of the evaluated file, the fission cross section is taken from Neutron Standards, and neutron capture includes fluctuations observed in recent experiments. Other channels are taken directly from model calculations. New evaluation is validated against ICSBEP criticality benchmarks with fast neutron spectra with excellent results.

  9. Evaluation of the neutron induced reactions on 235U from 2.25 keV up to 30 MeV

    Directory of Open Access Journals (Sweden)

    Trkov Andrej

    2017-01-01

    Full Text Available An evaluation of fast neutron induced reactions on 235U is performed in the 2.25 keV–30 MeV incident energy range with the code EMPIRE–3.2 Malta, combined with selected experimental data. The reaction model includes a dispersive optical model potential (RIPL 2408 that couples seven levels of the ground-state rotational band and a triple-humped fission barrier with absorption in the wells described within the optical model for fission. EGSM nuclear level densities are used in Hauser-Feshbach calculations of the compound-nuclear decay. The starting values for the model parameters are retrieved from the RIPL-3 data-base. Excellent agreement is achieved with available experimental data for neutron emission, neutron capture and fission, which gives confidence that the quantities for which there is no experimental information are also predicted accurately. In the fast neutron region of the evaluated file, the fission cross section is taken from Neutron Standards, and neutron capture includes fluctuations observed in recent experiments. Other channels are taken directly from model calculations. New evaluation is validated against ICSBEP criticality benchmarks with fast neutron spectra with excellent results.

  10. Neutron induced degradation in nitrided pyrogenic field oxide MOS capacitors

    CERN Document Server

    Vaidya, S J; Shaikh, A M; Chandorkar, A N

    2002-01-01

    Neutron induced oxide charge trapping and generation of interface states in MOS capacitors with pyrogenic and nitrided pyrogenic field oxides have been studied. In order to assess the damage due to neutrons alone, it is necessary to account for the damage produced by the accompanying gamma rays from neutron radiation. This is done by measuring the intensity of gamma radiation accompanying neutrons at different neutron fluences at the irradiation position. MOS capacitor structures were subjected to neutron radiation in a swimming pool type of reactor. Other samples from the same batch were then subjected to an equivalent dose of gamma radiation from a Co sup 6 sup 0 source. The difference in the damage observed was used to characterize the damage caused by neutrons. It is observed that neutrons, though uncharged, are capable of causing ionization damage. This damage is found to be significant when the radiation is performed under biased conditions. Nitridation in different ambients is found to improve the radi...

  11. Boron determination in tourmaline by neutron induced radiography

    Energy Technology Data Exchange (ETDEWEB)

    Qureshi, A.A. E-mail: aaqureshi@pinstech.org.pk; Akram, M.; Ayub Khan, M.; Khattak, N.U.; Qureshi, I.E.; Khan, H.A

    2001-06-01

    The technique of neutron induced radiography has been applied to determine the boron concentration and its spatial distribution in mineral tourmaline collected from Swat Tourmaline Granite, Northern Pakistan. The technique involves the simultaneous irradiation of sample and a standard fixed on a track detector with thermal neutrons and the counting of alpha and {sup 7}Li tracks produced in the detector from the nuclear reaction {sup 10}B(n,{alpha}){sup 7}Li. Boron concentration is determined by comparing the {sup 7}Li and alpha particle tracks density with that of a standard of known boron concentration. Boron concentration in tourmaline has been found to be (3.40{+-}0.01)% in this study which is on the upper side within the normal range (2.5-3.8)% reported in the world. The presence of somewhat higher concentration of boron in tourmaline indicates that the Swat Tourmaline Granite was generated as a late stage hydrothermal activity during the Himalayan Orogeny.

  12. Neutron kinetics in moderators and SNM detection through epithermal-neutron-induced fissions

    Energy Technology Data Exchange (ETDEWEB)

    Gozani, Tsahi, E-mail: tgmaven@gmail.com [1050 Harriet St., Palo Alto, CA 94301 (United States); King, Michael J. [Rapiscan Laboratories Inc., 520 Almanor Ave., Sunnyvale, CA 94085 (United States)

    2016-01-01

    Extension of the well-established Differential Die Away Analysis (DDAA) into a faster time domain, where more penetrating epithermal neutrons induce fissions, is proposed and demonstrated via simulations and experiments. In the proposed method the fissions stimulated by thermal, epithermal and even higher-energy neutrons are measured after injection of a narrow pulse of high-energy 14 MeV (d,T) or 2.5 MeV (d,D) source neutrons, appropriately moderated. The ability to measure these fissions stems from the inherent correlation of neutron energy and time (“E–T” correlation) during the process of slowing down of high-energy source neutrons in common moderating materials such as hydrogenous compounds (e.g., polyethylene), heavy water, beryllium and graphite. The kinetic behavior following injection of a delta-function-shaped pulse (in time) of 14 MeV neutrons into such moderators is studied employing MCNPX simulations and, when applicable, some simple “one-group” models. These calculations served as a guide for the design of a source moderator which was used in experiments. Qualitative relationships between slowing-down time after the pulse and the prevailing neutron energy are discussed. A laboratory system consisting of a 14 MeV neutron generator, a polyethylene-reflected Be moderator, a liquid scintillator with pulse-shape discrimination (PSD) and a two-parameter E–T data acquisition system was set up to measure prompt neutron and delayed gamma-ray fission signatures in a 19.5% enriched LEU sample. The measured time behavior of thermal and epithermal neutron fission signals agreed well with the detailed simulations. The laboratory system can readily be redesigned and deployed as a mobile inspection system for SNM in, e.g., cars and vans. A strong pulsed neutron generator with narrow pulse (<75 ns) at a reasonably high pulse frequency could make the high-energy neutron induced fission modality a realizable SNM detection technique.

  13. Modified automatic term selection v2: A faster algorithm to calculate inelastic scattering cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    Rusz, Ján, E-mail: jan.rusz@fysik.uu.se

    2017-06-15

    Highlights: • New algorithm for calculating double differential scattering cross-section. • Shown good convergence properties. • Outperforms older MATS algorithm, particularly in zone axis calculations. - Abstract: We present a new algorithm for calculating inelastic scattering cross-section for fast electrons. Compared to the previous Modified Automatic Term Selection (MATS) algorithm (Rusz et al. [18]), it has far better convergence properties in zone axis calculations and it allows to identify contributions of individual atoms. One can think of it as a blend of MATS algorithm and a method described by Weickenmeier and Kohl [10].

  14. Numerical calculation of the cross section by the solution of the wave equation

    International Nuclear Information System (INIS)

    Drewko, J.

    1982-01-01

    A numerical method of solving of the wave equation is described for chosen vibrational eigenfunctions. A prepared program calculates the total cross sections for the resonant vibrational excitation for diatomic molecules on the basis of introduced molecular data. (author)

  15. KAPSIES: A program for the calculation of multi-step direct reaction cross sections

    International Nuclear Information System (INIS)

    Koning, A.J.; Akkermans, J.M.

    1994-09-01

    We present a program for the calculation of continuum cross sections, sepctra, angular distributions and analyzing powers according to various quantum-mechanical theories for statistical multi-step direct nuclear reactions. (orig.)

  16. Calculation of inelastic cross sections for H+ + Cs → H(n=2) + Cs+

    International Nuclear Information System (INIS)

    Valance, A.; Spiess, G.

    1975-01-01

    The cross sections for the processes H + +Cs → H(2p and 2s) +Cs + were calculated in the center of mass energy range 250--2400 eV using a simple pseudopotential formalism for the potential curves and coupling matrix elements and a perturbed stationary state (pss) formulation for the calculation of the cross sections. The results are found to be in reasonable agreement with experiment. (auth)

  17. Weisskopf–Ewing and Hauser–Feshbach calculations of photonuclear cross sections used for electromagnetic dissociation

    International Nuclear Information System (INIS)

    Adamczyk, Anne M.; Norbury, John W.; Townsend, Lawrence W.

    2013-01-01

    The Weisskopf–Ewing (WE) and Hauser–Feshbach (HF) theory are statistical methods, which are often used to calculate photonuclear cross sections for compound nucleus reactions. In our past work, WE methodology was presented and photonuclear reaction cross sections for nucleon emission were calculated using WE theory. Here, our previous results, which neglect pre-equilibrium emissions and do not include multiple particle emission, are compared to those calculated with HF theory and experimental data. For the reactions considered herein, it is found that the WE theory and HF method are in reasonable agreement below the two neutron separation energy assuming an energy dependent branching ratio for intermediate and heavy nuclei. In addition, qualitative confidence of WE theory for electromagnetic dissociation (EMD) cross section calculations was found. - Author-Highlights: • The Weisskopf–Ewing (WE) theory is reviewed. • Photonuclear cross sections calculated with WE theory are compared to HF predictions. • The WE theory and the HF method give similar photonuclear cross sections. • Qualitative confidence of WE theory for EMD cross section calculations is found

  18. On calculating phase shifts and performing fits to scattering cross sections or transport properties

    International Nuclear Information System (INIS)

    Hepburn, J.W.; Roy, R.J. Le

    1978-01-01

    Improved methods of calculating quantum mechanical phase shifts and for performing least-squares fits to scattering cross sections or transport properties, are described. Their use in a five-parameter fit to experimental differential cross sections reduces the computer time by a factor of 4-7. (Auth.)

  19. Total and Compound Formation Cross Sections for Americium Nuclei: Recommendations for Coupled-Channels Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Escher, J. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-04-11

    Calculations for total cross sections and compound-nucleus (CN) formation cross sections for americium isotopes are described, for use in the 2017 NA-22 evaluation effort. The code ECIS 2006 was used in conjunction with Frank Dietrich's wrapper `runtemplate'.

  20. MUXS: a code to generate multigroup cross sections for sputtering calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1982-10-01

    This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

  1. Optical and statistical model calculation of the americium 242m capture cross section

    International Nuclear Information System (INIS)

    Tellier, Henry.

    1981-04-01

    The capture cross sections of Am 242m can be deduced from resonances analysis at low energy and computed with theoretical models at high energy. In this work, a coherent set of cross sections which reproduced the experimental values of the fission cross sections is computed. These calculations were performed for an energy of the incoming neutron between 1 keV and 1 MeV

  2. INTER, ENDF/B Thermal Cross-Sections, Resonance Integrals, G-Factors Calculation

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: INTER calculates thermal cross sections, g-factors, resonance integrals, fission spectrum averaged cross sections and 14.0 MeV (or other energy) cross sections for major reactions in an ENDF-6 or ENDF-5 format data file. Version 7.01 (Jan 2005): set success flag after return from beginning. 2 - Method of solution: INTER performs integrations by using the trapezoidal rule

  3. Calculations of H+ + Cs→H(2s or 2p) + Cs+ reaction cross sections

    International Nuclear Information System (INIS)

    Valance, A.; Spiess, G.

    1975-01-01

    The H(2s) and H(2p) atom production cross-sections are calculated and compared with experimental results in the incident proton energy range 250-2400eV. The calculation method used involves a perturbation of the stationary molecular states, these adiabatic potentials being obtained from a pseudo-potential describing the core of cesium [fr

  4. MCNP6 Fission Cross Section Calculations at Intermediate and High Energies

    OpenAIRE

    Mashnik, Stepan G.; Sierk, Arnold J.; Prael, Richard E.

    2013-01-01

    MCNP6 has been Validated and Verified (V&V) against intermediate- and high-energy fission cross-section experimental data. An error in the calculation of fission cross sections of 181Ta and a few nearby target nuclei by the CEM03.03 event generator in MCNP6 and a "bug: in the calculation of fission cross sections with the GENXS option of MCNP6 while using the LAQGSM03.03 event generator were detected during our V&V work. After fixing both problems, we find that MCNP6 using CEM03.03 and LAQGSM...

  5. Improved calculation of displacements per atom cross section in solids by gamma and electron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Piñera, Ibrahin, E-mail: ipinera@ceaden.edu.cu [Centro de Aplicaciones Tecnológicas y Desarrollo Nuclear, CEADEN, 30 St. 502, Playa 11300, Havana (Cuba); Cruz, Carlos M.; Leyva, Antonio; Abreu, Yamiel; Cabal, Ana E. [Centro de Aplicaciones Tecnológicas y Desarrollo Nuclear, CEADEN, 30 St. 502, Playa 11300, Havana (Cuba); Espen, Piet Van; Remortel, Nick Van [University of Antwerp, CGB, Groenenborgerlaan 171, 2020 Antwerpen (Belgium)

    2014-11-15

    Highlights: • We present a calculation procedure for dpa cross section in solids under irradiation. • Improvement about 10–90% for the gamma irradiation induced dpa cross section. • Improvement about 5–50% for the electron irradiation induced dpa cross section. • More precise results (20–70%) for thin samples irradiated with electrons. - Abstract: Several authors had estimated the displacements per atom cross sections under different approximations and models, including most of the main gamma- and electron-material interaction processes. These previous works used numerical approximation formulas which are applicable for limited energy ranges. We proposed the Monte Carlo assisted Classical Method (MCCM), which relates the established theories about atom displacements to the electron and positron secondary fluence distributions calculated from the Monte Carlo simulation. In this study the MCCM procedure is adapted in order to estimate the displacements per atom cross sections for gamma and electron irradiation. The results obtained through this procedure are compared with previous theoretical calculations. An improvement in about 10–90% for the gamma irradiation induced dpa cross section is observed in our results on regard to the previous evaluations for the studied incident energies. On the other hand, the dpa cross section values produced by irradiation with electrons are improved by our calculations in about 5–50% when compared with the theoretical approximations. When thin samples are irradiated with electrons, more precise results are obtained through the MCCM (in about 20–70%) with respect to the previous studies.

  6. Heavy Ion SEU Cross Section Calculation Based on Proton Experimental Data, and Vice Versa

    CERN Document Server

    Wrobel, F; Pouget, V; Dilillo, L; Ecoffet, R; Lorfèvre, E; Bezerra, F; Brugger, M; Saigné, F

    2014-01-01

    The aim of this work is to provide a method to calculate single event upset (SEU) cross sections by using experimental data. Valuable tools such as PROFIT and SIMPA already focus on the calculation of the proton cross section by using heavy ions cross-section experiments. However, there is no available tool that calculates heavy ion cross sections based on measured proton cross sections with no knowledge of the technology. We based our approach on the diffusion-collection model with the aim of analyzing the characteristics of transient currents that trigger SEUs. We show that experimental cross sections could be used to characterize the pulses that trigger an SEU. Experimental results allow yet defining an empirical rule to identify the transient current that are responsible for an SEU. Then, the SEU cross section can be calculated for any kind of particle and any energy with no need to know the Spice model of the cell. We applied our method to some technologies (250 nm, 90 nm and 65 nm bulk SRAMs) and we sho...

  7. New techniques for multi-level cross section calculation and fitting

    International Nuclear Information System (INIS)

    Froehner, F.H.

    1981-01-01

    A number of recent developments in multi-level cross section work are described. A new iteration scheme for the conversion of Reich-Moore resonance parameters to Kapur-Peierls parameters allows application of Turing's method for Gaussian broadening of meromorphic functions directly to multi-level cross section expressions, without recourse to the Voigt profiles psi and chi. This makes calculation of Doppler-broadened Reich-Moore and MLBW cross sections practically as fast as SLBW and Adler-Adler cross section calculations involving the Voigt profiles. A convenient distant-level treatment utilizing average resonance parameters is presented. Apart from effectively dealing with edge effects in resonance fitting work it also leads to a simple prescription for the determination of bound levels which reproduce the thermal cross sections correctly. A brief discussion of improved resonance shape fitting techniques is included, with emphasis on the importance of correlated errors and proper use of prior information by application of Bayes' theorem

  8. Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

    International Nuclear Information System (INIS)

    Blideanu, Valentin; Garcia, Mauricio; Joyer, Philippe; Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco; Ortiz, Felix; Sanz, Javier; Sauvan, Patrick

    2011-01-01

    In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

  9. The calculation of nucleus-nucleus interaction cross sections at high energy in the Glauber approach

    International Nuclear Information System (INIS)

    Gal'perin, A.G.; Uzhinskij, V.V.

    1994-01-01

    Total, inelastic and elastic cross sections of nucleus-nucleus (AA)-interactions at high energy (HE) are calculated on the base of Glauber approach. The calculation scheme is realized as a set of routines. The statistical average method is used in calculations. Program runs in an interactive regime. User is prompted about charge and mass numbers of nuclei and NN-interaction characters at the energy he is interested in: total cross section, the slope parameter of differential cross section of elastic scattering and ratio of real part to imaginary part of elastic scattering amplitude at zero momentum transfer. These data can be extracted from proper compilations. Results of calculations are displayed and are written on user defined output file. The program runs on PC. 21 refs., 1 tab

  10. Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

    Energy Technology Data Exchange (ETDEWEB)

    Blideanu, Valentin [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Garcia, Mauricio [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Joyer, Philippe, E-mail: philippe.joyer@cea.fr [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Ortiz, Felix [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Sanz, Javier; Sauvan, Patrick [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain)

    2011-10-01

    In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

  11. Analysis methods of neutrons induced resonances in the transmission experiments by time-of-flight and automation of these methods on IBM 7094 II computer

    International Nuclear Information System (INIS)

    Corge, C.

    1967-01-01

    The neutron induced resonances analysis aims to determine the neutrons characteristics, leading to the excitation energies, de-excitation probabilities by gamma radiation emission, by neutron emission or by fission, their spin, their parity... This document describes the methods developed, or adapted, the calculation schemes and the algorithms implemented to realize such analysis on a computer, from data obtained during time-of-flight experiments on the linear accelerator of Saclay. (A.L.B.)

  12. A database of fragmentation cross section measurements applicable to cosmic ray propagation calculations

    International Nuclear Information System (INIS)

    Crawford, H.J.; Engelage, J.; Jones, F.C.

    1989-08-01

    A database of single particle inclusive fragment production cross section measurements has been established and is accessible over common computer networks. These measurements have been obtained from both published literature and direct communication with experimenters and include cross sections for nuclear beams on H, He, and heavier targets, and for H and He beams on nuclear targets, for energies >30 MeV/nucleon. These cross sections are directly applicable to calculations involving cosmic ray nuclear interactions with matter. The data base includes projectile, target, and fragment specifications, beam energy, cross section with uncertainty, literature reference, and comment code. It is continuously updated to assure accuracy and completeness. Also available are widely used semi-empirical formulations for calculating production cross sections and excitation functions. In this paper we discuss the database in detail and describe how it can be accessed. We compare the measurements with semi-empirical calculations and point out areas where improved calculations and further cross section measurements are required. 5 refs., 2 figs

  13. Cross-Sections for Low-Energy Neutron-Induced Fission; Sections Efficaces de Fission pour des Neutrons de Faible Energie; 0421 0415 0427 0415 041d 0418 042f 0414 0415 041b 0415 041d 0418 042f , 0412 042b 0417 0412 0410 041d 041d 041e 0413 041e 041d 0415 0419 0422 0420 041e 041d 0410 041c 0418 041d 0418 0417 K 041e 0419 042d 041d 0415 0420 0413 0418 0418 ; Secciones Eficaces en la Fision Inducida por Neutrones de Baja Energia

    Energy Technology Data Exchange (ETDEWEB)

    Rae, E. R. [Atomic Energy Research Establishment, Harwell, Didcot, Berks. (United Kingdom)

    1965-07-15

    The cross-sections of the heavy nuclei for neutron-induced fission are of fundamental importance to the technology of nuclear energy production. The manner in which these cross-sections vary with the neutron energy and with the mass and charge of the target nuclei also provides much information on the structure of heavy nuclei. In the case of a thermally fissile target nucleus, as the neutron energy is increased from thermal the cross-section first exhibits an inverse velocity dependence, followed by a region in which sharp resonance peaks appear, and finally a continuum region where the cross-section exhibits relatively smooth steps and breaks. All these phenomena can be explained in principle, but certain features of the data have proved very difficult to explain quantitatively. Recent cross-section measurements, stimulated by the needs of reactor technology, have concentrated on improving the energy resolution and accuracy of the data on fuel materials, and this has led to a more detailed study of the resonance region, which is of considerable interest for reactor Doppler effect calculations. More accurate measurements at rather higher energies have established the existence of appreciable fission cross-sections below the so-called fission threshold in certain cases. Careful measurements of this nature have in turn stimulated interest in the interpretation of the cross-sections in terms of nuclear models on a firm quantitative basis. This paper outlines the main features of neutron-induced-fission cross-sections and their interpretation. Some emphasis is placed on recent improvements in the quality of the measurements and in attempts at quantitative interpretation of certain aspects of the data. (author) [French] Les sections efficaces de fission des noyaux lourds ont une importance essentielle pour la technologie de la production d'energie d'origine nucleaire. La maniere dont ces sections efficaces varient Selon l'energie des neutrons et Selon la masse et la

  14. Calculation of the fine spectrum and integration of the resonance cross sections in the cells

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1986-10-01

    The code BOXER is used for the neutronics calculations of two-dimensional LWR arrays. During the calculation of the group constants of the cells (pin, clad and moderator), the program SLOFIN, a BOXER module, allows taking into account the self-shielding of the resonances. The resonance range is devided into two parts: - above 907 eV the cross sections are condensed into groups by the library code ETOBOX. In SLOFIN, these values are interpolated over the equivalent cross section and the temperature. The interpolation formula chosen gives an accuracy better than 1% for values of the equivalent cross section larger than 5 barns. - between 4 and 907 eV, the cross sections are given in pointwise form as a function of the lethargy. At first a list of pointwise macroscopic cross section is established. Then the fine spectrum in the cell is calculated in 2 or 3 zones by means of the collision probability theory. In the central zone one resonant pseudo-nuclide is considered for the calculation of the scattering source, while the light nuclides are explicitly treated but under the assumption of energy independent cross sections. The fine spectrum is then used as a weihting function for the condensation of the pointwise cross sections of the resonant nuclides into energy groups. The procedure was checked on the basis of the TRX-1 to -4 and BAPL-UO 2 -1 to -3 experiments which are used as benchmarks for the tests of the ENDF/B libraries. The comparisons with other calculation results show that the deviations observed are typical for the basic cross sections. The method proposed shows a good accuracy in the application range foreseen for BOXER. It is also fast enough to be used as a standard method in a cell code. (author)

  15. CREST : a computer program for the calculation of composition dependent self-shielded cross-sections

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1977-01-01

    A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)

  16. Calculation of total and ionization cross sections for electron scattering by primary benzene compounds

    Science.gov (United States)

    Singh, Suvam; Naghma, Rahla; Kaur, Jaspreet; Antony, Bobby

    2016-07-01

    The total and ionization cross sections for electron scattering by benzene, halobenzenes, toluene, aniline, and phenol are reported over a wide energy domain. The multi-scattering centre spherical complex optical potential method has been employed to find the total elastic and inelastic cross sections. The total ionization cross section is estimated from total inelastic cross section using the complex scattering potential-ionization contribution method. In the present article, the first theoretical calculations for electron impact total and ionization cross section have been performed for most of the targets having numerous practical applications. A reasonable agreement is obtained compared to existing experimental observations for all the targets reported here, especially for the total cross section.

  17. Calculation of total and ionization cross sections for electron scattering by primary benzene compounds

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Suvam; Naghma, Rahla; Kaur, Jaspreet; Antony, Bobby, E-mail: bka.ism@gmail.com [Atomic and Molecular Physics Lab, Department of Applied Physics, Indian School of Mines, Dhanbad (India)

    2016-07-21

    The total and ionization cross sections for electron scattering by benzene, halobenzenes, toluene, aniline, and phenol are reported over a wide energy domain. The multi-scattering centre spherical complex optical potential method has been employed to find the total elastic and inelastic cross sections. The total ionization cross section is estimated from total inelastic cross section using the complex scattering potential-ionization contribution method. In the present article, the first theoretical calculations for electron impact total and ionization cross section have been performed for most of the targets having numerous practical applications. A reasonable agreement is obtained compared to existing experimental observations for all the targets reported here, especially for the total cross section.

  18. To the calculation of differential and total cross sections of γπ interactions

    International Nuclear Information System (INIS)

    Duplij, S.A.

    1980-01-01

    The differential and total cross sections of different charge channels of the γπ→ππ process are calculated. At the threshold energies the vector dominance model predicts twice as large values of the total cross sections than the current algebra. In resonance the total cross section of photoproduction on a neutral pion is 10-50 μb, on a charged pion - 5-10μb, at near-threshold energies (Esub(γ)=300-600 MeV) both cross sections are of the 20-40 nb order. For the γπ→ππ process the differential cross sections according to the invariant mass of two pions are obtained for different charge channels. At the threshold energies the total cross sections of the γπ→ππ process is of the 0.1-1 nb order

  19. Neutron induced degradation in nitrided pyrogenic field oxide MOS capacitors

    Science.gov (United States)

    Vaidya, S. J.; Sharma, D. K.; Shaikh, A. M.; Chandorkar, A. N.

    2002-09-01

    Neutron induced oxide charge trapping and generation of interface states in MOS capacitors with pyrogenic and nitrided pyrogenic field oxides have been studied. In order to assess the damage due to neutrons alone, it is necessary to account for the damage produced by the accompanying gamma rays from neutron radiation. This is done by measuring the intensity of gamma radiation accompanying neutrons at different neutron fluences at the irradiation position. MOS capacitor structures were subjected to neutron radiation in a swimming pool type of reactor. Other samples from the same batch were then subjected to an equivalent dose of gamma radiation from a Co 60 source. The difference in the damage observed was used to characterize the damage caused by neutrons. It is observed that neutrons, though uncharged, are capable of causing ionization damage. This damage is found to be significant when the radiation is performed under biased conditions. Nitridation in different ambients is found to improve the radiation performance of pyrogenic field oxides with respect to positive charge build up as well as interface state generation. Pyrogenic oxide nitrided in N 2O is found to be the best oxynitride as damage due to neutrons is the least.

  20. Neutron induced autoradiography of some minerals from the Allchar mine

    International Nuclear Information System (INIS)

    Lazaru, A.; Ilic, R.; Skvarc, J.; Kristof, E.S.; Stafilov, T.

    1999-01-01

    The mineral lorandite from the Allchar mine (Kavadarci, Macedonia) will be used to estimate the average solar neutrino flux. Here, the amount of 205 Pb isotope induced by the 205 Tl(ν e , e - ) 205 Pb reaction is measured and converted to neutrino flux. To determine the few 205 Pb atoms that are produced by solar neutrinos in the Tl ore it is necessary to know all the interfering reactions and/or impurities producing 205 Pb. The concentration and/or spatial distribution of some impurities such as U in lorandite should be known as accurately as possible. In the present work uranium and boron concentrations in some minerals from the Allchar mine (lorandite, realgar, stibnite, orpiment and dolomite) were measured by neutron induced autoradiography. The tracks of 10 B(n, α) and 235 U(n, f) reaction products were recorded by CR-39 and phosphate glass (PSK-50) etched track detectors, respectively. Results showed that uranium is nonuniformly distributed in some of the minerals (orpiment, realgar and lorandite). Average uranium concentration levels varied from 0.01 to 1 μg g -1 . The highest boron concentration (about 6.7 μg g -1 ) was found in stibnite while its concentration in other minerals was below the detection limit (about 1 μg g -1 ) of the technique

  1. Neutron induced electron radiography; Radiografia com eletrons induzida por neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Marcos Leandro Garcia

    2008-07-01

    In the present paper a new radiography technique, the 'Neutron Induced Electron Radiography' - NIER, to inspect low thickness samples on the order of micra, has been developed. This technique makes use of low energy electrons as penetrating radiation generated from metallic gadolinium screens when irradiated by thermal neutrons. The conditions to obtain the best image for the conventional X-ray film Kodak-AA were determined by using a digital system to quantify the darkening level of the film. The irradiations have been performed at a radiography equipment installed at the beam-hole no. 8 of the 5 MW IEA-R1 nuclear research reactor of IPEN-CNEN/SP. The irradiation time to obtain the best radiography was 100 seconds and for such condition the technique was able to discern 1 {mu}m in 24 {mu}m of aluminum at a resolution of 32 {mu}m. By visual comparison the images obtained by the NIER shown a higher quality when compared with the ones from other usual techniques the make use of electrons a penetrating radiation and films for image registration. Furthermore the use of the digital system has provided a smaller time for data acquisition and data analysis as well as an improvement in the image visualization. (author)

  2. Neutron induced autoradiography of some minerals from the Allchar mine

    CERN Document Server

    Lazaru, A; Skvarc, J; Kristof, E S; Stafilov, T

    1999-01-01

    The mineral lorandite from the Allchar mine (Kavadarci, Macedonia) will be used to estimate the average solar neutrino flux. Here, the amount of sup 2 sup 0 sup 5 Pb isotope induced by the sup 2 sup 0 sup 5 Tl(nu sub e , e sup -) sup 2 sup 0 sup 5 Pb reaction is measured and converted to neutrino flux. To determine the few sup 2 sup 0 sup 5 Pb atoms that are produced by solar neutrinos in the Tl ore it is necessary to know all the interfering reactions and/or impurities producing sup 2 sup 0 sup 5 Pb. The concentration and/or spatial distribution of some impurities such as U in lorandite should be known as accurately as possible. In the present work uranium and boron concentrations in some minerals from the Allchar mine (lorandite, realgar, stibnite, orpiment and dolomite) were measured by neutron induced autoradiography. The tracks of sup 1 sup 0 B(n, alpha) and sup 2 sup 3 sup 5 U(n, f) reaction products were recorded by CR-39 and phosphate glass (PSK-50) etched track detectors, respectively. Results showed...

  3. The Upgrade of the Neutron Induced Positron Source NEPOMUC

    Science.gov (United States)

    Hugenschmidt, C.; Ceeh, H.; Gigl, T.; Lippert, F.; Piochacz, C.; Pikart, P.; Reiner, M.; Weber, J.; Zimnik, S.

    2013-06-01

    In summer 2012, the new NEutron induced POsitron Source MUniCh (NEPOMUC) was installed and put into operation at the research reactor FRM II. At NEPOMUC upgrade 80% 113Cd enriched Cd is used as neutron-gamma converter in order to ensure an operation time of 25 years. A structure of Pt foils inside the beam tube generates positrons by pair production. Moderated positrons leaving the Pt front foil are electrically extracted and magnetically guided to the outside of the reactor pool. The whole design, including Pt-foils, the electric lenses and the magnetic fields, has been improved in order to enhance both the intensity and the brightness of the positron beam. After adjusting the potentials and the magnetic guide and compensation fields an intensity of about 3·109 moderated positrons per second is expected. During the first start-up, the measured temperatures of about 90°C ensure a reliable operation of the positron source. Within this contribution the features and the status of NEPOMUC upgrade are elucidated. In addition, an overview of recent positron beam experiments and current developments at the spectrometers is given.

  4. A new approach to make collapsed cross section for burnup calculation of subcritical system

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Kondo, Keitaro; Miyamaru, Hiroyuki; Murata, Isao

    2008-01-01

    A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)

  5. Analysis of fusion neutronics calculations and appraisal of UW cross-section library

    International Nuclear Information System (INIS)

    Xie Jianping; Li Xingzhong; Ying Chuntong

    1989-01-01

    A series of calculations for different cases (especially for the values of tritium breeding ratio T, and the fuel breeding ratio F in the blanket of a hybrid reactor) were carried out by using ANISN program and UW cross-section library. The comparison with other results in China and abroad kalso was done. It was shownwn that the installation and execution of ANISN program on ELXSI machine at Tsinghua University are successful, and the UW cross-section library is reliable. It may be used for fusion neutronics calculation in the future. The paper also points out that the difference between the calculations and by the authors are due to jthe different in cross-section data used

  6. FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations

    International Nuclear Information System (INIS)

    Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.

    1980-05-01

    Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described

  7. Polynomial expansion methodology for microscopic cross sections to use in spatial burnup calculations

    International Nuclear Information System (INIS)

    Conti Filho, P.; Oliveira Barroso, A.C. de

    1985-01-01

    It was developed a computer code to generate polynomial coefficients which represent homogenized microscopic cross sections in function of the local accumulated burnup and concentration of soluble boron, presented in fuel element, for each step of burnup reactor. Afterward, it was developed a coupling between LEOPARD-GERADOR DE POLINOMIOS - CITATION computer codes to interpret and build homogenized microscopic cross sections according with local characteristics of each fuel element during the burnup calculation of reactor core. (M.C.K.) [pt

  8. A computer code for calculating neutron cross-sections from resonance parameter data

    International Nuclear Information System (INIS)

    Mill, A.J.

    1979-08-01

    A computer code, XSEC, has been written which calculates neutron cross-sections from resonance data. Although the program was originally written in order to identify neutron 'windows' in enriched nuclides, it may be used to evaluate the total neutron cross-section of any medium mass nuclide at intermediate energies. XSEC has proved very useful in identifying suitable nuclides for use as neutron filters at intermediate energies. (author)

  9. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  10. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H.

    2005-01-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  11. Basis calculation of phase cross section library in a low power fast reactor neutronic simulation

    International Nuclear Information System (INIS)

    Jachic, J.

    1993-09-01

    In order to implement the utilization of the efficient multidimensional cubic SPLINE interpolation, we determine the phase library bases for net like relevant state components. A generic cubic surface and a weighted plane pertinent alternative interpolating methods used capable to generate cross sections values for fixed coordinates from cell code calculated data points is used. It is verified that the phase library bases increases or decrease smoothly and monotonically with the spectrum asymmetry and total flux buckling. This justifies its use in cross section updating avoiding cell calculations. (author)

  12. Calculation of the extinction cross section and lifetime of a gold nanoparticle using FDTD simulations

    International Nuclear Information System (INIS)

    Radhakrishnan, Archana; Murugesan, Dr V.

    2014-01-01

    The electromagnetic theory of light explains the behavior of light in most of the domains quite accurately. The problem arises when the exact solution of the Maxwell's equation is not present, in case of objects with arbitrary geometry. To find the extinction cross-section and lifetime of the gold nanoparticle, the software FDTD solutions 8.6 by Lumerical is employed. The extinction cross-sections and lifetimes of Gold nanospheres of different sizes and arrangements are studied using pulse lengths of the order of femtoseconds. The decay constant and other properties are compared. Further, the lifetimes are calculated using frequency and time domain calculations

  13. Calculation of the extinction cross section and lifetime of a gold nanoparticle using FDTD simulations

    Energy Technology Data Exchange (ETDEWEB)

    Radhakrishnan, Archana, E-mail: anju.archana@gmail.com [B.Tech, Engineering Physics, National Institute Of Technology, Calicut (India); Murugesan, Dr V., E-mail: murugesh@serc.iisc.in [Assistant Professor, Supercomputer Education and Research Centre, Indian Institute of Science, Bangalore (India)

    2014-10-15

    The electromagnetic theory of light explains the behavior of light in most of the domains quite accurately. The problem arises when the exact solution of the Maxwell's equation is not present, in case of objects with arbitrary geometry. To find the extinction cross-section and lifetime of the gold nanoparticle, the software FDTD solutions 8.6 by Lumerical is employed. The extinction cross-sections and lifetimes of Gold nanospheres of different sizes and arrangements are studied using pulse lengths of the order of femtoseconds. The decay constant and other properties are compared. Further, the lifetimes are calculated using frequency and time domain calculations.

  14. Relativistic total and differential cross section proton--proton electron--positron pair production calculation

    International Nuclear Information System (INIS)

    Rubinstein, J.E.

    1976-01-01

    Circle Feynman diagrams for a specific permutation of variables along with their corresponding algebraic expressions are presented to evaluate [H] 2 for proton-proton electron-positron pair production. A Monte Carlo integration technique is introduced and is used to set up the multiple integral expression for the total pair production cross section. The technique is first applied to the Compton scattering problem and then to an arbitrary multiple integral. The relativistic total cross section for proton-proton electron-positron pair production was calculated for eight different values of incident proton energy. A variety of differential cross sections were calculated for the above energies. Angular differential cross section distributions are presented for the electron, positron, and proton. Invariant mass differential cross section distributions are done both with and without the presence of [H] 2 . Both WGHT and log 10 (TOTAL) distributions were also obtained. The general behavioral trends of the total and differential cross sections for proton-proton electron-positron pair production are presented. The range of validity for this calculation is from 0 to about 200 MeV

  15. Molecular-state cross-section calculations for H+Csarrow-right-leftH-+Cs+

    International Nuclear Information System (INIS)

    Olson, R.E.; Kimura, M.; Sato, H.

    1984-01-01

    Pseudopotential molecular-structure calculations have been used to obtain the seven lowest 1 Σ and 1 Pi states of CsH. These states and their associated radial and rotational coupling terms have been used to calculate the cross sections for H - +Cs + ion-pair production in H+Cs(6s) and H+Cs((6p) collisions at energies from 0.1 to 10 keV. The ion-ion mutual neutralization cross section, H - +Cs + →H+Cs, is also presented. The cross-section calculations were done with the perturbed-stationary-state method, modified to include two-electron translation factors. The ion-pair production cross section for ground-state reactants is in good agreement with experiment; collisions of H with excited Cs((6p) show an order-of-magnitude enhancement of the ion-pair production cross section at 100 eV. The ion-ion mutual neutralization cross section is found to be large, attaining a value of 1.3 x 10 -14 cm 2 at 0.1 keV

  16. GNASH: a preequilibrium, statistical nuclear-model code for calculation of cross sections and emission spectra

    International Nuclear Information System (INIS)

    Young, P.G.; Arthur, E.D.

    1977-11-01

    A new multistep Hauser--Feshbach code that includes corrections for preequilibrium effects is described. The code can calculate up to 60 decay reactions (cross sections and energy spectra) in one computation, and thereby provide considerable flexibility for handling processes with complicated reaction chains. Input parameter setup, problem output, and subroutine descriptions are given along with a sample problem calculation. A brief theoretical description is also included. 8 figures, 3 tables

  17. The PSIMECX medium-energy neutron activation cross-section library. Part II: Calculational methods for light to medium mass nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F.

    1998-09-01

    The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: {sup 12}C, {sup 13}C, {sup 16}O, {sup 17}O, {sup 18}O, {sup 23}Na, {sup 24}Mg, {sup 25}Mg, {sup 26}Mg, {sup 27}Al, {sup 28}Si, {sup 29}Si, {sup 30}Si, {sup 31}P, {sup 32}S, {sup 33}S, {sup 34}S, {sup 36}S, {sup 35}Cl, {sup 37}Cl, {sup 39}K, {sup 40}K, {sup 41}K, {sup 40}Ca, {sup 42}Ca, {sup 43}Ca, {sup 44}Ca, {sup 46}Ca, {sup 48}Ca, {sup 46}Ti, {sup 47}Ti, {sup 48}Ti, {sup 49}Ti, {sup 50}Ti, {sup 50}V, {sup 51}V, {sup 50}Cr, {sup 52}Cr, {sup 53}Cr, {sup 54}Cr, {sup 55}Mn, {sup 54}Fe, {sup 56}Fe, {sup 57}Fe, {sup 58}Fe, {sup 58}Ni, {sup 60}Ni, {sup 61}Ni, {sup 62}Ni, {sup 64}Ni, {sup 63}Cu, {sup 65}Cu, {sup 64}Zn, {sup 66}Zn, {sup 67}Zn, {sup 68}Zn, {sup 70}Zn, {sup 92}Mo, {sup 94}Mo, {sup 95}Mo, {sup 96}Mo, {sup 97}Mo, {sup 98}Mo, {sup 100}Mo, {sup 121}Sb, {sup 123}Sb, {sup 204}Pb, {sup 206}Pb, {sup 207}Pb, {sup 208}Pb, {sup 232}Th and {sup 238}U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are principal constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This second report, of a series of three, describes and discusses the calculational methods used for the stable isotopes up to and including {sup 123}Sb. The library itself has been described in the first report of the series and the treatment for the heavy nuclei is given in the third. (author)

  18. A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Marullo, G C

    1971-02-15

    In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

  19. Calculation of total cross sections for electron and positron scattering on sodium and potassium

    International Nuclear Information System (INIS)

    McCarthy, I.E.; Ratnavelu, K.; Zhou, Y.

    1993-02-01

    Total cross sections for electron and positron scattering on sodium and potassium are calculated at various energies and compared with experiment. The method use is the coupled-channels-optical method with the equivalent-local polarisation potential, which takes all channels into account. For electrons the calculations are checked by comparison with coupled-channels-optical calculations using a detailed polarisation potential that makes only one approximation, that of weak coupling in the ionisation space. The polarisation potential for positrons includes effects of ionisation and positronium formation. 13 refs., 2 tabs

  20. Calculational tools for the evaluation of nuclear cross-section and spectra data

    International Nuclear Information System (INIS)

    Gardner, M.A.

    1985-01-01

    A technique based on discrete energy levels rather than energy level densities is presented for nuclear reaction calculations. The validity of the technique is demonstrated via theoretical and experimental agreement for cross sections, isomer-ratios and gamma-ray strength functions. 50 refs., 7 figs

  1. Adjust of effective cross sections of some actinides in inventory calculation with HAMOR-2

    International Nuclear Information System (INIS)

    Guimaraes, L.N.F.; Marzo, M.A.S.

    1985-01-01

    A comparative study of the adjustment of effective cross sections generated by HAMOR-2 for the following actinides U-238, Pu-239 and Pu-240 is done. The adjustment were made to calculate the inventory of two different PWRs reactors. (M.C.K.) [pt

  2. First step in the process of calculating the cross section for muonic antihydrogen

    CERN Document Server

    Maher, Niamh

    2016-01-01

    The end goal of the project is to measure the charge radius of the antiproton with muons. However a necessary step first is to calculate cross section of Muonium and antiprotons in the production of antihydrogen to determine the feasibility of such an experiment.

  3. Comparison Between Calculated and Measured Cross Section Changes in Natural Uranium Irradiated in NRX

    Energy Technology Data Exchange (ETDEWEB)

    Ahlstroem, P E

    1961-03-15

    It is desirable to obtain an experimental check of the reliability of the methods currently used to determine reactivity changes in a reactor and, with a view to meeting this requirement to some extent, a preliminary comparison has been made between calculated and measured cross-section changes in rods of natural uranium irradiated in NRX. The measurements were made at Harwell in the GLEEP reactor and a description has been given by, inter alia, Ward and Craig. The theory of the calculations, which is briefly described in this report, has been indicated by Littler. The investigation showed that the methods for calculating burn up used at present provides a good illustration of the long-term variations in isotope contents. A satisfactory agreement is obtained with experimental results when calculating apparent cross-section changes in uranium rods due to irradiation if the fission cross- section for {sup 239}Pu is set to 780 b. This is 34 b higher than the figure quoted in BNL - 325 (1958). However, in order to get a good idea as to whether the calculated long-term variations in reactivity really correspond to reality, it is necessary to make further investigations. For this reason the results quoted in this report should be regarded as preliminary.

  4. Construction of PWR nuclear cross sections for transient calculations. Test of the ANTI program against TWODIM

    International Nuclear Information System (INIS)

    Thorlaksen, B.

    1981-05-01

    Nuclear cross sections for fuel assemblies of the more recent Westinghouse designs, representing two different PWR reactor cores, are calculated as functions of average fuel temperature, moderator density, and moderator poison concentration. The cross-section functions are verified by referring to Westinghouse power-shape calculations and other analysis. Computations on the side reflector resulted in significantly higher albedo values than used previously for BWR's in similar nodal codes. This led to an investigation of the influence of the internodal coupling coefficients on the power shape. It is concluded that the calculated power shape is strongly dependent, on the choise of coupling coefficients. However, it is shown that ''the correct'' set of coupling coefficients depends mostly on the nodal configuration, and that it is fairly independent of the power condition. (author)

  5. MC2-2: a code to calculate fast neutron spectra and multigroup cross sections

    International Nuclear Information System (INIS)

    Henryson, H. II; Toppel, B.J.; Stenberg, C.G.

    1976-06-01

    MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers

  6. ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

    International Nuclear Information System (INIS)

    1985-01-01

    1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

  7. Filtered thermal neutron captured cross sections measurements and decay heat calculations

    International Nuclear Information System (INIS)

    Pham Ngoc Son; Vuong Huu Tan

    2015-01-01

    Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R ed ) of 420 and neutron flux (Φ th ) of 1.6*10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross sections for nuclide of 51 V, by the activation method relative to the standard reaction 197 Au(n,γ) 198 Au. In addition to the activities of neutron capture cross sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U are introduced in this report. (author)

  8. Yield calculations for a facility for short-lived nuclear beams

    CERN Document Server

    Jiang, C L; Gomes, I; Heinz, A M; Nolen, Jerry A; Rehm, K E; Savard, G; Schiffer, J P

    2002-01-01

    Yields for a broad range of radioactive nuclei produced by spallation reactions, neutron-induced fission, in-flight projectile fragmentation and in-flight fission have been calculated for beams of stable nuclei at energies of 100-1000 MeV/u. Calculations of cross-sections and yields, attenuation effects due to absorption, production from secondary reactions, and transport efficiencies for mass selection are discussed. Rare isotope yields as functions of bombarding energies for both reaccelerated and directly produced fast-fragmentation beams are presented. This information provides a foundation for a cost-effective design of a next generation rare isotope accelerator.

  9. CWF and TABLE - Two Fortran programmes for the calculation of Coulomb penetration and shift factors

    International Nuclear Information System (INIS)

    Norton, D.S.; James, M.F.

    1965-12-01

    CWF and TABLE are Fortran programmes, written for the IBM 7090 and English-Electric Leo Marconi KDF9 computers, that calculate the penetration and shift factors for a charged particle in a Coulomb field. The numerical methods used are those of Lutz and Karvelis. The two programmes are very similar. Input to TABLE is in the form of the centre-of-mass co-ordinates. CWF is intended for use in calculating cross-sections for neutron-induced reactions which result in charged particle emission, and the input is in the form of the neutron energy in the laboratory frame of reference, together with other necessary reaction data. (author)

  10. Evaluation of Neutron Induced Reactions for 32 Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Il

    2007-02-15

    Neutron cross sections for 32 fission products were evaluated in the neutron-incident energy range from 10{sup -5} eV to 20 MeV. The list of fission products consists of the priority materials for several applications, extended to cover complete isotopic chains for three elements. The full list includes 8 individual isotopes, {sup 95}Mo, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, and 24 isotopes in complete isotopic chains for Nd (8), Sm (9) and Dy (7). Our evaluation methodology covers both the low energy region and the fast neutron region.In the low energy region, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. This resource was used to infer both the thermal values and the resolved resonance parameters that were validated against the capture resonance integrals. In the unresolved resonance region we performed the additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data.In the fast neutron region our evaluations are based on the nuclear reaction model code EMPIRE-2.19 validated against the experimental data. EMPIRE is the modular system of codes consisting of many nuclear reaction models, including the spherical and deformed Optical Model, Hauser-Feshbach theory with the width fluctuation correction and complete gamma-ray emission cascade, DWBA, Multi-step Direct and Multi-step Compound models, and several versions of the phenomenological preequilibrium models. The code is equipped with a power full GUI, allowing an easy access to support libraries such as RIPL and CSISRS, the graphical package, as well the utility codes for formatting and checking. In general, in our calculations we used the Reference Input Parameter Library, RIPL, for the initial set model parameters. These parameters were properly adjusted to reproduce the available experimental data taken from the CSISRS library. Our evaluations cover cross

  11. Evaluation of Neutron Induced Reactions for 32 Fission Products

    International Nuclear Information System (INIS)

    Kim, Hyeong Il

    2007-02-01

    Neutron cross sections for 32 fission products were evaluated in the neutron-incident energy range from 10 -5 eV to 20 MeV. The list of fission products consists of the priority materials for several applications, extended to cover complete isotopic chains for three elements. The full list includes 8 individual isotopes, 95 Mo, 101 Ru, 103 Rh, 105 Pd, 109 Ag, 131 Xe, 133 Cs, 141 Pr, and 24 isotopes in complete isotopic chains for Nd (8), Sm (9) and Dy (7). Our evaluation methodology covers both the low energy region and the fast neutron region.In the low energy region, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. This resource was used to infer both the thermal values and the resolved resonance parameters that were validated against the capture resonance integrals. In the unresolved resonance region we performed the additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data.In the fast neutron region our evaluations are based on the nuclear reaction model code EMPIRE-2.19 validated against the experimental data. EMPIRE is the modular system of codes consisting of many nuclear reaction models, including the spherical and deformed Optical Model, Hauser-Feshbach theory with the width fluctuation correction and complete gamma-ray emission cascade, DWBA, Multi-step Direct and Multi-step Compound models, and several versions of the phenomenological preequilibrium models. The code is equipped with a power full GUI, allowing an easy access to support libraries such as RIPL and CSISRS, the graphical package, as well the utility codes for formatting and checking. In general, in our calculations we used the Reference Input Parameter Library, RIPL, for the initial set model parameters. These parameters were properly adjusted to reproduce the available experimental data taken from the CSISRS library. Our evaluations cover cross sections for almost all reaction channels

  12. Calculation of neutron cross sections on iron up to 40 MeV

    International Nuclear Information System (INIS)

    Arthur, E.D.; Young, P.G.

    1980-01-01

    The development of high energy d + Li neutron sources for fusion materials radiation damage studies will require neutron cross sections up to 40 MeV. Experimental data above 15 MeV are generally sparse or nonexistent, and reliance must be placed upon nuclear-model calculations to produce the needed cross sections. To satisfy such requirements for the Fusion Materials Irradiation Test Facility (FMIT), neutron cross sections have been calculated for 54 56 Fe between 3 and 40 MeV. These results were joined to the existing ENDF/B-V evaluation below 3 MeV. In this energy range, most neutron reactions can be described using the Hauser-Feshbach statistical model with corrections for preequilibrium and direct-reaction effects. To properly use these models to obtain realistic cross sections, emphasis must be placed upon the determination of suitable input parameters (optical model sets, gamma-ray strength functions, level densities) valid over the energy range of the calculation. To do this, several types of independent data were used to arrive at consistent parameter sets as described

  13. Prediction of fission mass-yield distributions based on cross section calculations

    International Nuclear Information System (INIS)

    Hambsch, F.-J.; G.Vladuca; Tudora, Anabella; Oberstedt, S.; Ruskov, I.

    2005-01-01

    For the first time, fission mass-yield distributions have been predicted based on an extended statistical model for fission cross section calculations. In this model, the concept of the multi-modality of the fission process has been incorporated. The three most dominant fission modes, the two asymmetric standard I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode are taken into account. De-convoluted fission cross sections for S1, S2 and SL modes for 235,238 U(n, f) and 237 Np(n, f), based on experimental branching ratios, were calculated for the first time in the incident neutron energy range from 0.01 to 5.5 MeV providing good agreement with the experimental fission cross section data. The branching ratios obtained from the modal fission cross section calculations have been used to deduce the corresponding fission yield distributions, including mean values also for incident neutron energies hitherto not accessible to experiment

  14. Achievement and qualification of multigroup cross-section library for light water reactor calculation

    International Nuclear Information System (INIS)

    Gastaldi, B.

    1986-07-01

    This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr

  15. New techniques for multi-level cross section calculation and fitting

    International Nuclear Information System (INIS)

    Froehner, F.H.

    1980-09-01

    A number of recent developments in multi-level cross section work are described. A new iteration scheme for the conversion of Reich-Moore resonance parameters to Kapur-Peierls parameters allows application of Turing's method for Gaussian broadening of meromorphic functions directly to multi-level cross section expressions, without recourse to the Voigt profiles psi and chi. This makes calculation of Doppler-broadened Reich-Moore and MLBW cross sections practically as fast as SLBW and Adler-Adler cross section calculations involving the Voigt profiles. A convenient distant-level treatment utilizing average resonance parameters is presented. Apart from effectively dealing with edge effects in resonance fitting work it also leads to a simple prescription for the determination of bound levels which reproduce the thermal cross sections correctly. A brief discussion of improved resonance shape fitting techniques is included, with empahsis on the importance of correlated errors and proper use of prior information by application of Bayes' theorem. (orig.) [de

  16. Covariance matrices for nuclear cross sections derived from nuclear model calculations

    International Nuclear Information System (INIS)

    Smith, D. L.

    2005-01-01

    The growing need for covariance information to accompany the evaluated cross section data libraries utilized in contemporary nuclear applications is spurring the development of new methods to provide this information. Many of the current general purpose libraries of evaluated nuclear data used in applications are derived either almost entirely from nuclear model calculations or from nuclear model calculations benchmarked by available experimental data. Consequently, a consistent method for generating covariance information under these circumstances is required. This report discusses a new approach to producing covariance matrices for cross sections calculated using nuclear models. The present method involves establishing uncertainty information for the underlying parameters of nuclear models used in the calculations and then propagating these uncertainties through to the derived cross sections and related nuclear quantities by means of a Monte Carlo technique rather than the more conventional matrix error propagation approach used in some alternative methods. The formalism to be used in such analyses is discussed in this report along with various issues and caveats that need to be considered in order to proceed with a practical implementation of the methodology

  17. HATHOR. HAdronic Top and Heavy quarks crOss section calculatoR

    International Nuclear Information System (INIS)

    Aliev, M.; Lacker, H.; Langenfeld, U.; Uwer, P.; Wiedermann, M.

    2010-07-01

    We present a program to calculate the total cross section for top-quark pair production in hadronic collisions. The program takes into account recent theoretical developments such as approximate next-to-next-to-leading order perturbative QCD corrections and it allows for studies of the theoretical uncertainty by separate variations of the factorization and renormalization scales. In addition it offers the possibility to obtain the cross section as a function of the running top-quark mass. The program can also be applied to a hypothetical fourth quark family provided the QCD couplings are standard. (orig.)

  18. HATHOR. HAdronic Top and Heavy quarks crOss section calculatoR

    Energy Technology Data Exchange (ETDEWEB)

    Aliev, M.; Lacker, H.; Langenfeld, U.; Uwer, P.; Wiedermann, M. [Berlin Univ. (Germany). Inst. fuer Physik; Moch, S. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany)

    2010-07-15

    We present a program to calculate the total cross section for top-quark pair production in hadronic collisions. The program takes into account recent theoretical developments such as approximate next-to-next-to-leading order perturbative QCD corrections and it allows for studies of the theoretical uncertainty by separate variations of the factorization and renormalization scales. In addition it offers the possibility to obtain the cross section as a function of the running top-quark mass. The program can also be applied to a hypothetical fourth quark family provided the QCD couplings are standard. (orig.)

  19. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  20. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  1. Influence of nuclear cross section data at efficiency calculation of the 3He semiconductor neutron spectrometer

    International Nuclear Information System (INIS)

    Avdic, S.; Pesic, M.

    1992-01-01

    The ORTEC 580 Neutron Spectrometer system contains a detector unit in diode coincidence arrangement for measurement of fast neutron spectrum in the energy range from 1 MeV to 14 MeV. Numerical code HE3 for computation of semiconductor 3 He detector efficiency in a collimated neutron beam is based on analytical method in infinite diode approximation and Monte Carlo method for real spectrometer geometry. Calculations are performed in the first collision approximation in the detector active volume including evaluation of correction factors. Accuracy of relative detector efficiency calculation is improved by using neutron cross section from nuclear library ENDF/B-6. (author)

  2. Maxwellian-averaged cross sections calculated from JENDL-3.2

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-01

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

  3. A general algorithm for calculating jet cross sections in NLO QCD

    CERN Document Server

    Catani, S.; Catani, Stefano; Seymour, Michael H

    1997-01-01

    We present a new general algorithm for calculating arbitrary jet cross sections in arbitrary scattering processes to next-to-leading accuracy in perturbative QCD. The algorithm is based on the subtraction method. The key ingredients are new factorization formulae, called dipole formulae, which implement in a Lorentz covariant way both the usual soft and collinear approximations, smoothly interpolating the two. The corresponding dipole phase space obeys exact factorization, so that the dipole contributions to the cross section can be exactly integrated analytically over the whole of phase space. We obtain explicit analytic results for any jet observable in any scattering or fragmentation process in lepton, lepton-hadron or hadron-hadron collisions. All the analytical formulae necessary to construct a numerical program for next-to-leading order QCD calculations are provided. The algorithm is straightforwardly implementable in general purpose Monte Carlo programs.

  4. Validation of new 240Pu cross section and covariance data via criticality calculation

    International Nuclear Information System (INIS)

    Kim, Do Heon; Gil, Choong-Sup; Kim, Hyeong Il; Lee, Young-Ouk; Leal, Luiz C.; Dunn, Michael E.

    2011-01-01

    Recent collaboration between KAERI and ORNL has completed an evaluation for 240 Pu neutron cross section with covariance data. The new 240 Pu cross section data has been validated through 28 criticality safety benchmark problems taken from the ICSBEP and/or CSEWG specifications with MCNP calculations. The calculation results based on the new evaluation have been compared with those based on recent evaluations such as ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. In addition, the new 240 Pu covariance data has been tested for some criticality benchmarks via the DANTSYS/SUSD3D-based nuclear data sensitivity and uncertainty analysis of k eff . The k eff uncertainty estimates by the new covariance data has been compared with those by JENDL-4.0, JENDL-3.3, and Low-Fidelity covariance data. (author)

  5. K- and L-shell ionization cross sections for deuterons calculated in the ECPSSR theory

    International Nuclear Information System (INIS)

    Cohen, D.D.

    1989-01-01

    Ionization cross sections for K and L subshells are tabulated according to target atomic number and incident deuteron energy. Deuteron energies between 100 keV and 10 MeV and selected targets between C and Am for the K shell and between Ar and Am for the L subshells are used. The cross sections have been calculated in the plane-wave Born approximation (PWBA) with corrections for energy loss (E), Coulomb deflection (C), perturbed stationary states (PSS), and relativistic (R) effects (ECPSSR). Differences between the computational approach of Cohen and Harrigan and that of Brandt and Lapicki are delineated, and the ratios of the resulting cross sections are tabulated. Copyright 1989 Academic Press, Inc

  6. Effect of XCOM photoelectric cross-sections on dosimetric quantities calculated with EGSnrc

    International Nuclear Information System (INIS)

    Hobeila, F.; Seuntjens, J.P.

    2002-01-01

    The EGSnrc Monte-Carlo code system incorporates improved low energy photon physics such as atomic relaxations and the implementation of bound Compton cross-sections using the impulse approximation. The total cross-section for photoelectric absorption however, still relies on the data by Storm and Israel (S and I). Yet, low energy applications such as brachytherapy (e.g. 125 I) require up-to-date low-energy photoelectric cross-section data. In this paper, we study the dosimetric effects of a simple implementation of NIST XCOM-based photoelectric cross-sections in EGSnrc. This is done by calculating mass energy-absorption coefficients, absorbed dose from point sources, kilovoltage x-ray beams and ion chamber response. In the EGS code system, the PEGS4 routine reads the photoelectric and pair cross-sections for elements from a file (pgspepr.dat) and provides numerical fits for compounds which will be used by EGSnrc. We updated the photoelectric cross-sections of the pgspepr.dat file with the XCOM total photoelectric absorption cross-sections from NIST. After validation of this new implementation, we studied its effects on a number of dosimetrically relevant quantities. Firstly, we calculated mass energy-absorption coefficients by scoring energy transferred in a thin slab of water and air using the DOSRZnrc user code. Secondly, we calculated inverse-square corrected absorbed dose distributions from point sources in water by using an internally developed user code, KERNELph. Thirdly, we studied the differences in free-air ion chamber response calculations. Ion chamber response is defined as the dose to the cavity of an ionization chamber, D gas , positioned with its effective point of measurement at a reference point divided by air-kerma measured free-in-air at the same point. The ion chamber response was calculated using monoenergetic photon beams of energy 10 keV to 200 keV. The comparison of the Storm and Israel photoelectric cross-sections with the XCOM cross

  7. Effects of shape differences in the level densities of three formalisms on calculated cross-sections

    International Nuclear Information System (INIS)

    Fu, C.Y.; Larson, D.C.

    1998-01-01

    Effects of shape differences in the level densities of three formalisms on calculated cross-sections and particle emission spectra are described. Reactions for incident neutrons up to 20 MeV on 58 Ni are chosen for illustrations. Level density parameters for one of the formalisms are determined from the available neutron resonance data for one residual nuclide in the binary channels and from fitting the measured (n,n'), (n,p) and (n,α) cross-sections for the other two residual nuclides. Level density parameters for the other two formalisms are determined such that they yield the same values as the above one at two selected energies. This procedure forces the level densities from the three formalisms used for the binary pat of the calculation to be as close as possible. The remaining differences are in their energy dependences (shapes). It is shown that these shape differences alone are enough to cause the calculated cross-sections and particle emission spectra to be different by up to 60%. (author)

  8. Neutron-induced 2.2 MeV background in gamma ray telescopes

    International Nuclear Information System (INIS)

    Zanrosso, E.M.; Long, J.L.; Zych, A.D.; White, R.S.; Hughes Aircraft Co., Los Angeles, CA)

    1985-01-01

    Neutron-induced gamma ray production is an important source of background in Compton scatter gamma ray telescopes where organic scintillator material is used. Most important is deuteron formation when atmospheric albedo and locally produced neutrons are thermalized and subsequently absorbed in the hydrogenous material. The resulting 2.2 MeV gamma line essentially represents a continuous isotropic source within the scintillator itself. Interestingly, using a scintillator material with a high hydrogen-to-carbon ratio to minimize the neutron-induced 4.4 MeV carbon line favors the np reaction. The full problem of neutron-induced background in Compton scatter telescopes has been previously discussed. Results are presented of observations with the University of California balloon-borne Compton scatter telescope where the 2.2 MeV induced line emission is prominently seen

  9. Evaluation Procedures of Random Uncertainties in Theoretical Calculations of Cross Sections and Rate Coefficients

    International Nuclear Information System (INIS)

    Kokoouline, V.; Richardson, W.

    2014-01-01

    Uncertainties in theoretical calculations may include: • Systematic uncertainty: Due to applicability limits of the chosen model. • Random: Within a model, uncertainties of model parameters result in uncertainties of final results (such as cross sections). • If uncertainties of experimental and theoretical data are known, for the purpose of data evaluation (to produce recommended data), one should combine two data sets to produce the best guess data with the smallest possible uncertainty. In many situations, it is possible to assess the accuracy of theoretical calculations because theoretical models usually rely on parameters that are uncertain, but not completely random, i.e. the uncertainties of the parameters of the models are approximately known. If there are one or several such parameters with corresponding uncertainties, even if some or all parameters are correlated, the above approach gives a conceptually simple way to calculate uncertainties of final cross sections (uncertainty propagation). Numerically, the statistical approach to the uncertainty propagation could be computationally expensive. However, in situations, where uncertainties are considered to be as important as the actual cross sections (for data validation or benchmark calculations, for example), such a numerical effort is justified. Having data from different sources (say, from theory and experiment), a systematic statistical approach allows one to compare the data and produce “unbiased” evaluated data with improved uncertainties, if uncertainties of initial data from different sources are available. Without uncertainties, the data evaluation/validation becomes impossible. This is the reason why theoreticians should assess the accuracy of their calculations in one way or another. A statistical and systematic approach, similar to the described above, is preferable.

  10. Neutron Elastic Scattering Cross Sections Experimental Data and Optical Model Cross Section Calculations. A Compilation of Neutron Data from the Studsvik Neutron Physics Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Holmqvist, B; Wiedling, T

    1969-06-15

    Neutron elastic scattering cross section measurements have been going on for a long period at the Studsvik Van de Graaff laboratory. The cross sections of a range of elements have been investigated in the energy interval 1.5 to 8 MeV. The experimental data have been compared with cross sections calculated with the optical model when using a local nuclear potential.

  11. Analysis and reevaluation of the neutron cross sections for 23Na

    International Nuclear Information System (INIS)

    Trykov, E.L.; Svinin, I.R.

    2000-05-01

    The reaction model calculations of the cross sections of neutron-induced reactions on 23 Na have been carried out for incident energies up to 20 MeV. The results of the calculations are compared to all available experimental data, including the most recent ones, and also to the previous evaluations. The discrepancies between the data and the present evaluation and also between evaluations themselves were analyzed. The probable reasons of these discrepancies were considered. On the whole, the calculation results agree well enough with the experimental data. (author)

  12. Calculation and Evaluation of Fission Yields and Capture Cross Sections Leading to the Production of Therapeutic Radionuclide by Means of Nuclear Reactors

    International Nuclear Information System (INIS)

    Sublet, J.C.

    2009-01-01

    Much progress has been made in nuclear medicine that involves the use of radionuclides for both diagnosis and therapy. Because of this qualitative and quantitative growth, the adoption of a set of established radionuclides for various applications, the methods of nuclide production need to be addressed and consideration given to other, emerging radionuclides that are judged to be developing in importance. The methods involved are characterized by the transmutation of isotopes by neutron-induced reactions and decays. Therefore, newly evaluated cross sections, fission yields and decay characteristics of relevance to the reactor production of those therapeutic radionuclides have been reviewed. Considerations of the decay schemes of all the nuclides involved are also included. (author)

  13. Cross sections needed for investigations into track phenomena and Monte-Carlo calculations

    International Nuclear Information System (INIS)

    Paretzke, H.G.

    1983-01-01

    Investigations into basic radiation action mechanisms as well as into applied radiation transport problems (e.g. electron microscopy) greatly benefit from detailed computer simulations of charged particle track structures in matter. The first and in fact most important and most difficult step in any such calculation is the derivation of reliable cross sections for the most relevant interaction processes in the material(s) under consideration. The second step in radiation transport calculations is the testing of results or intermediate results for quantitative or qualitative consistency with other experimental or theoretical information (e.g. yields, backscatter factors). This paper discusses the types of the most important collision cross sections for studies on track phenomena by detailed Monte-Carlo calculations, the necessary accuracy of such data and various means of consistency checks of calculated results. This will be done mainly with examples taken from radiation physics as applied to dosimetric and biological problems (i.e. to gaseous and condensed targets). 12 references, 8 figures

  14. Time-Dependent Wave Packet Dynamics Calculations of Cross Sections for Ultracold Scattering of Molecules

    Science.gov (United States)

    Huang, Jiayu; Liu, Shu; Zhang, Dong H.; Krems, Roman V.

    2018-04-01

    Because the de Broglie wavelength of ultracold molecules is very large, the cross sections for collisions of molecules at ultracold temperatures are always computed by the time-independent quantum scattering approach. Here, we report the first accurate time-dependent wave packet dynamics calculation for reactive scattering of ultracold molecules. Wave packet dynamics calculations can be applied to molecular systems with more dimensions and provide real-time information on the process of bond rearrangement and/or energy exchange in molecular collisions. Our work thus makes possible the extension of rigorous quantum calculations of ultracold reaction properties to polyatomic molecules and adds a new powerful tool for the study of ultracold chemistry.

  15. The characters and developments of therapy and research of neutron-induced acute radiation sickness

    International Nuclear Information System (INIS)

    Wang Xinru; Luo Qingliang; Wang Baoqing; Dong Shukui

    2003-01-01

    Neutron radiation will exert seriously lesions on body and lead to acute radiation sickness (ARS). Neutron induced ARS is characterized by complicated symptoms of gastrointestinal, high earlier death rate and lacking of specially therapeutic approaches. The primarily curable method is to treat patients with comprehensive means such as anti-infection, anti-bleeding and blood infusion, et. Cytokines can improve the hematopoietic functions of bone marrow. With the rapid development of the molecular biology and the emergence of cytokines such as recombinant human interleukin-11 that could help the regeneration of the gastrointestinal mucosa, neutron-induced ARS will be treated with new methods

  16. A new calculation formula of the nuclear cross-section of therapeutic protons

    Directory of Open Access Journals (Sweden)

    Waldemar Ulmer

    2014-03-01

    Full Text Available Purpose: We have previously developed for nuclear cross-sections of therapeutic protons a calculation model, which is founded on the collective model as well as a quantum mechanical many particle problem to derive the S matrix and transition probabilities. In this communication, we show that the resonances can be derived by shifted Gaussian functions, whereas the unspecific nuclear interaction compounds can be represented by an error function, which also provides the asymptotic behavior. Method: The energy shifts can be interpreted in terms of necessary domains of energy to excite typical nuclear processes. Thus the necessary formulas referring to previous calculations of nuclear cross-sections will be represented. The mass number AN determines the strong interaction range, i.e. RStrong = 1.2·10-13·AN1/3cm. The threshold energy ETh of the energy barrier is determined by the condition Estrong = ECoulomb. Results and Conclusion: A linear combination of Gaussians, which contain additional energy shifts, and an error function incorporate a possible representation of Fermi-Dirac statistics, which is applied here to nuclear excitations and reaction with release of secondary particles. The new calculation formula provides a better understanding of different types of resonances occurring in nuclear interactions with protons. The present study is mainly a continuation of published papers.1-3--------------------------------Cite this article as: Ulmer W. A new calculation formula of the nuclear cross-section of therapeutic protons. Int J Cancer Ther Oncol 2014; 2(2:020211. DOI: 10.14319/ijcto.0202.11

  17. Application of an enhanced cross-section interpolation model for highly poisoned LWR core calculations

    International Nuclear Information System (INIS)

    Palau, J.M.; Cathalau, S.; Hudelot, J.P.; Barran, F.; Bellanger, V.; Magnaud, C.; Moreau, F.

    2011-01-01

    Burnable poisons are extensively used by Light Water Reactor designers in order to preserve the fuel reactivity potential and increase the cycle length (without increasing the uranium enrichment). In the industrial two-steps (assembly 2D transport-core 3D diffusion) calculation schemes these heterogeneities yield to strong flux and cross-sections perturbations that have to be taken into account in the final 3D burn-up calculations. This paper presents the application of an enhanced cross-section interpolation model (implemented in the French CRONOS2 code) to LWR (highly poisoned) depleted core calculations. The principle is to use the absorbers (or actinide) concentrations as the new interpolation parameters instead of the standard local burnup/fluence parameters. It is shown by comparing the standard (burnup/fluence) and new (concentration) interpolation models and using the lattice transport code APOLLO2 as a numerical reference that reactivity and local reaction rate prediction of a 2x2 LWR assembly configuration (slab geometry) is significantly improved with the concentration interpolation model. Gains on reactivity and local power predictions (resp. more than 1000 pcm and 20 % discrepancy reduction compared to the reference APOLLO2 scheme) are obtained by using this model. In particular, when epithermal absorbers are inserted close to thermal poison the 'shadowing' ('screening') spectral effects occurring during control operations are much more correctly modeled by concentration parameters. Through this outstanding example it is highlighted that attention has to be paid to the choice of cross-section interpolation parameters (burnup 'indicator') in core calculations with few energy groups and variable geometries all along the irradiation cycle. Actually, this new model could be advantageously applied to steady-state and transient LWR heterogeneous core computational analysis dealing with strong spectral-history variations under

  18. Aorta cross-section calculation and 3D visualization from CT or MRT data using VRML

    Science.gov (United States)

    Grabner, Guenther; Modritsch, Robert; Stiegmaier, Wolfgang; Grasser, Simon; Klinger, Thomas

    2005-04-01

    Quantification of vessel diameters of artherosclerotic or congenital stenosis is very important for the diagnosis of vascular diseases. The aorta extraction and cross-section calculation is a software-based application that offers a three-dimensional, platform-independent, colorized visualization of the extracted aorta with augmented reality information of MRT or CT datasets. This project is based on different types of specialized image processing algorithms, dynamical particle filtering and complex mathematical equations. From this three-dimensional model a calculation of minimal cross sections is performed. In user specified distances, the aorta is cut in differently defined directions which are created through vectors with varying length. The extracted aorta and the derived minimal cross-sections are then rendered with the marching cube algorithm and represented together in a three-dimensional virtual reality with a very high degree of immersion. The aim of this study was to develop an imaging software that delivers cardiologists the possibility of (i) furnishing fast vascular diagnosis, (ii) getting precise diameter information, (iii) being able to process exact, local stenosis detection (iv) having permanent data storing and easy access to former datasets, and (v) reliable documentation of results in form of tables and graphical printouts.

  19. Calculation of homogenized Pickering NGS stainless steel adjuster rod neutron cross-sections using conservation of reaction rates

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, R C [Atlantic Nuclear Services Ltd. (Canada); Tran, F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

    1996-12-31

    A homogenization methodology for calculation of reactivity device incremental cross-sections has been developed using reaction rate conservation (RRC). A heterogeneous transport calculation of flux was utilised to produce the homogenized cross-sections for a finite difference two group diffusion code. The RRC cross-sections have been shown to improve significantly the prediction of reactivity worth for stainless steel adjuster rods installed in Pickering NGS reactors. (author). 10 refs., 3 tabs., 6 figs.

  20. Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions

    International Nuclear Information System (INIS)

    Santiago, Daniela Maiolino Norberto

    2011-01-01

    In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)

  1. A Comparison between Effective Cross Section Calculations using the Intermediate Resonance Approximation and More Exact Methods

    Energy Technology Data Exchange (ETDEWEB)

    Haeggblom, H

    1969-02-15

    In order to investigate some aspects of the 'Intermediate Resonance Approximation' developed by Goldstein and Cohen, comparative calculations have been made using this method together with more accurate methods. The latter are as follows: a) For homogeneous materials the slowing down equation is solved in the fundamental mode approximation with the computer programme SPENG. All cross sections are given point by point. Because the spectrum can be calculated for at most 2000 energy points, the energy regions where the resonances are accurately described are limited. Isolated resonances in the region 100 to 240 eV are studied for {sup 238}U/Fe and {sup 238}U/Fe/Na mixtures. In the regions 161 to 251 eV and 701 to 1000 eV, mixtures of {sup 238}U and Na are investigated. {sup 239}Pu/Na and {sup 239}Pu/{sup 238}U/Na mixtures are studied in the region 161 to 251 eV. b) For heterogeneous compositions in slab geometry the integral transport equation is solved using the FLIS programme in 22 energy groups. Thus, only one resonance can be considered in each calculation. Two resonances are considered, namely those belonging to {sup 238}U at 190 and 937 eV. The compositions are lattices of {sup 238}U and Fe plates. The computer programme DORIX is used for the calculations using the Intermediate Resonance Approximation. Calculations of reaction rates and effective cross sections are made at 0, 300 and 1100 deg K for homogeneous media and at 300 deg K for heterogeneous media. The results are compared to those obtained by using the programmes SPENG and FLIS and using the narrow resonance approximation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  3. Filtered thermal neutron captured cross-sections measurements and decay heat calculations

    International Nuclear Information System (INIS)

    Son, Pham Ngoc; Tan, Vuong Huu

    2014-01-01

    Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R cd ) of 420 and neutron flux (Φ th ) of 1.6x10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51 V, 55 Mn, 180 Hf and 186 W by the activation method relative to the standard reaction 197 Au(n,g) 198 Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U, 238 U, 239 Pu and 232 Th are introduced in this report. (author)

  4. Neutron-induced Single Event Upset on the RPC front-end chips for the CMS experiment

    Energy Technology Data Exchange (ETDEWEB)

    Abbrescia, M.; Colaleo, A.; Iaselli, G.; Loddo, F.; Maggi, M.; Marangelli, B.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Romano, F.; Altieri, S.; Belli, G.; Bruno, G.; Guida, R.; Merlo, M.; Ratti, S.P.; Riccardi, C.; Torre, P.; Vitulo, P. E-mail: paolo.vitulo@pv.infn.it; De Bari, A.; Manera, S

    2002-05-21

    Neutrons from a reactor and from a cyclotron have been used to characterise the CMS Resistive Plate Chambers (RPCs) front-end chip to neutron-induced damaging events. Single Event Upset (SEU) cross-sections have been measured up to 60 MeV for different chip thresholds. Tests at a reactor were done with an integrated fast (E{sub n}>3 MeV) neutron fluence of 1.7x10{sup 10} cm{sup -2} and a thermal neutron fluence of 9.5x10{sup 11} cm{sup -2}. High-energy neutrons from a cyclotron were used up to a fluence of 10{sup 12} cm{sup -2}. Data indicate the existence of a chip SEU sensitivity already at thermal energy and a saturated SEU cross-section from 3 to 60 MeV. Values of the SEU cross-sections from the thermal run well agree with those obtained by another CMS group that uses the same technology (0.8 {mu}m BiCMOS) though with different architecture. Cross-sections obtained with fast neutrons (from 3 MeV to about 10 MeV) are consistently higher by one order of magnitude compared to the thermal one. The average time between consecutive SEU events in each chip of the CMS barrel RPCs can be estimated to be 1 h.

  5. The LAW Library -- A multigroup cross-section library for use in radioactive waste analysis calculations

    International Nuclear Information System (INIS)

    Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.

    1994-08-01

    The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ( 14 N 7 , 15 N 7 , 16 O 8 , 154Eu 63 , and 155 Eu 63 ). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections

  6. Calculation of the differential cross sections of excitation and ionization of a helium atom by electrons

    International Nuclear Information System (INIS)

    Demkin, V.P.; Pecheritsyn, A.A.

    1995-01-01

    Equations for the amplitudes and differential cross sections of electronic excitation and ionization of a helium atom are derived in the approximation of a open-quotes frozenclose quotes ion core. The wave functions of the discrete states are chosen in the form of generalized hydrogenlike orbitals. The radial wave functions of the continuous spectrum are determined by solving the equation of motion numerically. The differential excitation cross sections of excitation of the 2p, 3p, and 4p levels and ionization of a helium atom by electrons are calculated in the energy range up to 50 eV. Estimates are obtained for the nonorthogonal wave functions in the amplitudes of the excitation and ionization processes. It is shown that the given method is more compatible with experiment than the Born method

  7. Accurate Calculations of Rotationally Inelastic Scattering Cross Sections Using Mixed Quantum/Classical Theory.

    Science.gov (United States)

    Semenov, Alexander; Babikov, Dmitri

    2014-01-16

    For computational treatment of rotationally inelastic scattering of molecules, we propose to use the mixed quantum/classical theory, MQCT. The old idea of treating translational motion classically, while quantum mechanics is used for rotational degrees of freedom, is developed to the new level and is applied to Na + N2 collisions in a broad range of energies. Comparison with full-quantum calculations shows that MQCT accurately reproduces all, even minor, features of energy dependence of cross sections, except scattering resonances at very low energies. The remarkable success of MQCT opens up wide opportunities for computational predictions of inelastic scattering cross sections at higher temperatures and/or for polyatomic molecules and heavier quenchers, which is computationally close to impossible within the full-quantum framework.

  8. Application of reactors for testing neutron-induced upsets in commercial SRAMs

    International Nuclear Information System (INIS)

    Griffin, P.J.; Luera, T.F.; Sexton, F.W.; Cooper, P.J.; Karr, S.G.; Hash, G.L.; Fuller, E.

    1997-01-01

    Reactor neutron environments can be used to test/screen the sensitivity of unhardened commercial SRAMs to low-LET neutron-induced upset. Tests indicate both thermal/epithermal (< 1 keV) and fast neutrons can cause upsets in unhardened parts. Measured upset rates in reactor environments can be used to model the upset rate for arbitrary neutron spectra

  9. 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)

  10. Study on the identification method of chemical warfare agents with spectroscopy of neutron induced γ rays

    International Nuclear Information System (INIS)

    Liu Boxue; Li Yun; Li Xiangbao

    1996-01-01

    The paper briefly describes some non-destructive verification technologies of chemical warfare agents in-site, and some application of neutron induced gamma ray analysis, such as multi-elements analysis of coal, hidden explosive detection and identification of chemical agents. It also describes some problems in developing the portable isotopic neutron spectroscopy for non-destructive evaluation of chemical warfare agents

  11. Preliminary Experiment on Neutron-Induced Mn Activity in Mn-Cd Solutions

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1942-07-01

    This report was written by E. Broda, J. Gueron and L. Kowarski at the Cavendish Laboratory (Cambridge) in June 1942 and is about a preliminary experiment on neutron-induced Mn activity in Mn-Cd solutions. The description of the experiment and the results can be found also in this report. (nowak)

  12. General calculation of the cross section for dark matter annihilations into two photons

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Cely, Camilo [Service de Physique Théorique, Université Libre de Bruxelles, Boulevard du Triomphe, CP225, 1050 Brussels (Belgium); Rivera, Andres, E-mail: Camilo.Alfredo.Garcia.Cely@ulb.ac.be, E-mail: afelipe.rivera@udea.edu.co [Instituto de Física, Universidad de Antioquia, Calle 70 No. 52-21, Medellín (Colombia)

    2017-03-01

    Assuming that the underlying model satisfies some general requirements such as renormalizability and CP conservation, we calculate the non-relativistic one-loop cross section for any self-conjugate dark matter particle annihilating into two photons. We accomplish this by carefully classifying all possible one-loop diagrams and, from them, reading off the dark matter interactions with the particles running in the loop. Our approach is general and leads to the same results found in the literature for popular dark matter candidates such as the neutralinos of the MSSM, minimal dark matter, inert Higgs and Kaluza-Klein dark matter.

  13. General calculation of the cross section for dark matter annihilations into two photons

    International Nuclear Information System (INIS)

    Garcia-Cely, Camilo; Rivera, Andres

    2017-01-01

    Assuming that the underlying model satisfies some general requirements such as renormalizability and CP conservation, we calculate the non-relativistic one-loop cross section for any self-conjugate dark matter particle annihilating into two photons. We accomplish this by carefully classifying all possible one-loop diagrams and, from them, reading off the dark matter interactions with the particles running in the loop. Our approach is general and leads to the same results found in the literature for popular dark matter candidates such as the neutralinos of the MSSM, minimal dark matter, inert Higgs and Kaluza-Klein dark matter.

  14. Theoretical calculation of fully differential cross sections for electron-impact ionization of hydrogen molecules

    International Nuclear Information System (INIS)

    Gao Junfang; Madison, D H; Peacher, J L

    2006-01-01

    We have recently proposed the orientation averaged molecular orbital (OAMO) approximation for calculating fully differential cross sections (FDCS) for electron-impact ionization of molecules averaged over all molecular orientations. Orientation averaged FDCS were calculated for electron-impact ionization of nitrogen molecules using the distorted wave impulse approximation (DWIA) and the molecular three-body distorted wave (M3DW) approximation. In this paper, we use the same methods to examine the FDCS for ionization of hydrogen molecules. It is found that the DWIA yields reasonable results for high-energy incident electrons. While the DWIA breaks down for low-energy electrons, the M3DW gives reasonable results down to incident-electron energies around 35 eV

  15. SCATLAW: a code of scattering law and cross sections calculation for liquids and solids

    International Nuclear Information System (INIS)

    Padureanu, I.; Rapeanu, S.; Rotarascu, G.; Craciun, C.

    1978-11-01

    A code for calculation of the scattering law S(Q,ω), differential and double differential cross sections and scattering kernels in the energy range E(0 - 683 meV) and wave-vector transfer Q(0 - 40 A -1 ) is presented. The code can be used both for solids and liquids which are coherent or incoherent scatterer. For liquids the calculations are based on the most recent theoretical models involving the correlation functions and generalized field approach. The phonon expansion model and the free gas model are also analysed in term of frequency spectra obtained from inelastic neutron scattering using time-of-flight technique. Several results on liquid sodium at T = 233 deg C and on liquid bismuth at T = 286 deg C and T = 402 deg C are presented. (author)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  17. Calculation of photoionization differential cross sections using complex Gauss-type orbitals.

    Science.gov (United States)

    Matsuzaki, Rei; Yabushita, Satoshi

    2017-09-05

    Accurate theoretical calculation of photoelectron angular distributions for general molecules is becoming an important tool to image various chemical reactions in real time. We show in this article that not only photoionization total cross sections but also photoelectron angular distributions can be accurately calculated using complex Gauss-type orbital (cGTO) basis functions. Our method can be easily combined with existing quantum chemistry techniques including electron correlation effects, and applied to various molecules. The so-called two-potential formula is applied to represent the transition dipole moment from an initial bound state to a final continuum state in the molecular coordinate frame. The two required continuum functions, the zeroth-order final continuum state and the first-order wave function induced by the photon field, have been variationally obtained using the complex basis function method with a mixture of appropriate cGTOs and conventional real Gauss-type orbitals (GTOs) to represent the continuum orbitals as well as the remaining bound orbitals. The complex orbital exponents of the cGTOs are optimized by fitting to the outgoing Coulomb functions. The efficiency of the current method is demonstrated through the calculations of the asymmetry parameters and molecular-frame photoelectron angular distributions of H2+ and H2 . In the calculations of H2 , the static exchange and random phase approximations are employed, and the dependence of the results on the basis functions is discussed. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

  18. Program TOTELA calculating basic cross sections in intermediate energy region by using systematics

    International Nuclear Information System (INIS)

    Fukahori, Tokio; Niita, Koji

    2000-01-01

    Program TOTELA can calculate neutron- and proton-induced total, elastic scattering and reaction cross sections and angular distribution of elastic scattering in the intermediate energy region from 20 MeV to 3 GeV. The TOTELA adopts the systematics modified from that by Pearlstein to reproduce the experimental data and LA150 evaluation better. The calculated results compared with experimental data and LA150 evaluation are shown in figures. The TOTELA results can reproduce those data almost well. The TOTELA was developed to fill the lack of experimental data of above quantities in the intermediate energy region and to use for production of JENDL High Energy File. In the case that there is no experimental data of above quantities, the optical model parameters can be fitted by using TOTELA results. From this point of view, it is also useful to compare the optical model calculation by using RIPL with TOTELA results, in order to verify the parameter quality. Input data of TOTELA is only atomic and mass numbers of incident particle and target nuclide and input/output file names. The output of TOTELA calculation is in ENDF-6 format used in the intermediate energy nuclear data files. It is easy to modify the main routine by users. Details are written in each subroutine and main routine

  19. Fission barrier theory and its application to the calculation of actinide neutron cross-sections

    International Nuclear Information System (INIS)

    Lynn, J.E.

    1980-01-01

    The lectures discuss the possibilities and realisations of applying nuclear fission theory to the calculation of unknown nuclear data required for applications, principally in the nuclear power field. A brief description of the fundamentals of fission theory, the nature of the potential energy surface in the deformation plane, and of the inertial tensor, is given, and the accuracy of the theoretical calculations is discussed. It is concluded that it is impracticable to obtain required quantities such as neutron cross-sections from such fundamental calculations at present. On the other hand the fundamental theory reveals a wealth of phenomenological aspects of the fission process which can be incorporated into nuclear reaction theory. It is then shown how reaction theory thus extended to take correct account of the structured (''double-humped'') fission barrier can be used to parametrise the barrier by analysis of experimental data, and subsequently to calculate new data. Descriptions of computer programmes and illustrations of the application of the methods to actual physical examples are included in this account. (author)

  20. Fully differential cross sections for the single ionization of helium by fast ions: Classical model calculations

    Science.gov (United States)

    Sarkadi, L.

    2018-04-01

    Fully differential cross sections (FDCSs) have been calculated for the single ionization of helium by 1- and 3-MeV proton and 100-MeV/u C6 + ion impact using the classical trajectory Monte Carlo (CTMC) method in the nonrelativistic, three-body approximation. The calculations were made employing a Wigner-type model in which the quantum-mechanical position distribution of the electron is approximated by a weighted integral of the microcanonical distribution over a range of the binding energy of the electron. In the scattering plane, the model satisfactorily reproduces the observed shape of the binary peak. In the region of the peak the calculated FDCSs agree well with the results of continuum-distorted-wave calculations for all the investigated collisions. For 1-MeV proton impact the experimentally observed shift of the binary peak with respect to the first Born approximation is compared with the shifts obtained by different higher-order quantum-mechanical theories and the present CTMC method. The best result was achieved by CTMC, but still a large part of the shift remained unexplained. Furthermore, it was found that the classical theory failed to reproduce the shape of the recoil peak observed in the experiments, it predicts a much narrower peak. This indicates that the formation of the recoil peak is dominated by quantum-mechanical effects. For 100-MeV/u C6 + ion impact the present CTMC calculations confirmed the existence of the "double-peak" structure of the angular distribution of the electron in the plane perpendicular to the momentum transfer, in accordance with the observation, the prediction of an incoherent semiclassical model, and previous CTMC results. This finding together with wave-packet calculations suggests that the "C6 + puzzle" may be solved by considering the loss of the projectile coherence. Experiments to be conducted using ion beams of anisotropic coherence are proposed for a more differential investigation of the ionization dynamics.

  1. A simplified unified Hauser-Feshbach/Pre-Equilibrium model for calculating double differential cross sections

    International Nuclear Information System (INIS)

    Fu, C.Y.

    1988-01-01

    A unified Hauser-Feshbach/Pre-Equilibrium model is extended and simplified. The extension involves the addition of correlations among states of different total quantum numbers (J and J') and the introduction of consistent level density formulas for the H-F and the P-E parts of the calculation. The simplification, aimed at reducing the computational cost, is achieved mainly by keeping only the off-diagonal terms that involve strongly correlated 2p-1h states. A correlation coefficient is introduced to fit the experimental data. The model has been incorporated into the multistep H-F model code TNG. Calculated double differential (n,xn) cross sections at 14 and 25.7 MeV for iron, niobium, and bismuth are in good agreement with experiments. In use at ORNL and JAERI, the TNG code in various stages of development has been applied with success to the evaluation of double differential (n,xn) cross sections from 1 to 20 MeV for the dominant isotopes of chromium, manganese, iron, nickel, copper, and lead. 11 refs., 2 figs

  2. Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

    International Nuclear Information System (INIS)

    Ku, L.P.; Price, W.G. Jr.

    1977-08-01

    The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model

  3. Monte Carlo Calculation of Thermal Neutron Inelastic Scattering Cross Section Uncertainties by Sampling Perturbed Phonon Spectra

    Science.gov (United States)

    Holmes, Jesse Curtis

    Nuclear data libraries provide fundamental reaction information required by nuclear system simulation codes. The inclusion of data covariances in these libraries allows the user to assess uncertainties in system response parameters as a function of uncertainties in the nuclear data. Formats and procedures are currently established for representing covariances for various types of reaction data in ENDF libraries. This covariance data is typically generated utilizing experimental measurements and empirical models, consistent with the method of parent data production. However, ENDF File 7 thermal neutron scattering library data is, by convention, produced theoretically through fundamental scattering physics model calculations. Currently, there is no published covariance data for ENDF File 7 thermal libraries. Furthermore, no accepted methodology exists for quantifying or representing uncertainty information associated with this thermal library data. The quality of thermal neutron inelastic scattering cross section data can be of high importance in reactor analysis and criticality safety applications. These cross sections depend on the material's structure and dynamics. The double-differential scattering law, S(alpha, beta), tabulated in ENDF File 7 libraries contains this information. For crystalline solids, S(alpha, beta) is primarily a function of the material's phonon density of states (DOS). Published ENDF File 7 libraries are commonly produced by calculation and processing codes, such as the LEAPR module of NJOY, which utilize the phonon DOS as the fundamental input for inelastic scattering calculations to directly output an S(alpha, beta) matrix. To determine covariances for the S(alpha, beta) data generated by this process, information about uncertainties in the DOS is required. The phonon DOS may be viewed as a probability density function of atomic vibrational energy states that exist in a material. Probable variation in the shape of this spectrum may be

  4. Collision cross section calculations for polyatomic ions considering rotating diatomic/linear gas molecules

    International Nuclear Information System (INIS)

    Larriba-Andaluz, Carlos; Hogan, Christopher J.

    2014-01-01

    Structural characterization of ions in the gas phase is facilitated by measurement of ion collision cross sections (CCS) using techniques such as ion mobility spectrometry. Further information is gained from CCS measurement when comparison is made between measurements and accurately predicted CCSs for model ion structures and the gas in which measurements are made. While diatomic gases, namely molecular nitrogen and air, are being used in CCS measurement with increasingly prevalency, the majority of studies in which measurements are compared to predictions use models in which gas molecules are spherical or non-rotating, which is not necessarily appropriate for diatomic gases. Here, we adapt a momentum transfer based CCS calculation approach to consider rotating, diatomic gas molecule collisions with polyatomic ions, and compare CCS predictions with a diatomic gas molecule to those made with a spherical gas molecular for model spherical ions, tetra-alkylammonium ions, and multiply charged polyethylene glycol ions. CCS calculations are performed using both specular-elastic and diffuse-inelastic collisions rules, which mimic negligible internal energy exchange and complete thermal accommodation, respectively, between gas molecule and ion. The influence of the long range ion-induced dipole potential on calculations is also examined with both gas molecule models. In large part we find that CCSs calculated with specular-elastic collision rules decrease, while they increase with diffuse-inelastic collision rules when using diatomic gas molecules. Results clearly show the structural model of both the ion and gas molecule, the potential energy field between ion and gas molecule, and finally the modeled degree of kinetic energy exchange between ion and gas molecule internal energy are coupled to one another in CCS calculations, and must be considered carefully to obtain results which agree with measurements

  5. Cross Sections Calculations of ( d, t) Nuclear Reactions up to 50 MeV

    Science.gov (United States)

    Tel, E.; Yiğit, M.; Tanır, G.

    2013-04-01

    In nuclear fusion reactions two light atomic nuclei fuse together to form a heavier nucleus. Fusion power is the power generated by nuclear fusion processes. In contrast with fission power, the fusion reaction processes does not produce radioactive nuclides. The fusion will not produce CO2 or SO2. So the fusion energy will not contribute to environmental problems such as particulate pollution and excessive CO2 in the atmosphere. Fusion powered electricity generation was initially believed to be readily achievable, as fission power had been. However, the extreme requirements for continuous reactions and plasma containment led to projections being extended by several decades. In 2010, more than 60 years after the first attempts, commercial power production is still believed to be unlikely before 2050. Although there have been significant research and development studies on the inertial and magnetic fusion reactor technology, there is still a long way to go to penetrate commercial fusion reactors to the energy market. In the fusion reactor, tritium self-sufficiency must be maintained for a commercial power plant. Therefore, for self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. Working out the systematics of ( d, t) nuclear reaction cross sections is of great importance for the definition of the excitation function character for the given reaction taking place on various nuclei at different energies. Since the experimental data of charged particle induced reactions are scarce, self-consistent calculation and analyses using nuclear theoretical models are very important. In this study, ( d, t) cross sections for target nuclei 19F, 50Cr, 54Fe, 58Ni, 75As, 89Y, 90Zr, 107Ag, 127I, 197Au and 238U have been investigated up to 50 MeV deuteron energy. The excitation functions for ( d, t) reactions have been calculated by pre-equilibrium reaction mechanism. Calculation results have been also compared with the available measurements in

  6. FPSPH DFPSPF, Line Shape Function for Doppler Broadened Resonance Cross-Sections Calculation

    International Nuclear Information System (INIS)

    Ribon, P.

    1982-01-01

    1 - Description of problem or function: In the computation of Doppler- broadened resonance cross sections, use is made of the symmetric and anti-symmetric line shape functions. These functions usually denoted as Psi and Phi (Psi and Chi in Anglo-Saxon formalism) are defined in terms of the real and imaginary parts of the error function for complex arguments. They are the product of the convolution of a Gaussian function with the symmetric and anti-symmetric Breit-Wigner functions, respectively. FPSPH and DFPSPH compute these functions. 2 - Method of solution: For (1+x 2 ) > 20 Beta 2 , the calculation is based upon the asymptotic expansion: Psi+(i*Phi) = 1/(1-ix)*(1-t+3t 2 -3.5t 3 +3.5+7t 4 ---), with: t = 1/(2z 2 ); z = (1-ix)/Beta. The half-plane (Beta,x) is split in several parts, and use is made of PADE approximants. For 1 + x 2 2 , the calculation is based upon the relation with the erf function: Psi + i*Phi = SQRT(Pi)/Beta*(e (z 2 ) )*(1-erf(z)) (z = (1-ix)/Beta, and erf(z) being calculated from its analytic expansion: erf(z) = 2/SQRT(Pi)*z*e (-z 2 ) *(1+z 2 /3+z 4 /(3*5) + z 6 /(3*5*7)+---). PADE approximants are used to compute the expansion and e z 2

  7. Angular distributions in the neutron-induced fission of actinides

    CERN Multimedia

    In 2003 the n_TOF Collaboration performed the fission cross section measurement of several actinides ($^{232}$Th, $^{233}$U, $^{234}$U, $^{237}$Np) at the n_TOF facility using an experImental setup made of Parallel Plate Avalanche Counters (PPAC). The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. We have been therefore able to cover the very broad neutron energy range 1eV-1GeV, taking full benefit of the unique characteristics of the n_TOF facility. Figure 1 shows an example obtained in the case of $^{237}$Np where the n_ TOF measurement showed that the cross section was underestimated by a large factor in the resonance region.

  8. Adjusts of control rod cross sections and its utilization in power distribution calculations for Angra-1 reactor

    International Nuclear Information System (INIS)

    Pina, C.M. de

    1981-01-01

    One of the most important part in neutronics calculations is the study of core behavior with inserted control rods. The first stage of this calculations consists in generating equivalent microscopic cross sections for the basic cells containing fuel or absorbed material. The cross sections will be then adjusted. The choice of parameters that help in those adjustments, were obtained by the comparisons of data coming from the control rod supercell calculations with the Hammer and Citation computer codes. The effect of those adjustments in core integral parameters was evaluated; in this work only the core power two-dimensional distribution calculations with the D bank completely inserted, is studied. (E.G.) [pt

  9. All-loop calculations of total, elastic and single diffractive cross sections in RFT via the stochastic approach

    International Nuclear Information System (INIS)

    Kolevatov, R. S.; Boreskov, K. G.

    2013-01-01

    We apply the stochastic approach to the calculation of the Reggeon Field Theory (RFT) elastic amplitude and its single diffractive cut. The results for the total, elastic and single difractive cross sections with account of all Pomeron loops are obtained.

  10. All-loop calculations of total, elastic and single diffractive cross sections in RFT via the stochastic approach

    Energy Technology Data Exchange (ETDEWEB)

    Kolevatov, R. S. [SUBATECH, Ecole des Mines de Nantes, 4 rue Alfred Kastler, 44307 Nantes Cedex 3 (France); Boreskov, K. G. [Institute of Theoretical and Experimental Physics, 117259, Moscow (Russian Federation)

    2013-04-15

    We apply the stochastic approach to the calculation of the Reggeon Field Theory (RFT) elastic amplitude and its single diffractive cut. The results for the total, elastic and single difractive cross sections with account of all Pomeron loops are obtained.

  11. Eikonal calculation of electron-capture cross sections in collisions of H atoms with fast projectiles

    International Nuclear Information System (INIS)

    Ho, T.S.; Lieber, M.; Chan, F.T.

    1981-01-01

    We have employed the eikonal method to calculate the cross section for the capture of an electron into an arbitrary nl subshell in collisions between hydrogen atoms and fast projectiles. the projectiles were protons, C 6+ , O 8+ , and Fe 24+ . The energy ranges considered were 20--100 keV in the proton case, and 40--200 keV per nucleon in the other cases. These projectiles were selected because of their importance in fusion plasmas. For the highly charged case of Fe 24+ we found that our formulas, while exact, involved a high degree of cancellation and produced unreliable numerical results, so that a numerical integration of the penultimate formula was substituted. In the proton case agreement with recent experimental data is excellent

  12. MORSEC-SP, Step Function Angular Distribution for Cross-Sections Calculation by Program MORSE

    International Nuclear Information System (INIS)

    1980-01-01

    1 - Description of problem or function: MORSEC-SP allows one to utilize a step distribution to describe the angular dependence of the multi- group function in the MORSEC cross section module of the MORSE Monte Carlo code. The step distribution is always non-negative and may be used in the random walk and for making point detector estimators. 2 - Method of solution: MORSEC-SP utilizes a table look up procedure to provide the probability of scattering when making point detector estimates for a given incident energy group and scattering angle. In the random walk, the step distributions are converted to cumulative distributions and an angle of scatter is selected from the cumulative distributions. Step distributions are obtained from calculation using the converted moments from the given Legendre coefficients of the scattering distributions. 3 - Restrictions on the complexity of the problem: Additional coding to the MORSEC module is variable dimensional and fully incorporated into blank common

  13. Coupled-channel calculations of partial capture cross sections in multiply charged ion collisions with hydrogen

    International Nuclear Information System (INIS)

    Hansen, J.P.; Taulbjerg, K.; University of Tennessee, Knoxville, Tennessee 37996)

    1989-01-01

    Partial cross sections for electron capture in 1--50-keV collisions of Ar 6+ and Ar 8+ with atomic hydrogen have been calculated using an atomic expansion including two complete principal shells of final states (n=4,5 for Ar 6+ and n=5,6 for Ar 8+ ). The qualitative structure of the results is in good accord with a reaction window picture. The results for Ar 6+ ions are in agreement with published experimental data when precaution is taken with respect to uncertainties in absolute normalization of the data and with respect to a proper analysis of translation energy spectra at lower impact energies. The limited experimental data for Ar 8+ do not agree with the present results

  14. Effect of different level density prescriptions on the calculated neutron nuclear reaction cross sections

    International Nuclear Information System (INIS)

    Garg, S.B.

    1991-01-01

    A detailed investigation is carried out to determine the effect of different level density prescriptions on the computed neutron nuclear data of Ni-58 in the energy range 5-25 MeV. Calculations are performed in the framework of the multistep Hauser-Feshbach statistical theory including the Kalbach exciton model and Brink-Axel giant dipole resonance model for radiative capture. Level density prescriptions considered in this investigation are based on the original Gilbert-Cameron, improved Gilbert-Cameron, backshifted Fermi-gas and the Ignatyuk, et al. approaches. The effect of these prescriptions is discussed, with special reference to (n,p), (n,2n), (n,alpha) and total particle-production cross sections. (author). 17 refs, 8 figs

  15. Measurement and calculation of cross section for (p,x) reactions on natural Fe for 650 MeV protons

    International Nuclear Information System (INIS)

    Janczyszyn, J.; Pohorecki, W.; Domanska, G.; Loska, L.; Taczanowski, S.; Shvetsov, V.

    2006-01-01

    Cross sections for production of radionuclides in (p,x) reactions on natural iron were measured for protons of 650 ± 4 MeV with the use of HPGe gamma spectrometry and calculated with the MCNPX code. The determined cross section values were compared with the computed and other experimental ones

  16. Microscopic calculation of absolute values of two-nucleon transfer cross sections

    International Nuclear Information System (INIS)

    Potel, G.; Bayman, B. F.; Barranco, F.

    2009-01-01

    Arguably, the greatest achievement of many-body physics in the fifties was that of providing the basis for a complete description and a thorough understanding of superconductivity in metals. At the basis of it one finds BCS theory and Josephson effect. The first recognized the central role played by the appearance of a macroscopic coherent field -usually viewed as a condensate of strongly overlapping Cooper pairs-, the quasiparticle vacuum. The second realized that a true gap is not essential for such a state of matter to exist, but rather a finite expectation value of the pair field. Consequently, the specific probe to study the superconducting state is Cooper pair tunneling. Important progress in the understanding of pairing in atomic nuclei may arise from the systematic study of two-particle transfer reactions. Although this subject of research started about the time of the BCS papers, the quantitative calculation of absolute cross sections taking properly into account the full non-locality of the Cooper pairs (correlation length much larger than nuclear dimensions) is still an open problem. We present in this talk the results obtained within a second order DWBA framework for two- nucleon transfer reactions around the Coulomb barrier induced both by heavy and light ions. The calculations were done using a computer code developed for this purpose including the sequential and simultaneous contributions to the process, with microscopic form factors which take into account the relevant structure aspects of the process, such as the nature of the single-particle wavefunctions, the spectroscopic factors, and the interaction potential responsible for the transfer. Reasonable agreement with the experimental absolute values of the differential cross section is obtained without any parameter adjustment (see Figure 1).(author)

  17. Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

    International Nuclear Information System (INIS)

    Cacuci, D.G.

    1978-01-01

    Some aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. In deriving these formulas, use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were greatly aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation, via FORMAC, of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, is one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table

  18. Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

    International Nuclear Information System (INIS)

    Cacuci, D.G.

    1978-04-01

    Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table

  19. Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

    Energy Technology Data Exchange (ETDEWEB)

    Cacuci, D.G.

    1978-04-01

    Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u/sup -5/. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M/sub 2/(u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table.

  20. Approach synthesis of superheavy nuclei from some aspects of cross section calculations

    International Nuclear Information System (INIS)

    Liu Zuhua

    2003-01-01

    Several important aspects in the cross section calculations for the synthesis of superheavy nuclei have been inquired. They are the effects of the coupled-channels, the damping of shell correction energy, the collective enhancements in the level density and the spin distributions of evaporation residues. The channel coupling of relative motion with internal degrees of freedom will enhance significantly the capture cross section at sub-barrier energies. However, recent measurements of spin distributions for the survived compound nucleus show that only low partial waves contribute to the evaporation residues, which should at least partially cancel out the enhancement due to the effects of the channel coupling. The fission barriers are determined mainly by the shell correction energy in the case of superheavy nuclei. Therefore, it is especially important to determine as accurate as possible the damping parameter which describes the decrease of the shell effects influence. In addition, the collective enhancement factor in the level density also plays a very important role in the synthesis of heavy spherical nuclei

  1. Cross sections for electron and photon processes required by electron-transport calculations

    International Nuclear Information System (INIS)

    Peek, J.M.

    1979-11-01

    Electron-transport calculations rely on a large collection of electron-atom and photon-atom cross-section data to represent the response characteristics of the target medium. These basic atomic-physics quantities, and certain qualities derived from them that are now commonly in use, are critically reviewed. Publications appearing after 1978 are not given consideration. Processes involving electron or photon energies less than 1 keV are ignored, while an attempt is made to exhaustively cover the remaining independent parameters and target possibilities. Cases for which data improvements can be made from existing information are identified. Ranges of parameters for which state-of-the-art data are not available are sought out, and recommendations for explicit measurements and/or calculations with presently available tools are presented. An attempt is made to identify the maturity of the atomic-physics data and to predict the possibilities for rapid changes in the quality of the data. Finally, weaknesses in the state-of-the-art atomic-physics data and in the conceptual usage of these data in the context of electron-transport theory are discussed. Brief attempts are made to weight the various aspects of these questions and to suggest possible remedies

  2. Technique of neutron-induced (fission-track) autoradiography with histological detail

    International Nuclear Information System (INIS)

    Smith, J.M.; Taylor, G.N.; Jee, W.S.S.

    1980-01-01

    The primary advantage of neutron-induced or fission-track autoradiography compared with conventional autoradiography is that for low concentrations of fissile nuclides prohibitively long exposure times may be avoided. However, it is difficult to produce imaging of biological structures on the neutron-induced autoradiograph which would allow localization of the nuclide histologically. The technique presented circumvents this difficulty using a thin polycarbonate film applied to the histologically stained tissue section mounted on a quartz substrate. After irradiation of the tissue section with an appropriate thermal neutron flux, the fission fragment tracks are revealed by etching the film with KOH. The tracks, superimposed on the stained tissue, may be observed under the light microscope in the same manner as for conventional nuclear emulsion autoradiography

  3. Neutron-induced single event upsets in static RAMS observed at 10 km flight altitude

    International Nuclear Information System (INIS)

    Olsen, J.; Becher, P.E.; Fynbo, P.B.; Raaby, P. Schultz, J.

    1993-01-01

    Neutron induced single event upsets (SEUs) in static memory devices (SRAMs) have so far been seen only in laboratory environments. The authors report observations of 14 neutron induced SEUs at commercial aircraft flight altitudes as well. The observed SEU rate at 10 km flight altitude based on exposure of 160 standard 256 Kbit CMOS SRAMs is 4.8 · 10 -8 upsets/bit/day. In the laboratory 117 SRAMs of two different brands were irradiated with fast neutrons from a Pu-Be source. A total of 176 SEUs have been observed, among these are two SEU pairs. The upset rates from the laboratory tests are compared to those found in the airborne SRAMS

  4. Hauser-Feshbach cross-section calculations for elastic and inelastic scattering of alpha particles-program CORA

    International Nuclear Information System (INIS)

    Hartman, A.; Siemaszko, M.; Zipper, W.

    1975-01-01

    The program CORA was prepared on the basis of Hauser and Feshbach compound reaction formalism. It allows the differential cross-section distributions for the elastic and inelastic scattering of alpha particles (via compound nucleus state) to be calculated. The transmission coefficients are calculated on the basis of a four parameter optical model. The search procedure is also included. (author)

  5. Comparison of fast neutron-induced tracks in plastics using the electrochemical etching method

    International Nuclear Information System (INIS)

    Cotter, S.J.; Gammage, R.B.; Thorngate, J.H.; Ziemer, P.L.

    1979-01-01

    Four plastics were examined by the electrochemical etching method for their suitability in registering fast neutron-induced recoil particle tracks. The plastics were cellulose acetate, cellulose triacetate, cellulose acetobutyrate and polycarbonate. Cellulose acetate and triacetate displayed high levels of water absorptivity during etching while the acetobutyrate foils cracked due to electromechanical stresses at high frequencies (>500 Hz). The clarity of the etched track was superior in the polycarbonate foils, suggesting the latter as the generally preferred dosimeter for fast neutrons. (author)

  6. Uranium analysis by neutron induced fissionography method using solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Akyuez, T.; Tretyakova, S. P.; Guezel, T.; Akyuz, S.

    1999-01-01

    In this study total twenty samples (eight reference materials and twelve sediment samples) were analysed for their uranium content which is in the range of 1-17 μg/g, by neutron induced fissionography (NIF) method using solid state nuclear track detectors (SSNTDs) in comparison with the results of neutron activation analysis (NAA), delayed neutron counting (DNC) technique or fluorometric method. It is found that NIF method using SSNTDs is very sensitive for analysis of uranium

  7. Uranium analysis by neutron induced fissionography method using solid state nuclear track detectors

    CERN Document Server

    Akyuez, T; Guezel, T; Akyuz, S

    1999-01-01

    In this study total twenty samples (eight reference materials and twelve sediment samples) were analysed for their uranium content which is in the range of 1-17 mu g/g, by neutron induced fissionography (NIF) method using solid state nuclear track detectors (SSNTDs) in comparison with the results of neutron activation analysis (NAA), delayed neutron counting (DNC) technique or fluorometric method. It is found that NIF method using SSNTDs is very sensitive for analysis of uranium.

  8. A semi-empirical concept for the calculation of electron-impact ionization cross sections of neutral and ionized fullerenes

    International Nuclear Information System (INIS)

    Deutsch, H.; Scheier, P.; Maerk, T.D.; Becker, K.

    2002-01-01

    A semi-empirical approach to the calculation of cross section functions (absolute value and energy dependence) for the electron-impact ionization of several neutral and ionized fullerenes C 60 n+ (n =0-3) was developed, for which reliable experimental data have been reported. In particular, it is proposed a modification of the simplistic assumption that the ionization cross section of a cluster/fullerene is given as the product of the monomer ionization cross section and a factor m a , where 'm' is the number of monomers in the ensemble and 'a' is a constant. A comparison between these calculations and the available experimental data reveals good agreement for n = 0,103. In the case of ionization of C 60 2+ (n = 2) the calculation lies significantly below the measured cross section which it was interpret as an indication that additional indirect ionization processes are present for this charge state. (nevyjel)

  9. Analysis of cosmic ray neutron-induced single-event phenomena

    International Nuclear Information System (INIS)

    Tukamoto, Yasuyuki; Watanabe, Yukinobu; Nakashima, Hideki

    2003-01-01

    We have developed a database of cross sections for the n+ 28 Si reaction in the energy range between 2 MeV and 3 GeV in order to analyze single-event upset (SEU) phenomena induced by cosmic-ray neutrons in semiconductor memory devices. The data are applied to calculations of SEU cross sections using the Burst Generation Rate (BGR) model including two parameters, critical charge and effective depth. The calculated results are compared with measured SEU cross-sections for energies up to 160 MeV, and the reaction products that provide important effects on SEU are mainly investigated. (author)

  10. Calculation of three-dimensional mass flow and temperature distributions of nuclear reactors using the hardy cross iterative global solution

    International Nuclear Information System (INIS)

    Silva Neto, A.J. da; Alvim, A.C.M.

    1989-01-01

    This work describes the thermalhydraulics code CROSS, designed for micro-computer calculation of heat and mass flow distributions in LWR nuclear reactor cores using the Hardy Cross method. Equations to calculate the pressure variations in the coolant channels are presented, along with derivation of a linear system of equations to calculate the energy balance. This system is solved through the Benachievicz method. A case study is presented, showing that the methodology developed in this work can be used in place of the forward marching multi-channel codes. (author) [pt

  11. Some time dependent aspects of fast neutron induced atomic cascades

    International Nuclear Information System (INIS)

    Williams, M.M.R.

    1976-01-01

    Analytical results are obtained for the time-energy distribution of neutrons and the associated displaced atoms slowing down in an amorphous medium according to a general force law. Explicit results are given for the inverse power law, and applications to hard-sphere and Coulomb scattering are discussed. Complete results are obtained for the steady state energy distribution of particles arising from a primary knock-on, and from a neutron initiated cascade. The speed of the slowing down process is assessed by calculating the slowing down time of particles. Two different concepts of slowing down time are discussed, one based upon a density average and the other on a slowing down density average. It is shown that the latter definition is physically more realistic and mathematically simpler. (author)

  12. Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

    International Nuclear Information System (INIS)

    Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

    1995-05-01

    The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%

  13. Neutron-induced hydrogen and helium production in iron

    Energy Technology Data Exchange (ETDEWEB)

    Haight, Robert C.

    2004-01-01

    In support of the Advanced Fuel Cycle Initiative, cross sections for hydrogen and helium production by neutrons are being investigated on structural materials from threshold to 100 MeV with the continuous-in-energy spallation neutron source at the Los Alamos Neutron Science Center (LANSCE). The present measurements are for elemental iron. The results are compared with values from the ENDF/B-VI library and its extension with LA150 evaluations. For designs in the Advanced Fuel Cycle Initiative, structural materials will be subjected to very large fluences of neutrons, and the selection of these materials will be guided by their resistance to radiation damage. The macroscopic effects of radiation damage result both from displacement of atoms in the materials as well as nuclear transmutation. We are studying the production of hydrogen and helium by neutrons, because these gases can lead to significant changes in materials properties such as embrittlement and swelling. Our experiments span the full range from threshold to 100 MeV. The lower neutron energies are those characteristic of fission neutrons, whereas the higher energies are relevant for accelerator-based irradiation test facilities. Results for the nickel isotopes, {sup 58,60}Ni, have been reported previously. The present studies are on natural iron.

  14. Angular distribution in the neutron-induced fission of actinides

    Directory of Open Access Journals (Sweden)

    Leong L.S.

    2013-12-01

    Full Text Available Above 1 MeV of incident neutron energy the fission fragment angular distribution (FFAD has generally a strong anisotropic behavior due to the combination of the incident orbital momentum and the intrinsic spin of the fissioning nucleus. This effect has to be taken into account for the efficiency estimation of devices used for fission cross section measurements. In addition it bears information on the spin deposition mechanism and on the structure of transitional states. We designed and constructed a detection device, based on Parallel Plate Avalanche Counters (PPAC, for measuring the fission fragment angular distributions of several isotopes, in particular 232Th. The measurement has been performed at n_TOF at CERN taking advantage of the very broad energy spectrum of the neutron beam. Fission events were recognized by back to back detection in coincidence in two position-sensitive detectors surrounding the targets. The detection efficiency, depending mostly on the stopping of fission fragments in backings and electrodes, has been computed with a Geant4 simulation and validated by the comparison to the measured case of 235U below 3 keV where the emission is isotropic. In the case of 232Th, the result is in good agreement with previous data below 10 MeV, with a good reproduction of the structures associated to vibrational states and the opening of second chance fission. In the 14 MeV region our data are much more accurate than previous ones which are broadly scattered.

  15. Relativistic quantum mechanic calculation of photoionization cross-section of hydrogenic and non-hydrogenic states using analytical potentials

    International Nuclear Information System (INIS)

    Rodriguez, R.; Gil, J.M.; Rubiano, J.G.; Florido, R.; Martel, P.; Minguez, E.

    2005-01-01

    Photoionization process is a subject of special importance in many areas of physics. Numerical methods must be used in order to obtain photoionization cross-sections for non-hydrogenic levels. The atomic data required to calculate them is huge so self-consistent calculations increase computing time considerably. Analytical potentials are a useful alternative because they avoid the iterative procedures typical in self-consistent models. In this work, we present a relativistic quantum calculation of photoionization cross-sections for isolated ions based on an analytical potential to obtain the required atomic data, which is valid both for hydrogenic and non-hydrogenic ions. Comparisons between our results and others obtained using either widely used analytical expressions for the cross-sections or more sophisticated calculations are done

  16. Methods of calculation of cross section of reaction 115In(gamma, n)114mIn

    International Nuclear Information System (INIS)

    Zhaba, V.I.; Parlag, A.M.

    2015-01-01

    The cross section of reaction 115 In(gamma, n) 114m In is expected by different methods. Results of the got cross section it is well comported inter se the Penfold-Leiss and Tikhonov's methods. The calculation of cross section is conducted the Penfold-Leiss method with smoothing out by the method of iterations. Number of iterations n = 1; 3; 5. In the programmatic package of TALYS-1.4 got cross section for five models of closeness of levels. Theoretical and experimental results well coincide in a maximum.

  17. Monte Carlo simulation of the neutron-induced prompt γ ray spectroscopy of the CW abandoned by Japan

    International Nuclear Information System (INIS)

    Wang Bairong; Yang Zhongping; Zhang Wenzhong

    2005-01-01

    This paper introduced the principle of identifying the chemical weapon by neutron-induced prompt γ ray, simulated and analyzed the neutron-induced prompt γ ray spectroscopy of chemical weapon abandoned by Japan in the different condition, using the MCNP-4C Monte Carlo program, whereby supply important datum and reference for the aftertime deeper research and disposal of Japan-abandoned chemical weapon. (authors)

  18. The Monte Carlo simulation of the neutron-induced prompt gamma ray spectroscopy of the CW abandoned by Japan

    International Nuclear Information System (INIS)

    Wang Bairong; Yang Zhongping; Zhan Wenzhong

    2003-01-01

    This paper introduced the principle of identifying the chemical weapon abandoned by Japan by neutron-induced prompt gamma ray. Using the MCNP-4C Monte Carlo program, this paper simulated and analyzed the neutron-induced prompt gamma ray spectroscopy of chemical weapon abandoned by Japan, whereby supply important datum and reference for the aftertime deeper research and disposal of Japan-abandoned chemical weapon. (authors)

  19. Two-level MOC calculation scheme in APOLLO2 for cross-section library generation for LWR hexagonal assemblies

    International Nuclear Information System (INIS)

    Petrov, Nikolay; Todorova, Galina; Kolev, Nikola; Damian, Frederic

    2011-01-01

    The accurate and efficient MOC calculation scheme in APOLLO2, developed by CEA for generating multi-parameterized cross-section libraries for PWR assemblies, has been adapted to hexagonal assemblies. The neutronic part of this scheme is based on a two-level calculation methodology. At the first level, a multi-cell method is used in 281 energy groups for cross-section definition and self-shielding. At the second level, precise MOC calculations are performed in a collapsed energy mesh (30-40 groups). In this paper, the application and validation of the two-level scheme for hexagonal assemblies is described. Solutions for a VVER assembly are compared with TRIPOLI4® calculations and direct 281g MOC solutions. The results show that the accuracy is close to that of the 281g MOC calculation while the CPU time is substantially reduced. Compared to the multi-cell method, the accuracy is markedly improved. (author)

  20. Neutron-induced mutation experiments and total radiation-induced genetic damage in entire genomes of Drosophila melanogaster. Final report, November 1, 1967-August 31, 1980

    International Nuclear Information System (INIS)

    Abrahamson, S.

    1981-02-01

    Neutron-induced mutation experiments with Drosophila oogonia were conducted at the University of Wisconsin, with irradiations being carried out at the RARAF facility at Brookhaven National Laboratory. X-linked recessive lethals and specific locus mutations were studied. Using the α value of the weighted linear regression equation for lethal data, RBE's relative to X-rays were calculated for the energies of neutrons studied. They are: 15 MeV to 2.0; 6 MeV to 2.9; 2 MeV to 3.2; .66 MeV to 4.0; .43 MeV to 4.8. The dose/frequency response curves for lethal data of all neutron energies studied was suggestive of a quadratic component. All data best fit a linear hypothesis, however. Control data for specific locus mutations was used to estimate the number of loci on the X-chromosome which are capable of mutating to lethals. Neutron-induced data for specific locus mutation was inconclusive due to the high error inherent in the frequencies obtained

  1. Calculation of actual cross sections and thermalization of neutrons; Calcul des sections efficaces effectives et thermalisation des neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R.

    1963-05-15

    This report gathers and presents in a simple way results of studies performed at the CEA on issues of spectra in thermal reactors. It is in fact a synthesis of results eventually published in different documents. It first presents the notion of actual cross section as it was introduced by Westcott to characterize the dependence of neutron behaviour on speed distribution. It addresses the case of a homogeneous medium with a conventional model, with the heavy gas model, and with the hydrogen gas model. It generalizes the approach by the differential model. The next part addresses the case of a heterogeneous medium, and the case of presence of moderator nuclei within the fuel [French] Le present rapport a pour objet de rassembler et de presenter de maniere simple les resultats des etudes effectuees au CE.A. sur les problemes de spectres dans les reacteurs thermiques. Ces resultats se trouvaient disperses dans plusieurs documents, ou n'etaient pas encore rediges, et bien que les etudes se poursuivent, il a paru utile d'en faire une synthese provisoire. On a cherche d'autre part a en donner une presentation elementaire, accessible aux lecteurs peu familiarises avec les problemes de thermalisation; dans cet esprit l'expose a une forme didactique, et comporte des rappels de notions bien connues comme par exemple le formalisme de Westcott. (auteur)

  2. Application of the Trojan Horse Method to study neutron induced reactions: the 17O(n, α14C reaction

    Directory of Open Access Journals (Sweden)

    Gulino M.

    2014-03-01

    Full Text Available The reaction 17O(n, α14C was studied using virtual neutrons coming from the quasi-free deuteron break-up in the three body reaction 17O+d → α+14C+p. This technique, called virtual neutron method, extends the Trojan Horse method to neutron-induced reactions allowing to study the reaction cross section avoiding the suppression effects coming from the penetrability of the centrifugal barrier. For incident neutron energies from thermal up to a few hundred keV, direct experiments have shown the population of two out of three expected excited states at energies 8213 keV and 8282 keV and the influence of the sub-threshold level at 8038 keV. In the present experiment the 18O excited state at E* = 8.125 MeV, missing in the direct measurement, is observed. The angular distributions of the populated resonances have been measured for the first time. The results unambiguously indicate the ability of the method to overcome the centrifugal barrier suppression effect and to pick out the contribution of the bare nuclear interaction.

  3. Comparison of Hansen--Roach and ENDF/B-IV cross sections for 233U criticality calculations

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1976-01-01

    A comparison is made between criticality calculations performed using ENDF/B-IV cross sections and the 16-group Hansen-- Roach library at ORNL. The area investigated is homogeneous systems of highly enriched 233 U in simple geometries. Calculations are compared with experimental data for a wide range of H/ 233 U ratios. Results show that calculations of k/sub eff/ made with the Hansen--Roach cross sections agree within 1.5 percent for the experiments considered. Results using ENDF/B-IV cross sections were in good agreement for well-thermalized systems, but discrepancies up to 7 percent in k/sub eff/ were observed in fast and epithermal systems

  4. Computer calculation of neutron cross sections with Hauser-Feshbach code STAPRE incorporating the hybrid pre-compound emission model

    International Nuclear Information System (INIS)

    Ivascu, M.

    1983-10-01

    Computer codes incorporating advanced nuclear models (optical, statistical and pre-equilibrium decay nuclear reaction models) were used to calculate neutron cross sections needed for fusion reactor technology. The elastic and inelastic scattering (n,2n), (n,p), (n,n'p), (n,d) and (n,γ) cross sections for stable molybdenum isotopes Mosup(92,94,95,96,97,98,100) and incident neutron energy from about 100 keV or a threshold to 20 MeV were calculated using the consistent set of input parameters. The hydrogen production cross section which determined the radiation damage in structural materials of fusion reactors can be simply deduced from the presented results. The more elaborated microscopic models of nuclear level density are required for high accuracy calculations

  5. Monte Carlo calculation of the cross-section of single event upset induced by 14MeV neutrons

    International Nuclear Information System (INIS)

    Li, H.; Deng, J.Y.; Chang, D.M.

    2005-01-01

    High-density static random access memory may experience single event upsets (SEU) in neutron environments. We present a new method to calculate the SEU cross-section. Our method is based on explicit generation and transport of the secondary reaction products and detailed accounting for energy loss by ionization. Instead of simulating the behavior of the circuit, we use the Monte Carlo method to simulate the process of energy deposition in sensitive volumes. Thus, we do not need to know details about the circuit. We only need a reasonable guess for the size of the sensitive volumes. In the Monte Carlo simulation, the cross-section of SEU induced by 14MeV neutrons is calculated. We can see that the Monte Carlo simulation not only can provide a new method to calculate SEU cross-section, but also can give a detailed description about random process of the SEU

  6. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Mendenhall, Marcus H., E-mail: marcus.h.mendenhall@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States); Weller, Robert A., E-mail: robert.a.weller@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States)

    2012-03-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10{sup 4} (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).

  7. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    International Nuclear Information System (INIS)

    Mendenhall, Marcus H.; Weller, Robert A.

    2012-01-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10 4 (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).

  8. Evaluation of neutron cross sections to 40 MeV for 5456Fe

    International Nuclear Information System (INIS)

    Arthur, E.D.; Young, P.G.

    1980-01-01

    Cross sections for neutron-induced reactions on 54 56 Fe were calculated by employing several nuclear models: optical, Hauser-Feshbach, preequilibrium and DWBA - in the energy range between 3 and 40 MeV. As a prelude to the calculations, the necessary input parameters were determined or verified through analysis of a large body of experimental data for both neutron- and proton-induced reactions in this mass and energy region. This technique also led to cross sections in which the simultaneous influence of available data types added to their consistency and reliability. Calculated cross sections as well as neutron and gamma-ray emission spectra were incorporated into an ENDF evaluation suitable for use to 40 MeV. 12 figures, 1 table

  9. Calculation of the 1s-2s two-photon excitation cross-section in atomic hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Celik, G.; Celik, E.; Kilic, H.S. [Selcuk Univ., Dept. of Physics, Faculty of Arts and Science (Turkey)

    2008-12-15

    The two-photon excitation cross-section of atomic hydrogen is calculated using explicit summation over intermediate states within the framework of dipole approximation. The matrix element for two-photon excitation is transformed into finite sums, consisting of the product of a radial and angular part. Nine intermediate states are employed in the calculation of the transition matrix element. The two-photon excitation cross-section obtained for the transition 1s{sup 2}S{sub 1/2}-2s{sup 2}S{sub 1/2} in atomic hydrogen is in good agreement with the literature. (authors)

  10. Calculation of the 1s-2s two-photon excitation cross-section in atomic hydrogen

    International Nuclear Information System (INIS)

    Celik, G.; Celik, E.; Kilic, H.S.

    2008-01-01

    The two-photon excitation cross-section of atomic hydrogen is calculated using explicit summation over intermediate states within the framework of dipole approximation. The matrix element for two-photon excitation is transformed into finite sums, consisting of the product of a radial and angular part. Nine intermediate states are employed in the calculation of the transition matrix element. The two-photon excitation cross-section obtained for the transition 1s 2 S 1/2 -2s 2 S 1/2 in atomic hydrogen is in good agreement with the literature. (authors)

  11. Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

    International Nuclear Information System (INIS)

    Bernnat, W.; Keinert, J.; Mattes, M.

    2004-01-01

    For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H 2 O, liquid He, liquid D 2 O, liquid and solid H 2 and D 2 , solid CH 4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S N -transport codes and the Monte Carlo Code MCNP. (orig.)

  12. Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)

    2004-03-01

    For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)

  13. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    OpenAIRE

    Mendenhall, Marcus H.; Weller, Robert A.

    2011-01-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analogous Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross section change. This makes it possible t...

  14. Determination of neutron-induced fission cross-sections of unstable ...

    Indian Academy of Sciences (India)

    2014-11-04

    Nov 4, 2014 ... powerful tool for a wide range of applications that involve compound nuclear reactions. In Bohr's hypothesis, formation and decay of a compound nucleus are considered to be independent of each other. This independence is exploited in the surrogate reaction. Pramana – J. Phys., Vol. 83, No. 5, November ...

  15. KOP program for calculating cross sections of neutron and charged particle interactions with atomic nuclei using the optical model

    International Nuclear Information System (INIS)

    Grudzevich, O.D.; Zelenetskij, A.V.; Pashchenko, A.B.

    1986-01-01

    The last version of the KOP program for calculating cross sections of neutron and charged particle interaction with atomic nuclei within the scope of the optical model is described. The structure and program organization, library of total parameters of the optical potential, program identificators and peculiarities of its operation, input of source data and output of calculational results for printing are described in detail. The KOP program is described in Fortran- and adapted for EC-1033 computer

  16. Assessment of the available {sup 233}U cross-section evaluations in the calculation of critical benchmark experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.; Wright, R.Q.

    1996-10-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U.S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the S{sub n} transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  17. Assessment of the Available (Sup 233)U Cross Sections Evaluations in the Calculation of Critical Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.

    1993-01-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U. S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  18. Doppler Temperature Coefficient Calculations Using Adjoint-Weighted Tallies and Continuous Energy Cross Sections in MCNP6

    Science.gov (United States)

    Gonzales, Matthew Alejandro

    The calculation of the thermal neutron Doppler temperature reactivity feedback co-efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective temperature derivatives within the system in order to accurately calculate the Doppler temperature feedback. A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) generated continuous energy cross sections within the Monte Carlo N-Particle (MCNP6) transport code. The adjoint-weighted tallies are generated during the continuous energy k-eigenvalue Monte Carlo calculation. The weighting is based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. The adjoint-weighted tallies are produced in a forward calculation and do not require an inversion of the random walk. The OTF cross section database uses a high order functional expansion between points on a user-defined energy-temperature mesh in which the coefficients with respect to a polynomial fitting in temperature are stored. The coefficients of the fits are generated before run- time and called upon during the simulation to produce cross sections at any given energy and temperature. The polynomial form of the OTF cross sections allows the possibility of obtaining temperature derivatives of the cross sections on-the-fly. The use of Monte Carlo sampling of adjoint-weighted tallies and the capability of computing derivatives of continuous energy cross sections with respect to temperature are used to calculate the Doppler temperature coefficient in a research

  19. Facility at CIRUS reactor for thermal neutron induced prompt γ-ray spectroscopic studies

    International Nuclear Information System (INIS)

    Biswas, D.C.; Danu, L.S.; Mukhopadhyay, S.; Kinage, L.A.; Prashanth, P.N.; Goswami, A.; Sahu, A.K.; Shaikh, A.M.; Chatterjee, A.; Choudhury, R.K.; Kailas, S.

    2013-01-01

    A facility for prompt γ-ray spectroscopic studies using thermal neutrons from a radial beam line of Canada India Research Utility Services (CIRUS) reactor, Bhabha Atomic Research Centre (BARC), has been developed. To carry out on-line spectroscopy experiments, two clover germanium detectors were used for the measurement of prompt γ rays. For the first time, the prompt γ–γ coincidence technique has been used to study the thermal neutron induced fission fragment spectroscopy (FFS) in 235 U(n th , f). Using this facility, experiments have also been carried out for on-line γ-ray spectroscopic studies in 113 Cd(n th , γ) reaction

  20. Independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239Pu

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Forman, L.

    1975-01-01

    The relative independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239 Pu have been measured on line using a mass spectrograph and thermalized neutrons from a burst reactor. Independent yields were derived by normalizing the measurements to products of chain yields and fractional independent yields, estimating the latter from measured cumulative yields of Kr and Xe. Comparing the independent yields with those from 238 U fission, the 239 Pu results show shifts in isotopic yield distribution toward lower mass for both Rb and Cs and also toward the production of more Cs and less Rb when 239 Pu is fissioned

  1. The progress of neutron induced prompt gamma analysis technique in 1988-2002

    International Nuclear Information System (INIS)

    Liu Yuren; Jing Shiwei

    2003-01-01

    The new development of the neutron induced prompt gamma-ray analysis (NIPGA) technology in 1988-2002 are described. The pulse fast-thermal neutron activation analysis method, which utilizes the inelastic reaction and capture reaction jointly is employed to measure the elemental content in the material more efficiently. The lifetime of the neutron generator is more than 10000 h and the capability of HPGe, TeZeCd and MCA (multi-channel analyser) reaches the high level. At the same time, Monte Carlo Library least-square method is used to solve the nonlinearity problem in the PGNAA (Prompt Gamma Neutron Activation Analysis)

  2. 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)

  3. Use of the Green function method for calculation of cross section of x-ray photoabsorption for solids

    International Nuclear Information System (INIS)

    Kozhakhmetov, S.K.

    1996-01-01

    Possibility of Green formalism use for calculation of photoabsorption of high-energy x-ray radiation is shown. Analytical expression for photoabsorption cross section is carried out. It does not contains wave functions in explicit form responding to finite states of photoelectron. 5 refs

  4. Comparison of the 12C(e,e'p) cross section at low momentum transfer with a relativistic calculation

    International Nuclear Information System (INIS)

    Tamae, T.; Sato, Y.; Yokokawa, T.; Asano, Y.; Kawabata, M.; Konno, O.; Nakagawa, I.; Nishikawa, I.; Hirota, K.; Yamazaki, H.; Kimura, R.; Miyase, H.; Tsubota, H.; Giusti, C.; Meucci, A.

    2009-01-01

    The (e,e ' p 0 ) cross section of 12 C has been measured at an energy transfer of 60 MeV and a momentum transfer of 104.4 MeV/c using a 197.5 MeV continuous electron beam. The cross section at missing momenta between 181.5 and 304.8 MeV/c obtained from the experiment is compared with theoretical calculations based on the relativistic distorted-wave impulse approximation with and without meson-exchange currents (MEC). The contribution of MEC due to the seagull current is large in the high-missing-momentum region, in particular for the longitudinal component. The cross sections calculated using three different current-conserving operators (cc1, cc2, and cc3) are similar, in contrast to the (γ,p) reaction, where the operators give very different results. The shape of the measured cross section is well described by the calculations, whereas its magnitude is slightly smaller than that described by the calculations.

  5. A comparison of methods for calculating NMR cross-relaxation rates (NOESY and ROESY intensities) in small peptides

    NARCIS (Netherlands)

    Feenstra, K Anton; Peter, Christine; Scheek, Ruud M; van Gunsteren, Wilfred F; Mark, Alan E

    Three methods for calculating nuclear magnetic resonance cross-relaxation rates from molecular dynamics simulations of small flexible molecules have been compared in terms of their ability to reproduce relaxation data obtained experimentally and to produce consistent descriptions of the system. The

  6. Differential α-production cross sections of iron and nickel for 4.3 to 14.1 MeV Neutrons

    International Nuclear Information System (INIS)

    Baba, Mamoru; Ito, Nobuo; Matsuyama, Isamu

    1994-01-01

    The cross section data for neutron-induced α-production are of prime importance in the evaluation of the radiation damage and nuclear heating in fusion and fast reactors. For the evaluation, energy and angular doubly differential cross sections are also required to calculate primary knock-on atom spectra. However, the experimental (n, xα) data are few and discrepant, therefore, the new experimental data are required urgently to improve the accuracy of the (n, xα) cross section data. The authors have measured the double differential (n, xα) cross sections of Fe and Ni in the neutron energy range of 4.3-14.1 MeV using a specially developed gridded ionization chamber. The present work was undertaken as a part of IAEA Coordinated Research Program for neutron-induced He production cross sections. The gridded ionization chamber and the experimental method were reported previously. Three-signals from the common cathode and two anodes were accumulated as two sets of two-dimensional data. The experimental two-dimensional data for the anode and cathode signals were transformed into the double differential cross sections. The results of the double differential cross sections, angular distributions, angle-integrated spectra in the center of mass system and total α-production cross sections are shown. (K.I.)

  7. Calculations of (n,2n) reaction cross sections for Barium isotopes from 5 to 20 MeV

    Science.gov (United States)

    Sahan, Halide; Sahan, Muhittin; Tel, Eyyup

    2017-09-01

    In this study, the excitation functions of (n,2n) reactions for 30,32,34,35,37,38Ba isotopes are calculated using TALYS 1.6, EMPIRE-3.2.2, and ALICE-GDH codes based on statistical model up to 20 MeV. Moreover, the cross section for each isotope have also been estimated at 14.2 MeV using semi empirical formula developed by four different authors. The calculated and estimated cross-sections are compared with experimental cross-sections from EXFOR and compared with the evaluation data in ENDF/B-VII.1 library. Results are close agreement with the experimental data from literature.

  8. Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS

    International Nuclear Information System (INIS)

    Perkasa, Y. S.; Waris, A.; Kurniadi, R.; Su'ud, Z.

    2014-01-01

    Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS have been conducted. In this work, fission cross section resulted from MCNP6 prediction will be compared with result from TALYS calculation. MCNP6 with its event generator CEM03.03 and LAQGSM03.03 have been validated and verified for several intermediate and heavy nuclides fission reaction data and also has a good agreement with experimental data for fission reaction that induced by photons, pions, and nucleons at energy from several ten of MeV to about 1 TeV. The calculation that induced within TALYS will be focused mainly to several hundred MeV for actinide and sub-actinide nuclides and will be compared with MCNP6 code and several experimental data from other evaluator

  9. Cross-section and rate coefficient calculation for electron impact excitation, ionisation and dissociation of H2 and OH molecules

    International Nuclear Information System (INIS)

    Riahi, R.; Ben Lakhdar, Z.; Teulet, Ph.; Gleizes, A.

    2006-01-01

    The weighted total cross-sections (WTCS) theory is used to calculate electron impact excitation, ionization and dissociation cross-sections and rate coefficients of OH, H 2 , OH + , H 2 + , OH - and H 2 - diatomic molecules in the temperature range 1500-15000 K. Calculations are performed for H 2 (X, B, C), OH(X, A, B), H 2 + (X), OH + (X, a, A, b, c), H 2 - (X) and OH - (X) electronic states for which Dunham coefficients are available. Rate coefficients are calculated from WTCS assuming Maxwellian energy distribution functions for electrons and heavy particles. One and 2 temperatures (θ e and θ g respectively for electron and heavy particles kinetic temperatures) results are presented and fitting parameters (a, b and c) are given for each reaction rate coefficient: k(θ) a(θ b )exp(-c/θ). (authors)

  10. The considering of the slowing down effect in the formalism of probability tables. Application to the effective cross section calculation

    International Nuclear Information System (INIS)

    Bouhelal, O.K.A.

    1990-01-01

    The exact determination of the effective multigroup cross sections imposes the numerical solution of the slowing down equation on a very fine energy mesh. Given the complexity of these calculations, different approximation methods have been developed but without a satisfactory treatment of the slowing-down effect. The usual methods are essentially based on interpolations using precalculated tables. The models that use the probability tables allow to reduce the amount of data and the computational effort. A variety of methods proposed by Soviets, then by Americans, and finally the French method, based on the ''moments of a probability distribution'' are incontestably valid within the framework of the statistical hypothesis. This stipulates that the collision densities do not depend on cross section and there is no ambiguity in the effective cross section calculation. The objective of our work is to show that the non statistical phenomena, such as the slowing-down effect which is taken into account, can be described by probability tables which are able to represent the neutronic values and collision densities. The formalism involved in the statistical hypothesis, is based on the Gauss quadrature of the cross sections moments. In the non-statistical hypothesis we introduce the crossed probability tables using the quadratures of double integrals of cross sections, comments. Moreover, a mathematical formalism allowing to establish a relationship between the crossed probability tables and the collision densities was developed. This method was applied on uranium-238 in the range of resolved resonances where the slowing down effect is significant. Validity of the method and the analysis of the obtained results are studied through a reference calculation based on a solution of a discretized slowing down equation using a very fine mesh in which each microgroup can be correctly defined via the statistical probability tables. 42 figs., 32 tabs., 49 refs. (author)

  11. 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

  12. Neutron-induced reactions on U and Th - A new approach via AMS

    International Nuclear Information System (INIS)

    Wallner, A.; Capote, R.; Christl, M.; Fifield, L.K.; Srncik, M.; Tims, S.; Hotchkis, M.; Krasa, A.; Lachner, J.; Lippold, J.; Plompen, A.; Semkova, V.; Steier, P.; Winkler, S.

    2014-01-01

    Recent studies exhibit discrepancies at keV and MeV energies between major nuclear data libraries for 238 U(n,γ), 232 Th(n,γ) and also for (n,xn) reactions. We have extended our initial (n,γ) measurements on 235,238 U to higher neutron energies and to additional reaction channels. Neutron-induced reactions on 232 Th and 238 U were measured by a combination of the activation technique and atom counting of the reaction products using accelerator mass spectrometry (AMS). Natural thorium and uranium samples were activated with quasi-monoenergetic neutrons at IRMM. Neutron capture data were produced for neutron energies between 0.5 and 5 MeV. Fast neutron-induced reactions were studied in the energy range from 17 to 22 MeV. Preliminary data indicate a fair agreement with data libraries; however at the lower band of existing data. This approach represents a complementary method to on-line particle detection techniques and also to conventional decay counting. (authors)

  13. Neutron-induced autoradiography used in the investigation of modern pigments in paintings of known composition

    International Nuclear Information System (INIS)

    Aderhold, H.C.; Taft, W.S.

    1992-01-01

    Neutron-Induced Autoradiography is an effective analytical technique for mapping the location of a number of specified pigments in paintings. Most paintings which have been examined through neutron-induced autoradiography to date were painted prior to the introduction of the most common of modern pigments. By understanding die nuclear properties of these pigments, as revealed by this technique, a more informed analysis of modem paintings may result This investigation is part of an ongoing program to develop case studies for presentation to an undergraduate class at Cornell University, 'Art, Isotopes and Analysis'. We have found that this technique is a graphic and effective method of presenting nuclear reactions and radioactivity to non-specialists. Sample paintings are produced using pigments of known composition. A sequence of discreet layers, each a separate image, is documented in order to establish a reference for accurately interpreting the autoradiographs. The painting is then activated in the Cornell TRIGA reactor and a series of autoradiographs produced Gamma spectra taken before and after each film exposure gives us detailed information on which radioisotopes (and therefore, which pigments), are active. (author)

  14. Implementation of neutron-induced gamma-ray spectroscopy in industrial applications

    International Nuclear Information System (INIS)

    Abernethy, D. A.; Lim, C. S.

    2006-01-01

    Full text: Neutron based analytical techniques are commonly used in a wide variety of industrial applications, with new applications continually being found. As a result, despite popular concerns about the harmful health effects of radiation the number of these analysers is increasing. This is because neutron-induced gamma-ray techniques have the capability of combining elemental sensitivity with significant penetrating power, enabling non-intrusive and non-destructive bulk elemental measurements to be averaged over a large volume of material. Neutron induced gamma ray spectroscopy has been developed by several groups, including CSIRO Minerals, for on-line measurement of elemental composition in a range of industrial applications in vessels, pipes and on conveyor belts. Compared to those typically found in a scientific laboratory, conditions in industrial plants differ substantially in a number of ways, such as environmental variability, operator skill and training, and shielding requirements. As a result of these differences, equipment and techniques which are used as a matter of course in a laboratory often have to undergo major modification to render them suitable for use in an industrial context. This paper will discuss some of the factors that have to be considered when deciding such matters with particular emphasis on the implications of radiation safety requirements

  15. The radioactive inventory of a decommissioned magnox power station structure. 1. Measurements of neutron induced activity in samples from the reactor island

    International Nuclear Information System (INIS)

    Woollam, P.B.

    1978-09-01

    This report is the first of a series which, together, aim to produce an accurate assessment of neutron induced activation levels in the fixed structural components of a reactor of the steel pressure vessel Magnox type. It describes the measurements made of induced activation, necessary in order to establish credibility in the complex calculations described in the subsequent reports. The report also attempts systematically to assess the potential contributions to the dose and disposal problem from all isotopes with a half-life in excess of 5 years. This is necessary in order to ensure that no isotope has been overlooked which could limit any part of the plan for the decommissioning of a Magnox reactor. In addition the report aims to determine concentrations, in each major material type, of trace elements which lead to the isotopes limiting in decommissioning. (author)

  16. Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

    DEFF Research Database (Denmark)

    Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

    2013-01-01

    Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....

  17. Calculations of intersection cross-slip activation energies in fcc metals using nudged elastic band method

    International Nuclear Information System (INIS)

    Rao, S.I.; Dimiduk, D.M.; Parthasarathy, T.A.; El-Awady, J.; Woodward, C.; Uchic, M.D.

    2011-01-01

    The nudged elastic band (NEB) method is used to evaluate activation energies for dislocation intersection cross-slip in face-centered cubic (fcc) nickel and copper, to extend our prior work which used an approximate method. In this work we also extend the study by including Hirth locks (HL) in addition to Lomer-Cottrell locks and glide locks (GL). Using atomistic (molecular statics) simulations with embedded atom potentials we evaluated the activation barrier for a dislocation to transform from fully residing on the glide plane to fully residing on the cross-slip plane when intersecting a 120 o forest dislocation in both Ni and Cu. The initial separation between the screw and the intersecting dislocation on the (1 1 1) glide plane is varied to find a minimum in the activation energy. The NEB method gives energies that are ∼10% lower than those reported in our prior work. It is estimated that the activation energies for cross-slip from the fully glide plane state to the partially cross-slipped state at the 120 o intersection forming GL in Ni and Cu are ∼0.47 and ∼0.65 eV, respectively, and from the fully cross-slip plane state to the partially cross-slipped state forming LC are ∼0.68 and ∼0.67 eV. The activation energies for cross-slip from the fully glide plane state to the partially cross-slipped state at the 120 o intersection forming HL in Ni and Cu are estimated to be ∼0.09 and ∼0.31 eV, respectively. These values are a factor of 3-20 lower than the activation energy for bulk cross-slip in Ni and, a factor of 2-6 lower than the activation energy for cross-slip in Cu estimated by Friedel-Escaig analysis. These results suggest that cross-slip should nucleate preferentially at selected screw dislocation intersections in fcc materials and the activation energies for such mechanisms are also a function of stacking fault energy.

  18. Improvements to the nuclear model code GNASH for cross section calculations at higher energies

    International Nuclear Information System (INIS)

    Young, P.G.; Chadwick, M.B.

    1994-01-01

    The nuclear model code GNASH, which in the past has been used predominantly for incident particle energies below 20 MeV, has been modified extensively for calculations at higher energies. The model extensions and improvements are described in this paper, and their significance is illustrated by comparing calculations with experimental data for incident energies up to 160 MeV

  19. New calculations of cyclotron production cross sections of some positron emitting radioisotopes in proton induced reactions

    International Nuclear Information System (INIS)

    Tel, E.; Aydin, E.G.; Kaplan, A.; Aydin, A.

    2009-01-01

    In this study, new calculations on the excitation functions of 13 C(p, n) 13 N, 14 N (p, α) 11 C, 15 N(p, n) 15 O, 16 O(p, α) 13 N, 18 O(p, n) 18 F, 62 Ni(p, n) 62 Cu, 68 Zn(p, n) 68 Ga and 72 Ge(p, n) 72 As reactions have been carried out in the 5-40 MeV incident proton energy range. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve hybrid model, geometry dependent hybrid model, the cascade exciton model and full exciton model. Equilibrium effects were calculated according to Weisskopf-Ewing model. The calculated results have been compared with experimental data taken from literature. (author)

  20. Influence of differentiation of potential parameters for each excited level of the target nucleus on neutron inelastic cross section calculations

    International Nuclear Information System (INIS)

    Cabezas, R.; Lubian, J.; Moreno, E.

    1992-01-01

    In this paper scattering of neutron in medium mass nuclei (48 < a < 64) at low energies (1-5 Mev) is analyzed. The Hauser-Feshbach-Moldauer formalism and the coupled channel method is used in a combined way. In both cases, the deformed optical potential in the frame of the harmonic vibrational models is considered of integral and total cross section and angular distribution enphasized. It's shown that the use of different set parameters has a mose influence at low energies and represented a contribution of 10% of the calculated cross section with the same potential

  1. Interference analysis of fission cross section

    International Nuclear Information System (INIS)

    Toshkov, S.A.; Yaneva, N.B.

    1976-01-01

    The formula for the reaction cross-section based on the R-matrix formalism considering the interference between the two neighbouring resonances, referred to the same value of total momentum was used for the analysis of the cross-section of resonance neutron induced fission of 230Pu. The experimental resolution and thermal motion of the target nuclei were accounted for numerical integration

  2. Electron ejection cross sections in electron and ion impact ionization: Ab initio and semiempirical calculations

    International Nuclear Information System (INIS)

    Manson, S.T.; Miller, J.H.; Pacific Northwest Lab., Richland, WA)

    1983-01-01

    Ionization cross sections for heavy ions and electrons incident on various atoms and molecules are required in the modeling of the interaction of radiation with matter. For each case, the energy distribution of secondary electrons (the single differential cross section, SDCS) is needed over a broad range of projectile and secondary electron (delta-ray) energies. In many cases the energy and angular distribution of secondary electrons (the double differential cross section, DDCS) is also necessary. Clearly, it would be desirable to have laboratory SDCS and DDCS measurements for all of the cases required. For a variety of reasons, this is not yet possible. Thus, one must turn elsewhere to obtain the needed cross sections. In this paper, we discuss cross sections obtained in two different ways; ab initio theory based on the first Born approximation, and a semi-empirical method based on the Bethe-Born Approximation. In both cases, results on helium will be presented since the largest amount of data is available in this case. Applications of both methods to other target species are given in the references. The accuracy of the methods and plans for the near future are also discussed. 23 references, 6 figures

  3. Cross-sections for homogenized BWR fuel elements in 2d-diffusion theory by 1d-transport calculations

    International Nuclear Information System (INIS)

    Ambrosius, G.

    1980-01-01

    Leakage has a large influence on the thermal spectrum in a fuel rod cell of a BWR and originates: a) from rods with different absorptions and; b) from the different distances to the water gaps. Due to reason a) Gd-rods are treated together with a ring of the homogenized eight nearest neighbours. The often used definition of homogenized cross-sections as the ratio of the integrated reaction rate to the integrated flux proved to be inadequate. This homogenization method is exact as far as the flux is constant over the boundary and as the leakag e during calculating the homogenized cross-sections is similar to that during application. With respect to the condition b) a 1d-transport calculation for the whole fuel element with rings or slabs of homogenized fuel rod cells is performed. With the definition above the flux distribution is that of the fluxes in the moderator regions. The spectrum within each fuel rod cell which includes the leakage is calculated by superimposing at each energy on the flux distribution in the cell the flux at the cell position from the bundle calculation. Changes in the flux ratio between fuel and moderator due to the leakage are taken into account in a final few group 2d-diffusion calculation with fuel and (moderator + cladding) taken separately

  4. Calculation of doses received while crossing a plume of radioactive material

    International Nuclear Information System (INIS)

    Scherpelz, R.I.; Desrosiers, A.E.

    1981-04-01

    A method has been developed for determining the dose received by a person while crossing a plume of radioactive material. The method uses a Gaussian plume model to arrive at a dose rate on the plume centerline at the position of the plume crossing. This dose rate may be due to any external or internal dose pathway. An algebraic formula can then be used to convert the plume centerline dose rate to a total dose integrated over the total time of plume crossing. Correction factors are presented for dose pathways in which the dose rate is not normally distributed about the plume centerline. The method is illustrated by a study done at the Pacific Northwest Laboratory, and results of this study are presented

  5. Basis for calculating cross sections for nuclear magnetic resonance spin-modulated polarized neutron scattering.

    Science.gov (United States)

    Kotlarchyk, Michael; Thurston, George M

    2016-12-28

    In this work we study the potential for utilizing the scattering of polarized neutrons from nuclei whose spin has been modulated using nuclear magnetic resonance (NMR). From first principles, we present an in-depth development of the differential scattering cross sections that would arise in such measurements from a hypothetical target system containing nuclei with non-zero spins. In particular, we investigate the modulation of the polarized scattering cross sections following the application of radio frequency pulses that impart initial transverse rotations to selected sets of spin-1/2 nuclei. The long-term aim is to provide a foundational treatment of the scattering cross section associated with enhancing scattering signals from selected nuclei using NMR techniques, thus employing minimal chemical or isotopic alterations, so as to advance the knowledge of macromolecular or liquid structure.

  6. Calculation of nucleon production cross sections for 200 MeV deuterons

    International Nuclear Information System (INIS)

    Ridikas, D.; Mittig, W.

    1997-01-01

    The differential neutron and proton production cross sections have been investigated for 200 MeV incident deuterons on thin and thick 9 Be, 56 Fe and 238 U targets using the LAHET code system. The examples of the deuteron beam on different target materials are analysed to determine the differences of converting the energy of the beam into the nucleons produced. Both double differential, energy and angle integrated nuclear production cross sections are presented together with the average nucleon multiplicities per incident deuteron. (K.A.)

  7. Calculation of nucleon production cross sections for 200 MeV deuterons

    Energy Technology Data Exchange (ETDEWEB)

    Ridikas, D.; Mittig, W.

    1997-12-31

    The differential neutron and proton production cross sections have been investigated for 200 MeV incident deuterons on thin and thick {sup 9}Be, {sup 56}Fe and {sup 238}U targets using the LAHET code system. The examples of the deuteron beam on different target materials are analysed to determine the differences of converting the energy of the beam into the nucleons produced. Both double differential, energy and angle integrated nuclear production cross sections are presented together with the average nucleon multiplicities per incident deuteron. (K.A.). 31 refs.

  8. Review and calculation of Mott scattering cross section by unscreened point nuclei

    International Nuclear Information System (INIS)

    Idoeta, R.; Legarda, F.

    1992-01-01

    A new tabulation of the ratio of the ''exact'' Mott cross section for unscreened point nuclei to the classical Rutherford cross section for electrons and positions has been made. Because of the infinite slowly converging series appearing in this ratio we have made two series transformations. With this evaluation the ratio reached convergence within six significant figures after less than a hundred terms and very low computing time. So the ratios evaluated have less relative error than those in the literature and covers a greater range of energy and atomic number. (orig.)

  9. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  10. Accurate calculation of the differential cross section of compton scattering with electron mixed chain propagator in SM

    International Nuclear Information System (INIS)

    Chen Xuewen; Fang Zhenyun; Shi Chengye

    2012-01-01

    By using the electroweak standard model (SM), we analyzed the framework of electron mixed chain propagator which composed of serious of different physical loops participating in electroweak interaction and completed the relevant analytical calculation. Then, we obtained the analytical result of electron mixed chain propagator. By applying our result to Compton scattering, the differential cross section of Compton scattering dσ SM (chain) /dcosθ is counted accurately. This result is compared with the lowest order differential cross section dσ (tree) /dcosθ and the electronic chain propagator Compton scattering differential cross section dσ QED (chain) /dcosθ in quantum electrodynamics (QED). It can be seen that dσ SM (chain ) /dcosθ can show the radiation correction more subtly than dσ QED (chain) /dcosθ. (authors)

  11. Analytical formulas for calculation of K X-ray production cross sections by alpha ions

    International Nuclear Information System (INIS)

    Abdellatif, A.; Kahoul, A.; Deghfel, B.; Nekkab, M.; Medjadi, D.E.

    2012-01-01

    In the present study, different procedures are followed to deduce the semi-empirical and the empirical K X-rayX-ray production cross sections induced by alpha ions from the available experimental data and the theoretical results of the ECPSSR model for elements with 20≤Z≤30. The deduced K X-ray production cross sections are compared with predictions from ECPSSR model and with other earlier works. Generally, the deduced K X-ray production cross sections obtained by fitting the available experimental data for each element separately give the most reliable values than those obtained by a global fit. - Highlights: ► The results were presented for elements with atomic numbers 20≤Z≤30 by alpha impact. ► The present semi-empirical formulas were derived from both theoretical and experimental values. ► The available experimental data are directly fitted to deduce the empirical one. ► The results obtained for each element separately give the most reliable values than those obtained by a global fit. ► This procedure is proposed as a black-box way to quickly estimate the cross section.

  12. Statistical Model Analysis of (n, α Cross Sections for 4.0-6.5 MeV Neutrons

    Directory of Open Access Journals (Sweden)

    Khuukhenkhuu G.

    2016-01-01

    Full Text Available The statistical model based on the Weisskopf-Ewing theory and constant nuclear temperature approximation is used for systematical analysis of the 4.0-6.5 MeV neutron induced (n, α reaction cross sections. The α-clusterization effect was considered in the (n, α cross sections. A certain dependence of the (n, α cross sections on the relative neutron excess parameter of the target nuclei was observed. The systematic regularity of the (n, α cross sections behaviour is useful to estimate the same reaction cross sections for unstable isotopes. The results of our analysis can be used for nuclear astrophysical calculations such as helium burning and possible branching in the s-process.

  13. Absolute values of inelastic neutron scattering cross-sections calculated with account taken of the pre-equilibrium mechanism

    International Nuclear Information System (INIS)

    Jahn, H.

    1980-01-01

    Absolute values of secondary energy-dependent inelastic neutron scattering cross sections can be calculated either with the master equation pre-equilibrium formalism of Cline and Blann or with Blann's more recent geometry-dependent hybrid model. The master equation formalism was used at Dubna and Dresden to reproduce experimental results for 14 MeV incident energy. The geometry-dependent hybrid model was used at Karlsruhe to cover for a number of materials the whole range from 5 to 14 MeV incident energy and to reproduce smoothed experimental spectra at 7.45 and 14 MeV. Only the geometry-dependent hybrid model accounts for scattering in the diffuse nuclear surface and thus for a certain average over the direct interaction. It is also free of any fit parameters other than those of the usual optical model. The master equation calculations, on the other hand, are based on nucleon-nucleon scattering cross sections inserted into the high-energy approximation of Kikuchi and Kawai for the intranuclear transition rate. Other approaches require either mass- or energy-dependent or more global fit parameters for a satisfactory reproduction of experimental results, but a genuine prediction of the incident-energy dependence of the inelastic neutron cross section, especially below 14 MeV, is needed for transport and shielding calculations for instance in connection with fusion reactor design studies. (author)

  14. Ab initio calculation of scattering length and cross sections at very low energies for electron-helium scattering

    International Nuclear Information System (INIS)

    Saha, H.P.

    1993-01-01

    The multiconfiguration Hartree-Fock method for continuum wave functions has been used to calculate the scattering length and phase shifts over extremely low energies ranging from 0 to 1 eV very accurately for electron-helium scattering. The scattering length is calculated very accurately with wave functions computed exactly at zero energy, resulting in an upper bound of 1.1784. The electron correlation and polarization of the target by the scattering electron, which are very important in these calculations, have been taken into account in an accurate ab initio manner through the configuration-interaction procedure by optimizing both bound and continuum orbitals simultaneously at each kinetic energy of the scattered electron. Detailed results for scattering length, differential, total, and momentum-transfer cross sections obtained from the phase shifts are presented. The present scattering length is found to be in excellent agreement with the experimental result of Andrick and Bitsch [J. Phys. B 8, 402 (1975)] and the theoretical result of O'Malley, Burke, and Berrington [J. Phys. B 12, 953 (1979)]. There is excellent agreement between the present total cross sections and the corresponding experimental measurements of Buckman and Lohmann [J. Phys. B 19, 2547 (1986)]. The present momentum-transfer cross sections also show remarkable agreement with the experimental results of Crompton, Elford, and Robertson [Aust. J. Phys. 23, 667 (1970)

  15. Model calculations of nuclear data for biologically-important elements

    International Nuclear Information System (INIS)

    Chadwick, M.B.; Blann, M.; Reffo, G.; Young, P.G.

    1994-05-01

    We describe calculations of neutron-induced reactions on carbon and oxygen for incident energies up to 70 MeV, the relevant clinical energy in radiation neutron therapy. Our calculations using the FKK-GNASH, GNASH, and ALICE codes are compared with experimental measurements, and their usefulness for modeling reactions on biologically-important elements is assessed

  16. Introducing MCgrid 2.0: Projecting cross section calculations on grids

    Science.gov (United States)

    Bothmann, Enrico; Hartland, Nathan; Schumann, Steffen

    2015-11-01

    MCgrid is a software package that provides access to interpolation tools for Monte Carlo event generator codes, allowing for the fast and flexible variation of scales, coupling parameters and PDFs in cutting edge leading- and next-to-leading-order QCD calculations. We present the upgrade to version 2.0 which has a broader scope of interfaced interpolation tools, now providing access to fastNLO, and features an approximated treatment for the projection of MC@NLO-type calculations onto interpolation grids. MCgrid 2.0 also now supports the extended information provided through the HepMC event record used in the recent SHERPA version 2.2.0. The additional information provided therein allows for the support of multi-jet merged QCD calculations in a future update of MCgrid.

  17. Neutron induced displacement damage in beryllium in the blanket of a (d,t)-fusion reactor

    International Nuclear Information System (INIS)

    Hermanutz, D.

    1995-09-01

    Beryllium is a favoured candidate for a neutron multiplier in solid breeder blankets of fusion reactors. This is mainly due to its low (n, 2n)-reaction threshold and because of its good thermal and mechanical properties. Its behaviour under intense neutron irradiation, however, is a crucial issue for its use in future fusion reactors. Displacement damage in beryllium so far has been calculated both with data related and methodological deficiencies. First of all, there is a need to have accurate cross-section data in order to obtain reliable spectra of primary knock-on atoms (PKA's). Furthermore, there are principal restrictions of the NRT-model in general used to calculate secondary displacements initiated by PKA's. The underlying theory of damage-energy (part of kinetic energy of PKA transferred elastically to matrix atoms) according to Lindhard is strictly valid only for medium and heavy mass ions with moderate energies in targets of the same element. In this work improved damage cross-sections and displacement rates (dpa/s) in beryllium have been calculated based on cross-section data from ENDF/B-VI (with a significantly improved (n, 2n)-evaluation) and on an appropriate treatment of damage-energy that is suitable for fusion relevant damage of light mass materials. ''This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Communities within the European Fusion Technology Program''. (orig.)

  18. Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

    International Nuclear Information System (INIS)

    Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.

    2012-01-01

    Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

  19. Fission cross section calculations for 209Bi target nucleus based on fission reaction models in high energy regions

    Directory of Open Access Journals (Sweden)

    Kaplan Abdullah

    2015-01-01

    Full Text Available Implementation of projects of new generation nuclear power plants requires the solving of material science and technological issues in developing of reactor materials. Melts of heavy metals (Pb, Bi and Pb-Bi due to their nuclear and thermophysical properties, are the candidate coolants for fast reactors and accelerator-driven systems (ADS. In this study, α, γ, p, n and 3He induced fission cross section calculations for 209Bi target nucleus at high-energy regions for (α,f, (γ,f, (p,f, (n,f and (3He,f reactions have been investigated using different fission reaction models. Mamdouh Table, Sierk, Rotating Liquid Drop and Fission Path models of theoretical fission barriers of TALYS 1.6 code have been used for the fission cross section calculations. The calculated results have been compared with the experimental data taken from the EXFOR database. TALYS 1.6 Sierk model calculations exhibit generally good agreement with the experimental measurements for all reactions used in this study.

  20. MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data

    International Nuclear Information System (INIS)

    2001-01-01

    1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation

  1. Consultants’ Meeting on Recommended Input Parameters for Fission Cross-Section Calculations. Summary Report

    International Nuclear Information System (INIS)

    Capote Noy, Roberto; Simakov, Stanislav; Goriely, Stephane; Hilaire, Stephane; Iwamoto, Osamu; Kawano, Toshihiko; Koning, Arjan

    2014-12-01

    A Consultants’ Meeting on “Recommended Input Parameters for Fission Cross-Section Calculations” was held at IAEA Headquarters, Vienna, Austria to define the scope, deliverables and appropriate work programme of a possible Coordinated Research Project (CRP) on the subject. Presentations are available online at https://www-nds.iaea.org/indexmeeting-crp/CM-RIPL-fission/. A new CRP was endorsed to recommend a comprehensive set of fission input parameters needed for the modelling of fission cross sections. Special attention will be given to the modelling of photon and nucleon induced reactions on actinides with emphasis on incident energies below 30 MeV. The goals and detailed deliverables of the planned CRP were proposed. A Hauser-Feshbach code intercomparison was recommended. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-15

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

  3. Standard Practice for Ensuring Test Consistency in Neutron-Induced Displacement Damage of Electronic Parts

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice sets forth requirements to ensure consistency in neutron-induced displacement damage testing of silicon and gallium arsenide electronic piece parts. This requires controls on facility, dosimetry, tester, and communications processes that affect the accuracy and reproducibility of these tests. It provides background information on the technical basis for the requirements and additional recommendations on neutron testing. In addition to neutrons, reactors are used to provide gamma-ray pulses of intensities and durations that are not achievable elsewhere. This practice also provides background information and recommendations on gamma-ray testing of electronics using nuclear reactors. 1.2 Methods are presented for ensuring and validating consistency in neutron displacement damage testing of electronic parts such as integrated circuits, transistors, and diodes. The issues identified and the controls set forth in this practice address the characterization and suitability of the radiation environm...

  4. Fast neutrons-induced apoptosis is Fas-independent in lymphoblastoid cells

    International Nuclear Information System (INIS)

    Fischer, Barbara; Benzina, Sami; Jeannequin, Pierre; Dufour, Patrick; Bergerat, Jean-Pierre; Denis, Jean-Marc; Gueulette, John; Bischoff, Pierre L.

    2005-01-01

    We have previously shown that ionizing radiation-induced apoptosis in human lymphoblastoid cells differs according to their p53 status, and that caspase 8-mediated cleavage of BID is involved in the p53-dependent pathway. In the present study, we investigated the role of Fas signaling in caspase 8 activation induced by fast neutrons irradiation in these cells. Fas and FasL expression was assessed by flow cytometry and by immunoblot. We also measured Fas aggregation after irradiation by fluorescence microscopy. We found a decrease of Fas expression after irradiation, but no change in Fas ligand expression. We also showed that, in contrast to the stimulation of Fas by an agonistic antibody, Fas aggregation did not occur after irradiation. Altogether, our data strongly suggest that fast neutrons induced-apoptosis is Fas-independent, even in p53-dependent apoptosis

  5. Neutron-induced particle production in the cumulative and noncumulative regions at intermediate energies

    International Nuclear Information System (INIS)

    Mashnik, S.G.

    1992-01-01

    The first systematic measurements of neutron-induced inclusive production of protons, deuterons, tritons and charged pions on carbon, copper, and bismuth in the bombarding energy range of 300-580 MeV and in the angular interval from 51 deg to 165 deg have been analyzed in the framework of the cascade-exciton model. The role of single-particle scattering, the effects of rescattering, the pre-equilibrium emission and 'coalescence' mechanism in particle production in the cumulative (i.e., kinematically - forbidden for quasi-free intranuclear projectile-nucleon collisions) and noncumulative regions are discussed. A week sensitivity of the inclusive distributions to the specific reaction mechanisms and a need of correlation and polarization measurements are noted. 27 refs.; 12 figs.; 1 tab

  6. Vacuum treatment of CR-39 for the reduction of background in neutron induced autoradiography of boron

    International Nuclear Information System (INIS)

    Freyer, K.; Treutler, H.C.; Dietze, K.; Hunyadi, I.; Csige, I.; Somogyi, G.

    1991-01-01

    The influence of etching time and vacuum processing before, during, and after neutron irradiation on the ''signal/noise'' ratio of the neutron induced autoradiography of boron using CR-39 track detectors has been investigated. The neutron irradiation was carried out in a vacuum chamber using the Cf-252 neutron source of the Central Institute of Isotope and Radiation Research at Leipzig. Hungarian-made CR-39 type track detectors, MA-ND/p and MA-ND/α, produced in different years, are compared. After a few hours of 2 kPa (about 10 -2 Torr) vacuum treatment, the ''signal/noise'' ratio for boron determination is improved remarkably in most cases of the detector and etching-time combinations used. (author)

  7. Vacuum treatment of CR-39 for the reduction of background in neutron induced autoradiography of boron

    Energy Technology Data Exchange (ETDEWEB)

    Freyer, K.; Treutler, H.C.; Dietze, K. (Central Inst. of Isotope and Radiation Research, Leipzig (Germany, F.R.)); Hunyadi, I.; Csige, I.; Somogyi, G. (Magyar Tudomanyos Akademia, Debrecen (Hungary). Atommag Kutato Intezete)

    1991-01-01

    The influence of etching time and vacuum processing before, during, and after neutron irradiation on the ''signal/noise'' ratio of the neutron induced autoradiography of boron using CR-39 track detectors has been investigated. The neutron irradiation was carried out in a vacuum chamber using the Cf-252 neutron source of the Central Institute of Isotope and Radiation Research at Leipzig. Hungarian-made CR-39 type track detectors, MA-ND/p and MA-ND/{alpha}, produced in different years, are compared. After a few hours of 2 kPa (about 10{sup -2} Torr) vacuum treatment, the ''signal/noise'' ratio for boron determination is improved remarkably in most cases of the detector and etching-time combinations used. (author).

  8. Improved adiabatic calculation of muonic-hydrogen-atom cross sections. I. Isotopic exchange and elastic scattering in asymmetric collisions

    International Nuclear Information System (INIS)

    Cohen, J.S.; Struensee, M.C.

    1991-01-01

    The improved adiabatic representation is used in calculations of elastic and isotopic-exchange cross sections for asymmetric collisions of pμ, dμ, and tμ with bare p, d, and t nuclei and with H, D, and T atoms. This formulation dissociates properly, correcting a well-known deficiency of the standard adiabatic method for muonic-atom collisions, and includes some effects at zeroth order that are normally considered nonadiabatic. The electronic screening is calculated directly and precisely within the improved adiabatic description; it is found to be about 30% smaller in magnitude than the previously used value at large internuclear distances and to deviate considerably from the asymptotic form at small distances. The reactance matrices, needed for calculations of molecular-target effects, are given in tables

  9. Analytical calculation of the average scattering cross sections using fourier series

    International Nuclear Information System (INIS)

    Palma, Daniel A.P.; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da

    2009-01-01

    The precise determination of the Doppler broadening functions is very important in different applications of reactors physics, mainly in the processing of nuclear data. Analytical approximations are obtained in this paper for average scattering cross section using expansions in Fourier series, generating an approximation that is simple and precise. The results have shown to be satisfactory from the point-of-view of accuracy and do not depend on the type of resonance considered. (author)

  10. Analytical calculation of the average scattering cross sections using fourier series

    Energy Technology Data Exchange (ETDEWEB)

    Palma, Daniel A.P. [Instituto Federal do Rio de Janeiro, Nilopolis, RJ (Brazil)], e-mail: dpalmaster@gmail.com; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: asilva@con.ufrj.br, e-mail: agoncalves@con.ufrj.br, e-mail: aquilino@lmp.ufrj.br, e-mail: fernando@con.ufrj.br

    2009-07-01

    The precise determination of the Doppler broadening functions is very important in different applications of reactors physics, mainly in the processing of nuclear data. Analytical approximations are obtained in this paper for average scattering cross section using expansions in Fourier series, generating an approximation that is simple and precise. The results have shown to be satisfactory from the point-of-view of accuracy and do not depend on the type of resonance considered. (author)

  11. Application of Gauss quadratures to the calculation of resolved resonance contribution in multigroup cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1989-01-01

    A program (RESQ) based on quadratures that evaluates, from ENDF/B data, the resolved resonance contribution in group-averaged cross sections (capture, fission and scattering) was developed. Single and Multilevel Breit-Wigner parameters are accepted. Constant weighting function and zero Kelvin were considered. To assure convergence, different quadrature orders may be analysed. Results are compared with other codes' reconstruction and integration methods. (author) [pt

  12. Theoretical methods for the calculation of the multiphoton ionisation cross-section of atoms and molecules

    International Nuclear Information System (INIS)

    Moccia, R.

    1991-01-01

    Some of the available theoretical methods to compute the two-photon ionisation cross-section of many-electron systems are reviewed. In particular the problems concerning the computation of (i) reliable approximations for the transition matrix elements and the excitation energies; and (ii) accurate results pertaining to the electronic continuum by the use of L 2 basis functions are considered. (author). 29 refs., 6 figs., 1 tab

  13. Improvement of decay and cross-section data libraries for activation calculations

    International Nuclear Information System (INIS)

    Attaya, H.

    1993-01-01

    A new decay data library has been completed. The new library contains up-to-date decay information (half-lives, branching ratios, decay energies, γ's energies and intensities). Activation responses such as the air and water biological hazard potentials, the waste disposal rating, and the biological dose are also included in this library. Recently developed cross-section libraries have been acquired to be used together with the decay data library

  14. Measurement and calculation of excitation cross sections and level ionization by electron impact

    International Nuclear Information System (INIS)

    Blanco Rames, F.

    1990-01-01

    The experimental and theoretical study of the atomic structure in neutral noble gases is studied in this work. It mainly deals with the determination of total cross sections by electron impact and transition probabilities, including: Chapter 1: Theoretical introduction and the intermediate coupling description obtained for 420 levels of s, p and d configurations in neutral noble gases. Chapter 2: Experimental and theoretical values for electron collision cross sections are obtained for several levels of He, Ne, Ar and Kr. Our results as well as those available from existing bibliography are sumarized and compared. By means of an intermediate coupling treatment a number of regularities is found that provides us with some useful approximate semi empirical expressions. Chapter 3: Determination of lifetime and transition probabilities. Lifetime measurements are carried out by means of laser excitation and multichannel delayed coincidences technique. The experimental setup and electronics are also described. Chapter 4: Details the experimental setup developed for electron cross sections measurement by the optical method. The difficulties of this method and their treatment are also shown. (Author)

  15. The background cross section method for calculating the epithermal neutron spectra

    International Nuclear Information System (INIS)

    Martinez, A.S.

    1983-01-01

    We have developed a new methodology to the multigroup constants calculations, for thermal and fast reactors. The method to obtain the constants is extremely fast and simple, and it avoid repeated computations of the detailed neutron spectrum for different cell configurations (composition, geometry and temperature). (author) [pt

  16. 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)

  17. Development of ionization technique for measurement of fast neutron induced fission products yields of {sup 237}Np

    Energy Technology Data Exchange (ETDEWEB)

    Goverdovski, A.A.; Khryachkov, V.A.; Ketlerov, V.V.; Mitrofanov, V.F.; Ostapenko, Yu.B.; Semenova, N.N.; Fomichev, A.N.; Rodina, L.F. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-03-01

    Twin gridded ionization chamber and corresponding software was designed for measurements of masses, kinetic energies and nuclear charges of fission fragments from fast neutron induced fission of {sup 237}Np. The ionization detector design, electronics, data acquisition and processing system and the test results are presented in this paper. (J.P.N.)

  18. Enrico Fermi's Discovery of Neutron-Induced Artificial Radioactivity:The Recovery of His First Laboratory Notebook

    Science.gov (United States)

    Acocella, Giovanni; Guerra, Francesco; Robotti, Nadia

    . We give a short description of the discovery of the first experimental notebook of Enrico Fermi (1901-1954) on his researches during March and April of 1934 on neutron-induced artificial radioactivity, and we point out its relevance for a proper historical and conceptual understanding of those researches.

  19. On a calculation of nucleon knock-out cross sections in a collision of relativistic nuclei

    International Nuclear Information System (INIS)

    Goryachev, B.I.; Lin'kova, N.V.

    1985-01-01

    It is shown that in the framework of the two-stage model one can obtain knock-out cross sections of the given number of nucleons from the nucleus-target at a certain number of nucleons knocked out from the nucleus-projectile. The first stage is considered as a fast process of nucleon collisions of interacting nuclei which is completed with knock out of one or several nucleons. The second stage-comparatively slow - is related to de-excitation of nuclei-fragments

  20. Kuang's Semi-Classical Formalism for Calculating Electron Capture Cross Sections: A Space- Physics Application

    Science.gov (United States)

    Barghouty, A. F.

    2014-01-01

    Accurate estimates of electroncapture cross sections at energies relevant to the modeling of the transport, acceleration, and interaction of energetic neutral atoms (ENA) in space (approximately few MeV per nucleon) and especially for multi-electron ions must rely on detailed, but computationally expensive, quantum-mechanical description of the collision process. Kuang's semi-classical approach is an elegant and efficient way to arrive at these estimates. Motivated by ENA modeling efforts for apace applications, we shall briefly present this approach along with sample applications and report on current progress.

  1. Configuration-interaction relativistic-many-body-perturbation-theory calculations of photoionization cross sections from quasicontinuum oscillator strengths

    International Nuclear Information System (INIS)

    Savukov, I. M.; Filin, D. V.

    2014-01-01

    Many applications are in need of accurate photoionization cross sections, especially in the case of complex atoms. Configuration-interaction relativistic-many-body-perturbation theory (CI-RMBPT) has been successful in predicting atomic energies, matrix elements between discrete states, and other properties, which is quite promising, but it has not been applied to photoionization problems owing to extra complications arising from continuum states. In this paper a method that will allow the conversion of discrete CI-(R)MPBT oscillator strengths (OS) to photoionization cross sections with minimal modifications of the codes is introduced and CI-RMBPT cross sections of Ne, Ar, Kr, and Xe are calculated. A consistent agreement with experiment is found. RMBPT corrections are particularly significant for Ar, Kr, and Xe and improve agreement with experimental results compared to the particle-hole CI method. As a result, the demonstrated conversion method can be applied to CI-RMBPT photoionization calculations for a large number of multivalence atoms and ions

  2. Microscopic calculation of sub-barrier fusion cross section and barrier distribution using M3Y-type forces

    International Nuclear Information System (INIS)

    Ismail, M.; Ramadan, Kh.A.

    2000-01-01

    The heavy-ion (HI) potential between spherical and deformed nuclei is derived using an M3Y-type nucleon-nucleon (NN) interaction. The calculation of the exchange part of the HI potential was improved by using a finite-range NN exchange force instead of the zero-range pseudo-potential which is usually used in deriving the potential between deformed nuclei. We consider an 154 Sm- 16 O nuclear pair as an example to show the effect of finite range on the nucleus-nucleus potential for different deformation parameters and at different orientation angles of the deformed target nucleus. We calculated the fusion cross section and the barrier distribution in the WKB approximation and studied their dependence on the orientation and deformation of the target nucleus. The variations found due to improving the exchange part enhance the fusion cross section below the Coulomb barrier by a factor of about four. It has been found that both the cross section and the barrier distribution are very sensitive to the deformation parameters at energies below the Coulomb barrier. (author)

  3. Microscopic calculation of sub-barrier fusion cross section and barrier distribution using M3Y-type forces

    Energy Technology Data Exchange (ETDEWEB)

    Ismail, M.; Ramadan, Kh.A. [Physics Department, Faculty of Science, Cairo University, Giza (Egypt)

    2000-10-01

    The heavy-ion (HI) potential between spherical and deformed nuclei is derived using an M3Y-type nucleon-nucleon (NN) interaction. The calculation of the exchange part of the HI potential was improved by using a finite-range NN exchange force instead of the zero-range pseudo-potential which is usually used in deriving the potential between deformed nuclei. We consider an {sup 154}Sm-{sup 16}O nuclear pair as an example to show the effect of finite range on the nucleus-nucleus potential for different deformation parameters and at different orientation angles of the deformed target nucleus. We calculated the fusion cross section and the barrier distribution in the WKB approximation and studied their dependence on the orientation and deformation of the target nucleus. The variations found due to improving the exchange part enhance the fusion cross section below the Coulomb barrier by a factor of about four. It has been found that both the cross section and the barrier distribution are very sensitive to the deformation parameters at energies below the Coulomb barrier. (author)

  4. Calculations of elastic, ionization and total cross sections for inert gases upon electron impact: threshold to 2 keV

    Energy Technology Data Exchange (ETDEWEB)

    Vinodkumar, Minaxi [V P and R P T P Science College, Vallabh Vidyanagar 388 120, Gujarat (India); Limbachiya, Chetan [P S Science College, Kadi 382 715, Gujarat (India); Antony, Bobby [Department of Environmental, Earth and Atmospheric Sciences, University of Massachusetts Lowell, 265 Riverside Street, Lowell, MA 01854-5045 (United States); Joshipura, K N [Department of Physics, Sardar Patel University, Vallabh Vidyanagar 388 120, Gujarat (India)

    2007-08-28

    In this paper we report comprehensive calculations of total elastic (Q{sub el}), total ionization (Q{sub ion}) and total (complete) cross sections (Q{sub T}) for the impact of electrons on inert gases (He, Ne, Ar, Kr and Xe) at energies from about threshold to 2000 eV. We have employed the spherical complex optical potential (SCOP) formalism to evaluate Q{sub el} and Q{sub T} and used the complex spherical potential-ionization contribution (CSP-ic) method to derive Q{sub ion}. The dependence of Q{sub T} on polarizability and incident energy is presented for these targets through an analytical formula. Mutual comparison of various cross sections is provided to show their relative contribution to the total cross sections Q{sub T}. Comparison of Q{sub T} for all these targets is carried out to present a general theoretical picture of collision processes. The present calculations also provide information, hitherto sparse, on the excitation processes of these atomic targets. These results are compared with available experimental and other theoretical data and overall good agreement is observed.

  5. Newly calculated absolute cross-section for the electron-impact ionization of C2H2+

    International Nuclear Information System (INIS)

    Deutsch, H.; Becker, K.; Defrance, P.; Probst, M.; Mark, T.D.; Limtrakul, J.

    2006-01-01

    New measurements of the cross-section for electron impact ionization of the molecular ion C 2 H 2 + have been carried out recently. These data differ significantly from earlier data, because cross-sections corresponding to all the possible dissociative ionization processes were determined. The new data in conjunction with the significant discrepancies between the earlier data and the results of various calculations, which disagreed among themselves by a factor of 3, motivated a renewed attempt to apply the semi-classical Deutsch-Mark (DM) formalism to the calculation of the absolute electron-impact ionization cross-section of this molecular ion. A quantum chemical molecular orbital population analysis for both the neutral molecule and the ion revealed that in the case of C 2 H 2 + the singly occupied molecular orbital (i.e. the 'missing' electron) is highly localized near the site of a C atom in the molecule. This information is explicitly incorporated in our formalism. The results obtained by taking the ionic character directly into account are in excellent agreement with the recent experimental data. (authors)

  6. Measurements and calculations of integral capture cross-sections of structural materials in fast reactor spectra

    International Nuclear Information System (INIS)

    Seth, S.; Brunson, G.; Gmuer, K.; Jermann, M.; McCombie, C.; Richmond, R.; Schmocker, U.

    1979-01-01

    This paper relates the checking of integral data of steel and iron in fast reactor lattices. The fully-rodded GCFR benchmark lattice of the zero-energy reactor PROTEUS was successively modified by replacing the PuO 2 -UO 2 fuel rods by steel-18/8 or steel-37 (iron) rods. The neutron spectra of the modified lattices in fact have median energies close to that of a typical LMFBR. The replacement of fuel by the structural material of interest was such that in each case the value of k(infinity) was reduced to near-unity. This allowed the measurement of the lattice-k(infinity) by the null-reactivity technique. In addition, the principal reaction rates (namely U238 capture and fission, relative to Pu239 fission) and the neutron spectrum were measured. These directly measured integral data which are particularly sensitive to the steel cross-sections can be used for the checking and systematic adjustment of data sets. The results may also be analysed so as to derive specific values for the integral capture cross-sections of steel and iron. Neutron balance equations were set-up for each of the lattices using the measured k(infinity) and reaction rates

  7. The development of a collapsing method for the mixed group and point cross sections and its application on multi-dimensional deep penetration calculations

    International Nuclear Information System (INIS)

    Bor-Jing Chang; Yen-Wan H. Liu

    1992-01-01

    The HYBRID, or mixed group and point, method was developed to solve the neutron transport equation deterministically using detailed treatment at cross section minima for deep penetration calculations. Its application so far is limited to one-dimensional calculations due to the enormous computing time involved in multi-dimensional calculations. In this article, a collapsing method is developed for the mixed group and point cross section sets to provide a more direct and practical way of using the HYBRID method in the multi-dimensional calculations. A testing problem is run. The method is then applied to the calculation of a deep penetration benchmark experiment. It is observed that half of the window effect is smeared in the collapsing treatment, but it still provide a better cross section set than the VITAMIN-C cross sections for the deep penetrating calculations

  8. IAEA CIELO Evaluation of Neutron-induced Reactions on 235U and 238U Targets

    Science.gov (United States)

    Capote, R.; Trkov, A.; Sin, M.; Pigni, M. T.; Pronyaev, V. G.; Balibrea, J.; Bernard, D.; Cano-Ott, D.; Danon, Y.; Daskalakis, A.; Goričanec, T.; Herman, M. W.; Kiedrowski, B.; Kopecky, S.; Mendoza, E.; Neudecker, D.; Leal, L.; Noguere, G.; Schillebeeckx, P.; Sirakov, I.; Soukhovitskii, E. S.; Stetcu, I.; Talou, P.

    2018-02-01

    Evaluations of nuclear reaction data for the major uranium isotopes 238U and 235U were performed within the scope of the CIELO Project on the initiative of the OECD/NEA Data Bank under Working Party on Evaluation Co-operation (WPEC) Subgroup 40 coordinated by the IAEA Nuclear Data Section. Both the mean values and covariances are evaluated from 10-5 eV up to 30 MeV. The resonance parameters of 238U and 235U were re-evaluated with the addition of newly available data to the existing experimental database. The evaluations in the fast neutron range are based on nuclear model calculations with the code EMPIRE-3.2 Malta above the resonance range up to 30 MeV. 235U(n,f), 238U(n,f), and 238U(n,γ) cross sections and 235U(nth,f) prompt fission neutron spectrum (PFNS) were evaluated within the Neutron Standards project and are representative of the experimental state-of-the-art measurements. The Standards cross sections were matched in model calculations as closely as possible to guarantee a good predictive power for cross sections of competing neutron scattering channels. 235U(n,γ) cross section includes fluctuations observed in recent experiments. 235U(n,f) PFNS for incident neutron energies from 500 keV to 20 MeV were measured at Los Alamos Chi-Nu facility and re-evaluated using all available experimental data. While respecting the measured differential data, several compensating errors in previous evaluations were identified and removed so that the performance in integral benchmarks was restored or improved. Covariance matrices for 235U and 238U cross sections, angular distributions, spectra and neutron multiplicities were evaluated using the GANDR system that combines experimental data with model uncertainties. Unrecognized systematic uncertainties were considered in the uncertainty quantification for fission and capture cross sections above the thermal range, and for neutron multiplicities. Evaluated files were extensively benchmarked to ensure good performance in

  9. A flow-based methodology for the calculation of TSO to TSO compensations for cross-border flows

    International Nuclear Information System (INIS)

    Glavitsch, H.; Andersson, G.; Lekane, Th.; Marien, A.; Mees, E.; Naef, U.

    2004-01-01

    In the context of the development of the European internal electricity market, several methods for the tarification of cross-border flows have been proposed. This paper presents a flow-based method for the calculation of TSO to TSO compensations for cross-border flows. The basic principle of this approach is the allocation of the costs of cross-border flows to the TSOs who are responsible for these flows. This method is cost reflective, non-transaction based and compatible with domestic tariffs. It can be applied when limited data are available. Each internal transmission network is then modelled as an aggregated node, called 'supernode', and the European network is synthesized by a graph of supernodes and arcs, each arc representing all cross-border lines between two adjacent countries. When detailed data are available, the proposed methodology is also applicable to all the nodes and lines of the transmission network. Costs associated with flows transiting through supernodes or network elements are forwarded through the network in a way reflecting how the flows make use of the network. The costs can be charged either towards loads and exports or towards generations and imports. Combination of the two charging directions can also be considered. (author)

  10. CANDU radiotoxicity inventories estimation: A calculated experiment cross-check for data verification and validation

    International Nuclear Information System (INIS)

    Pavelescu, Alexandru Octavian; Cepraga, Dan Gabriel

    2007-01-01

    This paper is related to the Clearance Potential Index, Ingestion and Inhalation Hazard Factors of the nuclear spent fuel and radioactive wastes. This study required a complex activity that consisted of various phases such us: the acquisition, setting up, validation and application of procedures, codes and libraries. The paper reflects the validation phase of this study. Its objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from a Pickering CANDU reactor with inventories predicted using a recent version of the ORIGEN-ARP from SCALE 5 coupled with the time dependent cross sections library, CANDU 28.lib, produced by the sequence SAS2H of SCALE 4.4a. In this way, the procedures, codes and libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors are being qualified and validated, in support for the safety management of the radioactive wastes. (authors)

  11. Ab initio calculation of the cross sections for electron impact vibrational excitation of CO via the {sup 2}Π shape resonance

    Energy Technology Data Exchange (ETDEWEB)

    Falcetta, Michael F., E-mail: mffalcetta@gcc.edu; Fair, Mark C.; Tharnish, Emily M.; Williams, Lorna M.; Hayes, Nathan J. [Department of Chemistry, Grove City College, Grove City, Pennsylvania 16127 (United States); Jordan, Kenneth D. [Department of Chemistry, University of Pittsburgh, Pittsburgh, Pennsylvania 15260 (United States)

    2016-03-14

    The stabilization method is used to calculate the complex potential energy curve of the {sup 2}Π state of CO{sup −} as a function of bond length, with the refinement that separate potentials are determined for p-wave and d-wave attachment and detachment of the excess electron. Using the resulting complex potentials, absolute vibrational excitation cross sections are calculated as a function of electron energy and scattering angle. The calculated cross sections agree well with experiment.

  12. Deuteron and neutron induced activation in the Eveda accelerator materials: implications for the accelerator maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, M.; Sanz, J.; Garcia, N.; Cabellos, O. [Madrid Univ. Politecnica, C/ Jose Gutierrez Abascal, lnstituto de Fusion Nuclear (Spain); Sauvan, R. [Universidad Nacional de Educacion a Distancia (UNED), Madrid (Spain); Moreno, C.; Sedano, L.A. [CIEMAT-Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, Association Euratom-CIEMAT, Madrid (Spain)

    2007-07-01

    Full text of publication follows: The IFMIF (International Fusion Materials Irradiation Facility) is an accelerator-based DLi neutron source designed to test fusion reactor candidate materials for high fluence neutrons. Before deciding IFMIF construction, an engineering design and associated experimental data acquisition, defined as EVEDA, has been proposed. Along the EVEDA accelerator, deuteron beam losses collide with the accelerator materials, producing activation and consequent radiations responsible of dose. Calculation of the dose rates in the EVEDA accelerator room is necessary in order to analyze the feasibility for manual maintenance. Dose rates due to the activation produced by the deuteron beam losses interaction with the accelerator materials, will be calculated with the ACAB activation code, using EAF2007 library for deuteron activation cross-sections. Also, dose rates from the activation induced by the neutron source produced by the interaction of deuteron beam losses with the accelerator materials and the deuterium implanted in the structural lattice, will be calculated with the SRIM2006, TMAP7, DROSG2000/NEUYIE, MCNPX and ACAB codes. All calculations will be done for the EVEDA accelerator with the room temperature DTL structure, which is based on copper cavities for the DTL. Some calculations will be done for the superconducting DTL structure, based on niobium cavities for the DTL working at cryogenic temperature. Final analysis will show the dominant mechanisms and major radionuclides contributing to the surface dose rates. (authors)

  13. The K + - Nucleus Microscopic Optical Potential and Calculations of the Corresponding Differential Elastic and Total Reaction Cross Sections

    International Nuclear Information System (INIS)

    Zemlyanaya, E.V.; Lukyanov, K.V.; Lukyanov, V.K.; Hanna, K.M.

    2009-01-01

    The microscopic optical potential (OP) is calculated for the K+-meson scattering on the 12 C and 40 Ca nuclei at intermediate energies. This potential has no free parameters and based on the known kaon-nucleon amplitude and nuclear density distribution functions. Then, the Klein-Gordon equation is written in the form of the relativistic Schrodinger equation where terms quadratic in the potential was estimated can be neglected. The latter equations adapted to the considered task and solved numerically. The effect of revitalization is shown to play a significant role. A good agreement with the experimental data on differential elastic cross sections is obtained. However, to explain the data on total reaction cross sections the additional surface term of OP was introduced to account for influence of the peripheral nuclear reaction channels

  14. CRSEC: a general purpose Hauser--Feshbach code for the calculation of nuclear cross-sections and thermonuclear reaction rates

    International Nuclear Information System (INIS)

    Woosley, S.; Fowler, W.A.

    1977-09-01

    CRSEC is a FORTRAN IV computer code designed for the efficient calculation of average nuclear cross sections in situations where a statistical theory of nuclear reactions is applicable and where compound nuclear formation is the dominant reaction mechanism. This code generates cross sections of roughly factor of 2 accuracy for incident particle energies in the range of 10 keV to 10 MeV for most target nuclei from magnesium to bismuth. Exceptions usually involve reactions that enter the compound nucleus at such a low energy that fewer than 10 levels are present in the ''energy window of interest.'' The incident particle must be a neutron, proton, or alpha particle, and only binary reactions resulting in the emission of a single n, p, α, or γ (cascade) are calculated. CRSEC is quite fast, a complete calculation of 12 different reactions over a grid of roughly 150 energy points and the generation of Maxwellian averaged rates taking about 30 seconds of CDC7600 time. Also the semi-empirical parameterization of nuclear properties contained in CRSEC is very general. Greater accuracy may be obtained, however, by furnishing specific low-lying excited states, level density parameterization, and nuclear strength functions. A more general version of CRSEC, called CRSECI, is available that conserves isospin properly in all reactions and allows the user to specify a given degree of isospin mixing in the highly excited states of the compound nucleus. Besides the cross section as a function of center-of-mass energy, CRSEC also generates the Maxwell--Boltzmann averaged thermonuclear reaction rate and temperature dependent nuclear partition function for a grid of temperatures from 10 8 to 10 10 0 K. Sections of this report describe in greater detail the physics employed in CRSEC and how to use the code. 2 tables

  15. Modelisation of the fission cross section

    International Nuclear Information System (INIS)

    Morariu, Claudia

    2013-03-01

    The neutron cross sections of four nuclear systems (n+ 235 U, n+ 233 U, n+ 241 Am and n+ 237 Np) are studied in the present document. The target nuclei of the first case, like 235 U and 239 Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237 Np and 241 Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author) [fr

  16. Statistical model analysis of fast-neutron-induced fission of U isotopes

    International Nuclear Information System (INIS)

    Lestone, J.P.; Gavron, A.

    1994-01-01

    We have obtained the first experimental evidence of the washing out of the collective level density enhancement associated with the amma deformation of the triaxial first barrier in the U isotope cross sections at neutron energies up to ∼20 MeV, with a statistical model which uses level densities obtained from Nilsson model singles particle levels, we find that it is necessary to (1) wash out the triaxial level density enhancement at an excitation energy of ∼7 MeV above the triaxial barriers with a width of ∼1 MeV, and (2) incorporate the effects of preequilibrium emission. These results imply a γ deformation of the first barriers in the range 10 degree--20 degree. Above an incoming neutron energy of ∼20 MeV where insufficient data exist to constrain optical model potentials, our statistical model U(n,f) cross sections increasingly overestimate the experimental data. A satisfactory reproduction of all the available U(n,f) cross sections above ∼20 MeV is obtained by scaling our calculated compound nucleus formation cross sections. This scaling factor falls from 1.0 at ∼20 MeV to 0.82 at 100 MeV

  17. HAUFES : a FORTRAN code for the calculation of compound nuclear cross-sections by Hauser-Feshbach theory

    International Nuclear Information System (INIS)

    Viyogi, Y.P.; Ganguly, N.K.

    1975-01-01

    The FORTRAN code described in the report has been developed for the BESM-6 computer with a view to calculate the cross-section of reactions proceeding via the formation of compound nucleus for all open two-body reaction channels using Hauser-Feshbach theory with Moldauer's correction for the fluctuation of level widths. The code can also be used to analyse data from 'crystal blocking' experiments to obtain nuclear level densities. The report describes the input-output specifications along with a short account of the algorithm of the program. (author)

  18. New approximations for the interference term applied to the calculation of scattering cross section of the {sup 238} U isotope

    Energy Technology Data Exchange (ETDEWEB)

    Palma, Daniel Artur Pinheiro [Centro Federal de Educacao Tecnologica de Quimica de Nilopolis, RJ (Brazil)]. E-mails: dpalma@cefeteq.br; Martinez, Aquilino Senra; Goncalves, Alessandro C. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mail: agoncalves@con.ufrj.br; aquilino@lmp.ufrj.br

    2008-07-01

    The calculation of the Doppler broadening function and the interference term are very important in the generation of nuclear data. Recent papers have proposed analytical formulations for both functions and, despite their being simple and precise, they contain the error function with a complex argument. With the intention of simplifying the mathematical treatment two approximations are proposed in this paper. The first one consists of using an expansion in the form of series to treat the error function. The other approximation is based on simplifications in the differential equations that govern the Doppler broadening function. For validation purpose the result obtained is compared to the one obtained in the calculation of the cross sections for isotope {sup 238}U for different resonances. Results obtained have proved satisfactory from the standpoint of accuracy. (author)

  19. New approximations for the interference term applied to the calculation of scattering cross section of the 238 U isotope

    International Nuclear Information System (INIS)

    Palma, Daniel Artur Pinheiro; Martinez, Aquilino Senra; Goncalves, Alessandro C.

    2008-01-01

    The calculation of the Doppler broadening function and the interference term are very important in the generation of nuclear data. Recent papers have proposed analytical formulations for both functions and, despite their being simple and precise, they contain the error function with a complex argument. With the intention of simplifying the mathematical treatment two approximations are proposed in this paper. The first one consists of using an expansion in the form of series to treat the error function. The other approximation is based on simplifications in the differential equations that govern the Doppler broadening function. For validation purpose the result obtained is compared to the one obtained in the calculation of the cross sections for isotope 238 U for different resonances. Results obtained have proved satisfactory from the standpoint of accuracy. (author)

  20. Neutron capture cross section measurements and theoretical calculation for the {sup 186}W(n,γ){sup 187}W reaction

    Energy Technology Data Exchange (ETDEWEB)

    Al-abyad, Mogahed; Mohamed, Gehan Y. [Atomic Energy Authority, Cairo (Egypt). Experimental Nuclear Physics Dept.

    2017-08-01

    Neutron capture cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the reaction {sup 186}W(n,γ){sup 187}W were measured experimentally using the research reactor (ETRR-2) and an Am-Be neutron source, also calculated using TALYS-1.6 code. The present results of σ{sub 0} are (39.08±2.6, 38.75±0.98 and 38.33 barn) and I{sub 0} are (418.5±74, 439.3±36 and 445.5 barn) by using the reactor, neutron source and TALYS-1.6, respectively. The present results are in acceptable agreement with most of the previous experimental and evaluated data as well as the theoretical calculations.

  1. Calculation of the positronium formation differential cross section for collision of electron with anti-hydrogen atoms

    International Nuclear Information System (INIS)

    Ghanbari Adivi, E.; Kanjuri, F.; Bolorizadeh, M.

    2006-01-01

    The positronium formation differential cross sections in collision of the high-energy but non-relativistic electrons with anti-hydrogen atoms are calculated by using the three-body Faddeev-Watson-Lovelace formalism. In a second-order approximation, the inter-nuclear and nuclear-electronic partial amplitudes therein the Faddeev-Watson series are calculated, analytically, in the range of 0-180 degrees of the scattering angles. The presence of the T homas peak a t 45 d egree i s investigated. The results are discussed for 1 and 10 keV impact energies and for electron transition from anti-hydrogen ground state into the different states therein the K-, L- and M- shells of the positronium atoms.

  2. Calculation of total cross sections and effective emission coefficients for B5+ collisions with ground-state and excited hydrogen

    International Nuclear Information System (INIS)

    Guzman, F; Errea, L F; Illescas, Clara; Mendez, L; Pons, B

    2010-01-01

    Classical and semiclassical calculations of nl-resolved charge exchange cross sections in B 5 + collisions with H(n i ) are performed to compute effective emission coefficients for the n = 7 → n = 6 transition in B 4 + for plasma conditions typical of the ASDEX-U tokamak. For n i = 1, the value of the emission coefficient is larger than that obtained from ADAS database by a factor of 2 at energies of 10 keV amu -1 , but no differences are found at energies above 50 keV amu -1 . For n i = 2, our calculation yields emission coefficients close to those derived from ADAS data from low to high impact energies. The emission coefficients corresponding to B 5 + + H(n i = 3) collisions are of the same order of magnitude as those for n i = 2.

  3. Generation and Verification of ENDF/B-VII.0 Cross section Libraries for Monte Carlo Calculations

    International Nuclear Information System (INIS)

    Park, Ho Jin; Kwak, Min Su; Joo, Han Gyu; Kim, Chang Hyo

    2007-01-01

    For Monte Carlo neutronics calculations, a continuous energy nuclear data library is needed. It can be generated from various evaluated nuclear data files such as ENDF/B using the ACER routine of the NJOY.code after a series of prior processing involving various other NJOY routines. Recently, a utility code, which generates the NJOY input decks in an automated mode, named ANJOYMC became available. The use of this code greatly reduces the user's effort and the possibility of input errors. In December 2006, the initial version of the ENDF/BVII nuclear data library was released. It was reported that the new data files have much better data which reduces the errors noted in the previous versions. Thus it is worthwhile to examine the performance of the new data files particularly using an independent Monte Carlo code, MCCARD and the ANJOYMC utility code. The verification of the newly generated library can be readily performed by analyzing numerous standard criticality benchmark problems

  4. Calculation and evaluation of the activation cross sections for 187Re(n,2n)186m,gRe reactions

    International Nuclear Information System (INIS)

    Huang Xiaolong; Lu Hanlin; Zhou Chunmei

    1998-01-01

    The activation cross sections for 187 Re(n,2n) 186m,g Re reactions are calculated using UNF code. The calculations are in good agreement with the re-evaluated measured data. Finally the excitation function for 187 Re(n,2n) 186m,g Re reactions are evaluated and recommended based on present calculations and evaluated decay data

  5. First-principles method for calculating the rate constants of internal-conversion and intersystem-crossing transitions.

    Science.gov (United States)

    Valiev, R R; Cherepanov, V N; Baryshnikov, G V; Sundholm, D

    2018-02-28

    A method for calculating the rate constants for internal-conversion (k IC ) and intersystem-crossing (k ISC ) processes within the adiabatic and Franck-Condon (FC) approximations is proposed. The applicability of the method is demonstrated by calculation of k IC and k ISC for a set of organic and organometallic compounds with experimentally known spectroscopic properties. The studied molecules were pyrromethene-567 dye, psoralene, hetero[8]circulenes, free-base porphyrin, naphthalene, and larger polyacenes. We also studied fac-Alq 3 and fac-Ir(ppy) 3 , which are important molecules in organic light emitting diodes (OLEDs). The excitation energies were calculated at the multi-configuration quasi-degenerate second-order perturbation theory (XMC-QDPT2) level, which is found to yield excitation energies in good agreement with experimental data. Spin-orbit coupling matrix elements, non-adiabatic coupling matrix elements, Huang-Rhys factors, and vibrational energies were calculated at the time-dependent density functional theory (TDDFT) and complete active space self-consistent field (CASSCF) levels. The computed fluorescence quantum yields for the pyrromethene-567 dye, psoralene, hetero[8]circulenes, fac-Alq 3 and fac-Ir(ppy) 3 agree well with experimental data, whereas for the free-base porphyrin, naphthalene, and the polyacenes, the obtained quantum yields significantly differ from the experimental values, because the FC and adiabatic approximations are not accurate for these molecules.

  6. Calculations of the main free path on neutron emission cross-section for spallation reaction of target and fuel nuclei

    International Nuclear Information System (INIS)

    Tel, E.; Kisoglu, H. F.; Topaksu, A. K.; Aydin, A.; Kaplan, A.

    2007-01-01

    There are several new technological application fields of fast neutrons such as accelerator-driven incineration/ transmutation of the long-lived radioactive nuclear wastes (in particular transuranium nuclides) to short-lived or stable isotopes by secondary spallation neutrons produced by high-intensity, intermediate-energy, charged-particle beams, prolonged planetary space missions, shielding for particle accelerators. Especially, accelerator driven subcritical systems (ADS) can be used for fission energy production and /or nuclear waste transmutation as well as in the intermediate-energy accelerator driven neutron sources, ions and neutrons with energies beyond 20 MeV, the upper limit of exiting data files that produced for fusion and fission applications. In these systems, the neutron scattering cross sections and emission differential data are very important for reactor neutronics calculations. The transition rate calculation involves the introduction of the parameter of mean free path determines the mean free path of the nucleon in the nuclear matter. This parameter allows an increase in mean free path, with simulation of effect, which is not considered in the calculations, such as conservation of parity and angular momentum in intra nuclear transitions. In this study, we have investigated the multiple preequilibrium matrix element constant from internal transition for Uranium, Thorium, (n,xn) neutron emission spectra. The neutron-emission spectra produced by (n,xn) reactions on nuclei of some target (for spallation) have been calculated. In the calculations, we have used the geometry dependent hybrid model and the cascade exciton model including the effects of the preequilibrium. The pre-equilibrium direct effects have been examined by using full exciton model. All calculated results have been compared with the experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other

  7. Calculated sputtering and atomic displacement cross-sections for applications to medium voltage analytical electron microscopy

    International Nuclear Information System (INIS)

    Bradley, C.R.; Zaluzec, N.J.

    1987-08-01

    The development of medium voltage electron microscopes having high brightness electron sources and ultra-high vacuum environments has been anticipated by the microscopy community now for several years. The advantages of such a configuration have been discussed to great lengths, while the potential disadvantages have for the most part been neglected. The most detrimental of these relative to microcharacterization are the effects of electron sputtering and atomic displacement to the local specimen composition. These effects have in the past been considered mainly in the high voltage electron microscope regime and generally were ignored in lower voltage instruments. Recent experimental measurements have shown that the effects of electron sputtering as well as radiation induced segregation can be observed in conventional transmission electron microscopes. It is, therefore, important to determine at what point the effects will begin to manifest themselves in the new generation of medium voltage analytical electron microscopes. In this manuscript we present new calculations which allow the individual experimentalist to determine the potential threshold levels for a particular elemental system and thus avoid the dangers of introducing artifacts during microanalysis. 12 refs., 3 figs

  8. Excitation and charge transfer in He++H collisions. A study of the origin dependence of calculated cross sections

    International Nuclear Information System (INIS)

    Macias, A.; Riera, A.; Yanez, M.

    1983-01-01

    A treatment of the He + -H collision is presented in an impact-parameter formalism for collision energies 0.5--30 keV. The origin dependence of the calculated total cross sections is studied in detail. It is shown that the branching ratio between reactions He + (1s)+H(1s)→He + (1s)+H(2n) and He + (1s)+H(1s)→He(1s2p)+H + oscillates as a function of the origin of electronic coordinates chosen in the calculation. This oscillation is strong enough so that at nuclear velocity 0.5 a.u., either both reactions are competitive or one of them can have a cross section for the reaction He(1s 2 )+H + →He + (1s)+H(1s) can either be negligble or comparable to those of the other reactions. We study the oscillatory behavior of the charge-exchange-transition probability as a function of 1/v. We show the similarity, for high velocity, between nonresonant and resonant change-exhange processes, the origin of the damping factor, and the influence of the rotatioal coupling on the transition probabilities as functions of 1/v. A connection between Lichten's and Denkov's models is established

  9. Excitation and charge transfer in He/sup +/+H collisions. A study of the origin dependence of calculated cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Macias, A.; Riera, A.; Yanez, M.

    1983-01-01

    A treatment of the He/sup +/-H collision is presented in an impact-parameter formalism for collision energies 0.5--30 keV. The origin dependence of the calculated total cross sections is studied in detail. It is shown that the branching ratio between reactions He/sup +/(1s)+H(1s)..-->..He/sup +/(1s)+H(2n) and He/sup +/(1s)+H(1s)..-->..He(1s2p)+H/sup +/ oscillates as a function of the origin of electronic coordinates chosen in the calculation. This oscillation is strong enough so that at nuclear velocity 0.5 a.u., either both reactions are competitive or one of them can have a cross section for the reaction He(1s/sup 2/)+H/sup +/..-->..He/sup +/(1s)+H(1s) can either be negligble or comparable to those of the other reactions. We study the oscillatory behavior of the charge-exchange-transition probability as a function of 1/v. We show the similarity, for high velocity, between nonresonant and resonant change-exhange processes, the origin of the damping factor, and the influence of the rotatioal coupling on the transition probabilities as functions of 1/v. A connection between Lichten's and Denkov's models is established.

  10. Analysis methods of neutrons induced resonances in the transmission experiments by time-of-flight and automation of these methods on IBM 7094 II computer; Methode d'analyse des resonances induites par les neutrons dans les experiences de transmission par temps-de-vol et automatisation de ces methodes sur ordinateur IBM-7094 II

    Energy Technology Data Exchange (ETDEWEB)

    Corge, C

    1967-07-01

    The neutron induced resonances analysis aims to determine the neutrons characteristics, leading to the excitation energies, de-excitation probabilities by gamma radiation emission, by neutron emission or by fission, their spin, their parity... This document describes the methods developed, or adapted, the calculation schemes and the algorithms implemented to realize such analysis on a computer, from data obtained during time-of-flight experiments on the linear accelerator of Saclay. (A.L.B.)

  11. Analysis methods of neutrons induced resonances in the transmission experiments by time-of-flight and automation of these methods on IBM 7094 II computer; Methode d'analyse des resonances induites par les neutrons dans les experiences de transmission par temps-de-vol et automatisation de ces methodes sur ordinateur IBM-7094 II

    Energy Technology Data Exchange (ETDEWEB)

    Corge, C

    1967-07-01

    The neutron induced resonances analysis aims to determine the neutrons characteristics, leading to the excitation energies, de-excitation probabilities by gamma radiation emission, by neutron emission or by fission, their spin, their parity... This document describes the methods developed, or adapted, the calculation schemes and the algorithms implemented to realize such analysis on a computer, from data obtained during time-of-flight experiments on the linear accelerator of Saclay. (A.L.B.)

  12. Calculation of a Tunnel Cross Section Subjected to Fire – with a New Advanced Transient Concrete Model for Reinforced Structures

    Directory of Open Access Journals (Sweden)

    U. Schneider

    2009-01-01

    Full Text Available The paper presents the structural application of a new thermal induced strain model for concrete – the TIS-Model. An advanced transient concrete model (ATCM is applied with the material model of the TIS-Model. The non-linear model comprises thermal strain, elastic strain, plastic strain and transient temperature strains, and load history modelling of restraint concrete structures subjected to fire.The calculations by finite element analysis (FEA were done using the SAFIR structural code. The FEA software was basically new with respect to the material modelling derived to use the new TIS-Model (as a transient model considers thermal induced strain. The equations of the ATCM consider a lot of capabilities, especially for considering irreversible effects of temperature on some material properties. By considering the load history during heating up, increasing load bearing capacity may be obtained due to higher stiffness of the concrete. With this model, it is possible to apply the thermal-physical behaviour of material laws for calculation of structures under extreme temperature conditions.A tunnel cross section designed and built by the cut and cover method is calculated with a tunnel fire curve. The results are compared with the results of a calculation with the model of the Eurocode 2 (EC2-Model. The effect of load history in highly loaded structures under fire load will be investigated.A comparison of this model with the ordinary calculation system of Eurocode 2 (EC2 shows that a better evaluation of the safety level was achieved with the new model. This opens a space for optimizing concrete structure design with transient temperature conditions up to 1000 °C. 

  13. The method of covariant calculation of the amplitudes of processes with polarized spin 1/2 particles and its application to calculation of interference terms in cross sections of these processes

    International Nuclear Information System (INIS)

    Bondarev, A.L.

    1993-01-01

    The method of covariant calculation of the amplitudes of processes with polarized spin 1/2 particles is suggested. It can be used for calculation of interference terms in cross sections of these processes. As an illustration the expressions for the lowest order amplitudes of electron-electron scattering and for electron current with radiation of two bremsstrahlung photons in ultrarelativistic limit are presented

  14. 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

  15. Simulated production rates of exotic nuclei from the ion guide for neutron-induced fission at IGISOL

    Energy Technology Data Exchange (ETDEWEB)

    Jansson, Kaj; Al-Adili, Ali; Nilsson, Nicklas; Norlin, Martin; Solders, Andreas [Uppsala University, Department of Physics and Astronomy, Uppsala (Sweden)

    2017-12-15

    An investigation of the stopping efficiency of fission products, in the new ion guide designed for ion production through neutron-induced fission at IGISOL in Jyvaeskylae, Finland, has been conducted. Our simulations take into account the new neutron converter, enabling measurements of neutron-induced fission yields, and thereby provide estimates of the obtained yields as a function of primary proton beam current. Different geometries, targets, and pressures, as well as models for the effective charge of the stopped ions were tested, and optimisations to the setup for higher yields are suggested. The predicted number of ions stopped in the gas lets us estimate the survival probability of the ions reaching the downstream measurements stations. (orig.)

  16. Complex Correlation Kohn-T Method of Calculating Total and Elastic Cross Sections. Part 1; Electron-Hydrogen Elastic Scattering

    Science.gov (United States)

    Bhatia, A. K.; Temkin, A.; Fisher, Richard R. (Technical Monitor)

    2001-01-01

    We report on the first part of a study of electron-hydrogen scattering, using a method which allows for the ab initio calculation of total and elastic cross sections at higher energies. In its general form the method uses complex 'radial' correlation functions, in a (Kohn) T-matrix formalism. The titled method, abbreviated Complex Correlation Kohn T (CCKT) method, is reviewed, in the context of electron-hydrogen scattering, including the derivation of the equation for the (complex) scattering function, and the extraction of the scattering information from the latter. The calculation reported here is restricted to S-waves in the elastic region, where the correlation functions can be taken, without loss of generality, to be real. Phase shifts are calculated using Hylleraas-type correlation functions with up to 95 terms. Results are rigorous lower bounds; they are in general agreement with those of Schwartz, but they are more accurate and outside his error bounds at a couple of energies,

  17. Neutron-induced fission fragment angular distribution at CERN n TOF: The Th-232 case

    CERN Document Server

    Tarrio, Diego; Paradela, Carlos

    This thesis work was done in the frame of the study of the neutron-induced fission of actinides and subactinides at the CERN n TOF facility using a fast Parallel Plate Avalanche Counters (PPACs) setup. This experimental setup provide us with an intense neutron beam with a white spectrum from thermal to 1 GeV and with an outstanding high resolution provided by its flight path of 185 m. In our experiment, fission events were identified by detection of both fission fragments in time coincidence in the two PPAC detectors flanking the corresponding target. This technique allowed us to discriminate the fission events from the background produced by α disintegration of radioactive samples and by particles produced in spallation reactions. Because PPAC detectors are insensitive to the γ flash, it is possible to reach energies as high as 1 GeV. The stripped cathodes provide the spatial position of the hits in the detectors, so that the emission angle of the fission fragments can be measured. Inside the reaction cham...

  18. Symmetry of neutron-induced 235U fission at individual resonances. III

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, G A; Bayhurst, B P; Prestwood, R J; Gilmore, J S; Knobeloch, G W [Los Alamos Scientific Laboratory, University of California, Los Alamos, NM (United States)

    1970-05-15

    A number of experiments have been described in recent years which document variations in the yields of symmetric or near-symmetric fission products at resonances in 235-U and 239-Pu neutron-induced fission. In the case of 239-Pu fission it has been demonstrated in a statistically significant sample of s-wave neutron resonances (J{sup {pi}} = 0{sup +} or 1{sup +}) that the 0{sup +} levels have a characteristic 115Cd yield which is a factor of four higher than the yield at 1{sup +} levels. The fission widths of the J = 0 levels are larger than the J = 1 levels by a factor of ten. The populations of the two groups are in reasonable agreement with the expected (2J + 1) distributions. Previous efforts to obtain equally detailed data in 235-U fission and 233-U fission by the 'wheel' technique have not been entirely successful due in large part to the high level densities in the epithermal excitation functions of these nuclides and the consequent difficulty in characterizing fission yields in a sufficiently large and well-resolved sample of levels. In a recent 'wheel' experiment (late summer, 1969) vith a 235-U target the energy resolution was sufficiently improved in the region 20 eV-60 eV to allow characterization of a sample of 38 reasonably well-resolved levels by their relative symmetry of fission. (author)

  19. Development of neutron induced prompt γ-ray spectroscopy system using 252Cf

    International Nuclear Information System (INIS)

    Park, Yong-Joon; Song, Byung-Chul; Jee, Kwang-Yong

    2003-01-01

    For the design and set-up of neutron induced prompt γ-ray spectroscopy system using 252 Cf neutron source, the effects of shielding and moderator materials have been examined. The 252 Cf source being used for TLD badge calibration in Korea Atomic Energy Research Institute was utilized for this preliminary experiment. The γ-ray background and prompt γ-ray spectrum of the sample containing Cl were measured using HPGe (GMX 69% relative efficiency) located at the inside of the system connected to notebook PC at the outside of the system (about 20 meter distance). The background activities of neutron and γ-rays were measured with neutron survey meter as well as γ-ray survey meters, respectively and the system was designed to minimize the activities. Prompt γ-ray spectrum was measured using γ-γ coincident system for reduce the background and the continuum spectrum. The optimum system was designed and set up using the experimental data obtained

  20. Neutron-induced gamma-ray spectroscopy: simulations for chemical mapping of planetary surfaces

    International Nuclear Information System (INIS)

    Brueckner, J.; Waenke, H.; Reedy, R.C.

    1986-01-01

    Cosmic rays interact with the surface of a planetary body and produce a cascade of secondary particles, such as neutrons. Neutron-induced scattering and capture reactions play an important role in the production of discrete gamma-ray lines that can be measured by a gamma-ray spectrometer on board of an orbiting spacecraft. These data can be used to determine the concentration of many elements in the surface of a planetary body, which provides clues to its bulk composition and in turn to its origin and evolution. To investigate the gamma rays made by neutron interactions, thin targets were irradiated with neutrons having energies from 14 MeV to 0.025 eV. By means of foil activation technique the ratio of epithermal to thermal neutrons was determined to be similar to that in the Moon. Gamma rays emitted by the targets and the surrounding material were detected by a high-resolution germanium detector in the energy range of 0.1 to 8 MeV. Most of the gamma-ray lines that are expected to be used for planetary gamma-ray spectroscopy were found in the recorded spectra and the principal lines in these spectra are presented. 58 refs., 7 figs., 9 tabs

  1. Radioactive ion beams produced by neutron-induced fission at ISOLDE

    CERN Document Server

    Catherall, R; Gilardoni, S S; Köster, U

    2003-01-01

    The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN, tests have been made on standard ISOLDE actinide targets using fast neutron bunches produced by bombarding thick, high-Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC2/graphite and ThO2 targets with tungsten an...

  2. A study on the behavior of boron in iron-base alloys by neutron induced autoradiography

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin Sung; Rhee, Chang Kyu; Cho, Hae Dong; Han, Chang Hee; Lee, Chang Hee; Jung, Jung Hwan; Kim, Yi Kyung; Lee, Yong Bok

    2001-02-01

    Boron is widely utilized in steel or alloy making to improve certain properties. However, due to its lightness boron is difficult to detect or characterize its behavior even through TEM/EDS or EELS techniques. Although many companies recognize the beneficial effects of boron, the role or mechanism of the boron is not yet clearly understood. Therefore it is required to develop the autoradiography technique to elucidate the boron behavior in alloys. As the only institute operating research reactor in the country, it would be the responsibility of the institute to develop the technique and provide it to the industries. Quantitative analyses of boron in type 316 L stainless steel by neutron induced autoradiography was attempted in this study. Nine experimental reference alloys with different amount of boron were prepared and reliable chemical composition data were obtained. Autoradiographs of reference materials with three different neutron fluences ( 1.9 10{sup 13}, 1.9 10{sup 14} and 1.9 10{sup 15}/cm{sup 2} ) were obtained and a trial calibration curve of boron content vs. track density was acquired.

  3. 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)

  4. SCATPI, a subroutine for calculating πN cross sections and polarizations for incident pion kinetic energies between 90 and 300 MeV

    International Nuclear Information System (INIS)

    Walter, J.B.; Rebka, G.A. Jr.

    1979-03-01

    A subroutine, SCATPI, was written which calculates π + p elastic differential cross sections for incident pion kinetic energies between 90 and 310 MeV for π - p. The calculation is based upon the phase shift analysis of Carter, Bugg, and Carter, and is reliable to about 2% for π + p and 3% for π - p differential cross sections. SCATPI also calculates other scattering parameters for the π+-p systems. The calculations are compared with the measurements used in the phase shift analysis, and with selected recent measurements. The use of SCATPI is described. 14 figures, 4 tables

  5. Experimental studies of keV energy neutron-induced reactions relevant to astrophysics and nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Shima, T.; Kii, T.; Kikuchi, T.; Okazaki, F.; Kobayashi, T.; Baba, T.; Nagai, Y. [Tokyo Inst. of Tech. (Japan). Faculty of Science; Igashira, M.

    1997-03-01

    Nuclear reactions induced by keV energy neutrons provide a plenty of informations for studies of both astrophysics and nuclear physics. In this paper we will show our experimental studies of neutron- induced reactions of light nuclei in the keV energy region by means of a pulsed keV neutron beam and high-sensitivity detectors. Also we will discuss astrophysical and nuclear-physical consequences by using the obtained results. (author)

  6. Calculations of standard-Higgs-boson production cross sections in e+e- collisions by means of a reasonable set of parameters

    International Nuclear Information System (INIS)

    Biyajima, M.; Shirane, K.; Terazawa, O.

    1987-01-01

    We calculate cross sections for production of the standard Higgs boson in e + e - collisions and compare our results with those of several authors. It is found that there are appreciable differences among them which can be attributed to the coupling constants used, α(0) ( = (1/137) and G/sub F/. We also observe that cross sections depend on the magnitudes of the total width of the Z particle. The use of a reasonable set of parameters in calculations is emphasized

  7. Cross section for calculating the helium formation rate in construction materials irradiated by nucleons at energies to 800 MeV

    International Nuclear Information System (INIS)

    Konobeev, A.Yu.; Korovin, Yu.A.

    1992-01-01

    Recently, effects related to the formation of helium in irradiated construction materials have been studied extensively. Data on the nuclear cross sections for producing helium in these materials form the initial information necessary for such investigations. If the spectrum of the incoming particles is known, the value of the helium production cross section makes it possible to calculate the helium generation rate. In recent years, plans and simulating experiments on radiating materials with high-energy particles made it necessary to determine the helium production cross sections in constructionmaterials, which are irradiated by protons and neutrons with energies to 800 MeV. Helium-formation cross sections have been calculated at these energies. However, a correct description of the experimental data for various construction materials does not yet exist. For example, the calculated helium-formation cross sections turned out to overestimate the experimental data, and to underestimate the experimental data. The objective here is to calculate the helium-formation cross sections for various construction materials, which are irradiated by protons and neutrons to energies from 20 to 800 MeV, and to analyze the probable causes of deviations between experimental and earlier calculated cross sections

  8. Covariances for neutron cross sections calculated using a regional model based on local-model fits to experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Guenther, P.T.

    1983-11-01

    We suggest a procedure for estimating uncertainties in neutron cross sections calculated with a nuclear model descriptive of a specific mass region. It applies standard error propagation techniques, using a model-parameter covariance matrix. Generally, available codes do not generate covariance information in conjunction with their fitting algorithms. Therefore, we resort to estimating a relative covariance matrix a posteriori from a statistical examination of the scatter of elemental parameter values about the regional representation. We numerically demonstrate our method by considering an optical-statistical model analysis of a body of total and elastic scattering data for the light fission-fragment mass region. In this example, strong uncertainty correlations emerge and they conspire to reduce estimated errors to some 50% of those obtained from a naive uncorrelated summation in quadrature. 37 references.

  9. Montecarlo calculation of the isomeric cross sections ratio for the reaction 237Np(n,2n)236Np

    International Nuclear Information System (INIS)

    Cleri, F.

    1988-01-01

    A Montecarlo calculation of the isomeric cross section ratio for the (n,2n) reaction on 237 Np has been carried out based on the Hauser-Feshbach formulation. A standard energy-dependent optical model potential was used, with zero deformation parameters and no spin-orbit coupling. Investigation was made about the role of the energy cut-off value, of the higher multipole (E2) transition, of the gamma-ray versus second neutron emission, of the value of the spin cutt-off parameter. The results give the correct qualitative energy dependence of the branching ratio, with the assumption that the 1 - level is the ground state. The spin cut-off value obtained indicates a less pronounced deviation of the nuclear moment of inertia from the rigid-body value, with respect to older evaluations for high-mass nuclei. (author)

  10. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Energy Technology Data Exchange (ETDEWEB)

    Díez, C.J., E-mail: cj.diez@upm.es [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Cabellos, O. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Martínez, J.S. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain)

    2015-01-15

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

  11. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    International Nuclear Information System (INIS)

    Díez, C.J.; Cabellos, O.; Martínez, J.S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties

  12. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Science.gov (United States)

    Díez, C. J.; Cabellos, O.; Martínez, J. S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

  13. MICROX-2: an improved two-region flux spectrum code for the efficient calculation of group cross sections

    International Nuclear Information System (INIS)

    Mathews, D.; Koch, P.

    1979-12-01

    The MICROX-2 code is an improved version of the MICROX code. The improvements allow MICROX-2 to be used for the efficient and rigorous preparation of broad group neutron cross sections for poorly moderated systems such as fast breeder reactors in addition to the well moderated thermal reactors for which MICROX was designed. MICROX-2 is an integral transport theory code which solves the neutron slowing down and thermalization equations on a detailed energy grid for two-region lattice cells. The fluxes in the two regions are coupled by transport corrected collision probabilities. The inner region may include two different types of grains (particles). Neutron leakage effects are treated by performing B 1 slowing down and P 0 plus DB 2 thermalization calculations in each region. Cell averaged diffusion coefficients are prepared with the Benoist cell homogenization prescription

  14. Covariances for neutron cross sections calculated using a regional model based on local-model fits to experimental data

    International Nuclear Information System (INIS)

    Smith, D.L.; Guenther, P.T.

    1983-11-01

    We suggest a procedure for estimating uncertainties in neutron cross sections calculated with a nuclear model descriptive of a specific mass region. It applies standard error propagation techniques, using a model-parameter covariance matrix. Generally, available codes do not generate covariance information in conjunction with their fitting algorithms. Therefore, we resort to estimating a relative covariance matrix a posteriori from a statistical examination of the scatter of elemental parameter values about the regional representation. We numerically demonstrate our method by considering an optical-statistical model analysis of a body of total and elastic scattering data for the light fission-fragment mass region. In this example, strong uncertainty correlations emerge and they conspire to reduce estimated errors to some 50% of those obtained from a naive uncorrelated summation in quadrature. 37 references

  15. Calculation of 3-D free electron laser gain: Comparison with simulation and generalization to elliptical cross section

    International Nuclear Information System (INIS)

    Chin, Yong Ho; Kim, Kwang-Je; Xie, Ming.

    1992-08-01

    In the previous paper, we have derived a dispersion relation for the free electron laser (FEL) gain in the exponential regime taking account the diffraction and electron's betatron oscillation. Here, we compare the growth rates obtained by solving the dispersion relation with those obtained by simulation calculation for the waterbag and the Gaussian models for the electron's transverse phase space distribution. The agreement is found to be good except for the limiting case where the Rayleigh length is much longer than the gain length (1-D limit). We also generalize the analysis to the case where the electron beam cross section is elliptical as is usually the case in storage rings, and derive the first-order dispersion relation

  16. Measurement of the $^{242}$Pu(n,f) reaction cross-section at the CERN n_TOF facility

    CERN Document Server

    AUTHOR|(CDS)2080481; Kokkoris, Michael; Vlachoudis, Vasilis

    The accurate knowledge of relevant nuclear data, such as the neutron-induced fission cross sections of various plutonium isotopes and other minor actinides, is crucial for the design of advanced nuclear systems as well as the development of comprehensive theoretical models of the fission process. The $^{242}$Pu(n,f) cross section was measured at the CERN n_TOF facility taking advantage of the wide energy range and the high instantaneous flux of the neutron beam. In this work, results for the $^{242}$Pu(n,f) measurement are presented along with a detailed description of the experimental setup, Monte-Carlo simulations and the analysis procedure, and a theoretical cross section calculation performed with the EMPIRE code.

  17. Improved algorithms for the calculation of resolved resonance cross sections with applications to the structural Doppler effect in fast reactors

    International Nuclear Information System (INIS)

    Hwang, R.N.; Toppel, B.J.; Henryson, H. II.

    1980-10-01

    Motivated by a need for an economical yet rigorous tool which can address the computation of the structural material Doppler effect, an extremely efficient improved RABANL capability has been developed utilizing the fact that the Doppler broadened line shape functions become essentially identical to the natural line shape functions or Lorentzian limits beyond about 100 Doppler widths from the resonance energy, or when the natural width exceeds about 200 Doppler widths. The computational efficiency has been further enhanced by preprocessing or screening a significant number of selected resonances during library preparation into composition and temperature independent smooth background cross sections. The resonances which are suitable for such pre-processing are those which are either very broad or those which are very weak. The former contribute very little to the Doppler effect and their self-shielding effect can readily be averaged into slowly varying background cross section data, while the latter contribute very little to either the Doppler or to self-shielding effects. To illustrate the accuracy and efficiency of the improved RABANL algorithms and resonance screening techniques, calculations have been performed for two systems, the first with a composition typical of the STF converter region and the second typical of an LMFBR core composition. Excellent agreement has been found for RABANL compared to the reference Monte Carlo solution obtained using the code VIM, and improved results have also been obtained for the narrow resonance approximation in the ultra-fine-group option of MC 2 -2

  18. A comparison of three time-dependent wave packet methods for calculating electron--atom elastic scattering cross sections

    International Nuclear Information System (INIS)

    Judson, R.S.; McGarrah, D.B.; Sharafeddin, O.A.; Kouri, D.J.; Hoffman, D.K.

    1991-01-01

    We compare three time-dependent wave packet methods for performing elastic scattering calculations from screened Coulomb potentials. The three methods are the time-dependent amplitude density method (TDADM), what we term a Cayley-transform method (CTM), and the Chebyshev propagation method of Tal-Ezer and Kosloff. Both the TDADM and the CTM are based on a time-dependent integral equation for the wave function. In the first, we propagate the time-dependent amplitude density, |ζ(t)right-angle=U|ψ(t)right-angle, where U is the interaction potential and |ψ(t)right-angle is the usual time-dependent wave function. In the other two, the wave function is propagated. As a numerical example, we calculate phase shifts and cross sections using a screened Coulomb, Yukawa type potential over the range 200--1000 eV. One of the major advantages of time-dependent methods such as these is that we get scattering information over this entire range of energies from one propagation. We find that in most cases, all three methods yield comparable accuracy and are about equally efficient computationally. However for l=0, where the Coulomb well is not screened by the centrifugal potential, the TDADM requires smaller grid spacings to maintain accuracy

  19. High energy neutron cross-sections and kerma values of biomedical interest calculated with a nuclear model applicable to light nuclei

    International Nuclear Information System (INIS)

    Brenner, D.J.

    1984-01-01

    A model has been developed for calculating fast neutron cross sections (E > 14 MeV) for light nuclei of biomedical interest. The model explicitly includes experimental nuclear structure information. Some calculations for 12 C, 14 N, and 16 O are presented

  20. Comparison Of Quantum Mechanical And Classical Trajectory Calculations Of Cross Sections For Ion-Atom Impact Ionization of Negative - And Positive -Ions For Heavy Ion Fusion Applications

    International Nuclear Information System (INIS)

    Kaganovich, Igor D.; Startsev, Edward A.; Davidson, Ronald C.

    2003-01-01

    Stripping cross sections in nitrogen have been calculated using the classical trajectory approximation and the Born approximation of quantum mechanics for the outer shell electrons of 3.2GeV I - and Cs + ions. A large difference in cross section, up to a factor of six, calculated in quantum mechanics and classical mechanics, has been obtained. Because at such high velocities the Born approximation is well validated, the classical trajectory approach fails to correctly predict the stripping cross sections at high energies for electron orbitals with low ionization potential