Validation of evaluated neutron standard cross sections
International Nuclear Information System (INIS)
Badikov, S.; Golashvili, T.
2008-01-01
Some steps of the validation and verification of the new version of the evaluated neutron standard cross sections were carried out. In particular: -) the evaluated covariance data was checked for physical consistency, -) energy-dependent evaluated cross-sections were tested in most important neutron benchmark field - 252 Cf spontaneous fission neutron field, -) a procedure of folding differential standard neutron data in group representation for preparation of specialized libraries of the neutron standards was verified. The results of the validation and verification of the neutron standards can be summarized as follows: a) the covariance data of the evaluated neutron standards is physically consistent since all the covariance matrices of the evaluated cross sections are positive definite, b) the 252 Cf spectrum averaged standard cross-sections are in agreement with the evaluated integral data (except for 197 Au(n,γ) reaction), c) a procedure of folding differential standard neutron data in group representation was tested, as a result a specialized library of neutron standards in the ABBN 28-group structure was prepared for use in reactor applications. (authors)
Covariance Evaluation Methodology for Neutron Cross Sections
Energy Technology Data Exchange (ETDEWEB)
Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.
2008-09-01
We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.
Evaluation methods for neutron cross section standards
International Nuclear Information System (INIS)
Bhat, M.R.
1980-01-01
Methods used to evaluate the neutron cross section standards are reviewed and their relative merits, assessed. These include phase-shift analysis, R-matrix fit, and a number of other methods by Poenitz, Bhat, Kon'shin and the Bayesian or generalized least-squares procedures. The problems involved in adopting these methods for future cross section standards evaluations are considered, and the prospects for their use, discussed. 115 references, 5 figures, 3 tables
Mechanized evaluation of neutron cross-sections
International Nuclear Information System (INIS)
Horsley, A.; Parker, J.B.
1967-01-01
The evaluation work to provide accurate and consistent neutron cross-section data for multigroup neutronics calculations is not fully exploiting the available theoretical and experimental results; this has been so particularly since the introduction of on-line data handling techniques enabled experimenters to turn out vast quantities of numbers. This situation can be radically improved only by mechanizing the evaluation processes. Systems such as the SC1SRS tape will not only largely overcome the task of collecting data but will provide speedy access to it; by using computers and graph-plotting machines to tabulate and display this data, the labour of evaluation can be very greatly reduced. With some types of cross-section there is hope that by using modern curve-fitting techniques the actual evaluation and statistical accounting of the data can be performed automatically. Some areas where automatic evaluation would seem likely to succeed are specified and a discussion of the mathematical difficulties incurred, such as the elimination of anomalous data, is given. Particularly promising is the use of splines in the mechanized evaluation of data. Splines are the mathematical analogues of the draughtsman's spline used in drawing smooth curves. Their principal properties are the excellent approximations they give to the derivatives of a function; in contrast to conventional polynomial fitting, this feature ensures good interpolation and, when required, stable extrapolation. Various methods of using splines in data graduation and the problem of marrying these methods to standard statistical procedures are examined. The results of work done at AWRE with cubic splines on the mechanized evaluation of neutron scattering total cross-section and angular distribution data are presented. (author)
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)
Evaluation of the 238U neutron total cross section
International Nuclear Information System (INIS)
Smith, A.; Poenitz, W.P.; Howerton, R.J.
1982-12-01
Experimental energy-averaged neutron total cross sections of 238 U were evaluated from 0.044 to 20.0 MeV using regorous numerical methods. The evaluated results are presented together with the associated uncertainties and correlation matrix. They indicate that this energy-averaged neutron total cross section is known to better than 1% over wide energy regions. There are somwewhat larger uncertainties at low energies (e.g., less than or equal to 0.2 MeV), near 8 MeV and above 15 MeV. The present evaluation is compard with values given in ENDF/B-V
Measured and evaluated neutron cross sections of elemental bismuth
International Nuclear Information System (INIS)
Smith, A.; Guenther, P.; Smith, D.; Whalen, J.; Howerton, R.
1980-04-01
Neutron total cross sections of elemental bismuth are measured with broad resolution from 1.2 to 4.5 MeV to accuracies of approx. = 1%. Neutron-differential-elastic-scattering cross sections of bismuth are measured from 1.5 to 4.0 MeV at incident neutron energy intervals of approx.< 0.2 MeV over the scattered-neutron angular range approx. = 20 to 160 deg. Differential neutron cross sections for the excitation of observed states in bismuth at 895 +- 12, 1606 +- 14, 2590 +- 15, 2762 +- 29, 3022 +- 21, and 3144 +- 15 keV are determined at incident neutron energies up to 4.0 MeV. An optical-statistical model is deduced from the measured values. This model, the present experimental results, and information available elsewhere in the literature are used to construct a comprehensive evaluated nuclear data file for elemental bismuth in the ENDF format. The evaluated file is particularly suited to the neutronic needs of the fusion-fission hybrid designer. 87 references, 10 figures, 6 tables
Evaluation of fission product neutron cross sections for JENDL
International Nuclear Information System (INIS)
1984-01-01
The recent activities on the evaluation of fission product (FP) neutron cross sections for JENDL (Japanese Evaluated Nuclear Data Library) are presented briefly. The integral test of JENDL-1 FP cross section file was performed using the CFRMF sample activation data and the STEK sample reactivity data, and the ratio of experiment to calculation was nearly constant for all the samples in the STEK measurement. Therefore, a tentative analysis was performed by applying the correction to the calculated scattering reactivity component. Better agreement with the experiment was obtained after applying this correction in most cases. The evaluation work on the JENDL-2 FP neutron cross sections is now in progress. The improvement of the data evaluation is presented in an itemized form. The JENDL-2 FP file will contain the evaluated data for 100 nuclides from Kr to Tb. The improvement and the future scope of the integral test for JENDL-2 FP data are summarized. (Asami, T.)
Measured and evaluated fast neutron cross sections of elemental nickel
International Nuclear Information System (INIS)
Guenther, P.; Smith, A.; Smith, D.; Whalen, J.; Howerton, R.
1975-07-01
Fast neutron total and scattering cross sections of elemental nickel are measured. Differential elastic scattering cross sections are determined from incident energies of 0.3 to 4.0 MeV. The cross sections for the inelastic neutron excitation of states at: 1.156 +- 0.015, 1.324 +- 0.015, 1.443 +- 0.015, 2.136 +- 0.013, 2.255 +- 0.030, 2.449 +- 0.030, 2.614 +- 0.020 and 2.791 +- 0.025 MeV are measured to incident neutron energies of 4.0 MeV. The total neutron cross sections are determined from 0.25 to 5.0 MeV. The experimental results are discussed in the context of optical and statistical models. It is shown that resonance width-fluctuation and correlation effects are significant. The present experimental and theoretical results, together with previously reported values, are used to construct a comprehensive evaluated elemental data file in the ENDF format. Some comparisons are made with previously reported evaluated files. In addition, some selected reactions which are widely used in dosimetry and other applications are presented as supplemental evaluated isotopic-data files. The numerical quantities are presented in tabular form. (3 tables, 29 figures)
Neutron cross section standards evaluations for ENDF/B-VI
International Nuclear Information System (INIS)
Carlson, A.D.; Poenitz, W.P.; Hale, G.M.; Peelle, R.W.
1985-01-01
The neutron cross section standards are now being evaluated as the initial phase in the development of the new ENDF/B-VI file. These standards evaluations are following a somewhat different process compared with that used for earlier versions of ENDF. The primary effort is concentrated on a simultaneous evaluation using a generalized least squares program, R-matrix evaluations, and a procedure for combining the results of these evaluations. The ENDF/B-VI standards evaluation procedure is outlined, and preliminary simultaneous evaluation and R-matrix results are presented. 16 refs., 7 figs
Evaluation of the neutron and gamma-ray production cross-sections for 55Mn
International Nuclear Information System (INIS)
Takahashi, H.
1974-11-01
The evaluation of neutron and gamma production cross sections for manganese-55 from 1.0 (10) -5 eV to 20.0 MeV for ENDF/ B-IV is summarized. Included are resonance parameters, neutron cross sections, angular and energy distribution of secondary neutrons, gamma multiplicities and transition probability array, gamma angular and energy distributions, nuclear model calculations, uncertainty estimates of cross sections, and evaluated cross sections. (U.S.)
International Nuclear Information System (INIS)
Vonach, H.
1980-06-01
The cross-sections for the four important neutron dosimetry reactions 19 F(n,2n) 18 F, 31 P(n,p) 31 Si, 93 Nb(n,n')sup(93m)Nb and 103 Rh(n,n')sup(103m)Rh were evaluated in the neutron energy range from threshold to 20 MeV. For the 19 F(n,2n) reaction the evaluation could be based entirely on experimental data; for the reactions 31 P(n,p) 31 Si and 103 Rh(n,n')sup(103m)Rh large gaps in the experimental excitation functions and large discrepancies between the existing data made it necessary to supplement the experimental data by cross-section calculations and to give about equal weight to the experimental and calculated cross-sections. For the 93 Nb(n,n')sup(93m)Nb reaction the evaluation had to be based entirely on the theoretically calculated cross-sections. The cross-section calculations were performed using the statistical model of nuclear reactions allowing for precompound processes in the first reaction step and errors of the calculated cross-sections were estimated from their sensitivity to the various input parameters. Cross-section values were evaluated for energy groups between 0.1 MeV and 1 MeV wide, the width depending on both the slope of the excitation functions and the density of the available data. For each evaluated cross-section also an uncertainty (on a 1 sigma confidence level) was derived taking into account the errors given by the experimentalists, the general consistency of the experimental data and the estimated errors of the theoretically calculated cross-sections. In addition relative correlation matrices were derived for each evaluated excitation function describing the correlations between the uncertainties of the cross-sections at different energies. The correlations between the cross-section uncertainties for different reactions were found to be negligible. The results of this evaluation as well as those of Ref. 1 will be combined with the ENDF/B-V dosimetry file into an international neutron dosimetry file by the nuclear data section of
Consistent evaluation of neutron cross sections for the 242-244Cm isotopes
International Nuclear Information System (INIS)
Ignatyuk, A.V.; Maslov, V.M.
1989-01-01
The knowledge of neutron cross-sections for Curium isotopes is necessary for solving the problems of the external fuel cycle. Experimental information on the cross-sections is very meager and does not satisfy requirements and existing evaluations in different libraries differ substantially for fission and (n,2n) reaction cross-sections. This situation requires a critical review of the entire set of evaluations of the neutron cross-sections for Curium. 17 refs, 3 figs
Evaluation of Cm-247 neutron cross sections in the resonance region
International Nuclear Information System (INIS)
Martinelli, T.; Menapace, E.; Motta, M.; Vaccari, M.
1980-01-01
The neutron cross sections of Cm-247 are evaluated in the resonance (resolved and unresolved) region up to 10 keV. Average resonance parameters (i.e. spacing D, fission and radiative widths, neutron strength functions) are determined for unresolved region calculations. Moreover for a better comparison with the experimental data, fission cross section is calculated up to 10 MeV. In addition, the average number of neutrons emitted per fission as a function of energy is estimated
Evaluation of neutron cross sections for 244Cm, 246Cm, and 248Cm
International Nuclear Information System (INIS)
Benjamin, R.W.; McCrosson, F.J.; Gettys, W.E.
1977-01-01
An evaluation of neutron cross sections for 244 246 248 Cm using the ENDF/B format is presented. Primary data input included differential measurements, integral measurements, nuclear model calculations, and reactor production experience
The evaluation of neutron total cross section for natural iron and aluminium
International Nuclear Information System (INIS)
Liu Shirui; Wang Chunhao; Zhao Defang
1990-05-01
The experimental data of total cross section were collected and evaluated for natural iron in the energy region from 1 keV to 20 MeV and for natural aluminium from 4.07 keV to 20 MeV. The evaluated data were recommended in the regions for them. The minimum values of Fe total cross section in the keV region were specially recommended. The resonance structures were briefly discussed for both Fe and Al. To make the evaluation better, all experimental measurements of neutron total cross section relative to Fe and Al were studied. Considering the resonance feature of medium weight nuclides, two criteria for selecting total cross section were presented: 1) the correlation between the precission of total cross section and neutron source; 2) the correlation between the accuracy of total cross section and the resolving power of the neutron spectrometer
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
Evaluation of the neutron cross sections for Pu-240
International Nuclear Information System (INIS)
Weston, L.W.; Arthur, E.D.
1987-04-01
The present evaluation is proposed to supersede the ENDF/B-V, Revision 2 file for 240 Pu. In this work, resonance parameters, cross sections, energy distributions, and angular distributions have been modified. These changes are outlined in detail and appropriate references included. 37 refs., 21 figs., 2 tabs
Contribution to uncertainties evaluation for fast reactors neutronic cross sections
International Nuclear Information System (INIS)
Privas, Edwin
2015-01-01
The thesis has been motivated by a wish to increase the uncertainty knowledge on nuclear data, for safety criteria. It aims the cross sections required by core calculation for sodium fast reactors (SFR), and new tools to evaluate its.The main objective of this work is to provide new tools in order to create coherent evaluated files, with reliable and mastered uncertainties. To answer those problematic, several methods have been implemented within the CONRAD code, which is developed at CEA of Cadarache. After a summary of all the elements required to understand the evaluation world, stochastic methods are presented in order to solve the Bayesian inference. They give the evaluator more information about probability density and they also can be used as validation tools. The algorithms have been successfully tested, despite long calculation time. Then, microscopic constraints have been implemented in CONRAD. They are defined as new information that should be taken into account during the evaluation process. An algorithm has been developed in order to solve, for example, continuity issues between two energy domains, with the Lagrange multiplier formalism. Another method is given by using a marginalization procedure, in order to either complete an existing evaluation with new covariance or add systematic uncertainty on an experiment described by two theories. The algorithms are well performed along examples, such the 238 U total cross section. The last parts focus on the integral data feedback, using methods of integral data assimilation to reduce the uncertainties on cross sections. This work ends with uncertainty reduction on key nuclear reactions, such the capture and fission cross sections of 238 U and 239 Pu, thanks to PROFIL and PROFIL-2 experiments in Phenix and the Jezebel benchmark. (author) [fr
Neutron cross section and covariance data evaluation of experimental data for {sup 27}Al
Energy Technology Data Exchange (ETDEWEB)
Chunjuan, Li; Jianfeng, Liu [Physics Department , Zhengzhou Univ., Zhengzhou (China); Tingjin, Liu [China Nuclear Data Center, China Inst. of Atomic Energy, Beijing (China)
2006-07-15
The evaluation of neutron cross section and covariance data for {sup 27}Al in the energy range from 210 keV to 20 MeV was carried out on the basis of the experimental data mainly taken from EXFOR library. After the experimental data and their errors were analyzed, selected and corrected, SPCC code was used to fit the data and merge the covariance matrix. The evaluated neutron cross section data and covariance matrix for {sup 27}Al given can be collected for the evaluated library and also can be used as the basis of theoretical calculation concerned. (authors)
Neutron cross section and covariance data evaluation of experimental data for 27Al
International Nuclear Information System (INIS)
Li Chunjuan; Liu Jianfeng; Liu Tingjin
2006-01-01
The evaluation of neutron cross section and covariance data for 27 Al in the energy range from 210 keV to 20 MeV was carried out on the basis of the experimental data mainly taken from EXFOR library. After the experimental data and their errors were analyzed, selected and corrected, SPCC code was used to fit the data and merge the covariance matrix. The evaluated neutron cross section data and covariance matrix for 27 Al given can be collected for the evaluated library and also can be used as the basis of theoretical calculation concerned. (authors)
International Nuclear Information System (INIS)
Belanova, T.S.
1994-12-01
Data on the thermal neutron fission and capture cross-sections as well as their corresponding resonance integrals are reviewed and analysed. The data are classified according to the form of neutron spectra under investigation. The weighted mean values of the cross-sections and resonance integrals for every type of neutron spectra were adopted as evaluated data. (author). 87 refs, 2 tabs
Comparative analysis of the neutron cross-sections of iron from various evaluated data libraries
International Nuclear Information System (INIS)
Bychkov, V.M.; Vozyakov, V.V.; Manokhin, V.N.; Smoll, F.; Resner, P.; Seeliger, D.; Hermsdorf, D.
1983-09-01
The comparative analysis of neutron cross-sections of iron from evaluated nuclear data libraries SOKRATOR, KEDAK, ENDL is done in energy interval from 0.025 eV to 20 MeV. Some of iron cross-sections from SOKRATOR library are revised and new data, which are obtained by using new experimental data and more comprehensive theoretical methods, are recommended. As a result the new version of the iron neutron cross-section file (BNF-2012) is produced for SOKRATOR library. (author)
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
New evaluated neutron cross section libraries for the GEANT4 code
International Nuclear Information System (INIS)
Mendoza, E.; Cano-Ott, D.; Guerrero, C.; Capote, R.
2012-04-01
The so-called High Precision neutron physics model implemented in the GEANT4 simulation package allows simulating the transport of neutrons with energies up to 20 MeV. It relies on the G4NDL cross section libraries, prepared by the GEANT4 collaboration from evaluated cross section files and distributed freely together with the code. Even though the performance of the G4NDL library has been improved over the time, users running complex simulations which involve the transport of neutrons do need more flexibility, in particular when assessing the uncertainties in the simulation results due to the neutron (and hence the nuclear) data library used. For this reason, a software tool has been developed for transforming any evaluated neutron cross section library in the ENDF-6 format into the G4NDL format. Furthermore, eight different releases of ENDF-B, JEFF, JENDL, CENDL and BROND national libraries have been translated into the G4NDL format and are distributed by the IAEA nuclear data service at www-nds.iaea.org/geant4. In this way, GEANT4 users have access to the complete list of standard evaluated neutron data libraries when performing Monte Carlo simulations with GEANT4. Consistency checks and a first validation of the libraries have been made following the methods described in this report. (author)
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
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
NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION
International Nuclear Information System (INIS)
OH, S.Y.; CHANG, J.; MUGHABGHAB, S.
2000-01-01
Neutron cross section evaluations of the fission-product isotopes, 95 Mo, 99 Tc, 101 Ru, 103 Rh, 105 Pd, 109 Ag, 131 Xe, 133 Cs, 141 Pr, 141 Nd, 147 Sm, 149 Sm, 150 Sm, 151 Sm, 152 Sm, 153 Eu, 155 Gd, and 157 Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of 155 Gd and 157 Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations
NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION
Energy Technology Data Exchange (ETDEWEB)
OH,S.Y.; CHANG,J.; MUGHABGHAB,S.
2000-05-11
Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.
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
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
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.
A unified Monte Carlo approach to fast neutron cross section data evaluation.
Energy Technology Data Exchange (ETDEWEB)
Smith, D.; Nuclear Engineering Division
2008-03-03
A unified Monte Carlo (UMC) approach to fast neutron cross section data evaluation that incorporates both model-calculated and experimental information is described. The method is based on applications of Bayes Theorem and the Principle of Maximum Entropy as well as on fundamental definitions from probability theory. This report describes the formalism, discusses various practical considerations, and examines a few numerical examples in some detail.
International Nuclear Information System (INIS)
Savin, M.V.; Nefedov, Yu.Ya; Livke, A.V.; Zvenigorodskij, A.G.
2001-01-01
Experimental data on the total gamma-ray production cross-sections for inelastic interaction of fast neutrons with iron nuclei were analysed. The total gamma-ray production cross-sections, grouped according to E γ , were evaluated in the neutron energy range 0.5-19 MeV. The statistical spline approximation method was used to evaluate the experimental data. Evaluated data stored in the ENDF, JENDL, BROND, and other libraries on gamma-ray production spectra and cross-sections for inelastic interaction of fast neutrons with iron nuclei, were analysed. (author)
EVALUATION OF NEUTRON CROSS SECTIONS FOR A COMPLETE SET OF Nd ISOTOPES.
Energy Technology Data Exchange (ETDEWEB)
KIM,H.; HERMAN, M.; MUGHABGHAB, S.F.; OBLOZINSKY, P.; ROCHMAN, D.; LEE. Y.-O.
2007-10-29
Neutron cross sections for a complete set of Nd isotopes, {sup 142,143,144,145,146,147,148,150}Nd, were evaluated in the incident energy range from 10{sup -5} eV to 20 MeV. In the low energy region, including thermal and resolved resonances, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. In the unresolved resonance region we performed additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data. In the fast neutron region, we used the nuclear reaction model code EMPIRE-2.19 validated against the experimental data. The results are compared to the existing nuclear data libraries, including ENDF/B-VI.8, JENDL-3.3 and JEFF-3.1, and to the available experimental data. The new evaluations are suitable for neutron transport calculations and they were adopted by the new evaluated nuclear data file of the United States, ENDF/B-VII.0, released in December 2006.
Evaluation of neutron cross sections for a complete set of Dy isotopes
International Nuclear Information System (INIS)
Kim, Hyeong Il; Herman, M.; Mughabghab, S.F.; Oblozinsky, P.; Lee, Young-Ouk
2008-01-01
Neutron cross sections for a complete set of Dy isotopes, 156,158,160,161,162,163,164 Dy, were evaluated in the incident energy range from 10 -5 eV to 20 MeV. In the low energy region, including thermal and resolved resonances, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. In the unresolved resonance region we performed additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data. In the fast neutron region, we used the nuclear reaction model code EMPIRE-2.19 with the model parameters adjusted to the experimental data. The results are compared with the available experimental data and with the existing nuclear data libraries, including ENDF/B-VI.8 and JEFF-3.1. The new evaluations are suitable for neutron transport calculations and they were adopted by the new US evaluated nuclear data library, ENDF/B-VII.0, released in December 2006
ZZ DOSCROS, Neutron Cross-Section Library for Spectra Unfolding and Integral Parameter Evaluation
International Nuclear Information System (INIS)
Zijp, Willem L.; Nolthenius, Henk J.; Rieffe, Henk Ch.
1987-01-01
1 - Description of problem or function: Format: SAND-II; Number of groups: 640 fine group cross section values; Nuclides: Li, B, F, Na, Mg, Al, S, Sc, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Br, Nb, Mo, Rh, Pd, Ag, In, Sb, I, Cs, La, Eu, Sm, Dy, Lu, Ta, W, Re, Au, Th, U, Np, Pu. Origin: ENDF/B-V mainly, ENDF/B-IV, INDL/V. This library forms in combination with the DAMSIG81 library a convenient source of evaluated energy dependent cross section sets which may be used in the determination of neutron spectra by means of adjustment (or unfolding) procedures or which can be used for the determination of integral parameters (such as damage-to-activation ratio) useful in characterising the neutron spectra. The energy dependent fine group cross section data are presented in a 640 group structure of the SAND-II type. This group structure has 45 energy groups per energy decade below 1 MeV and a group width of 100 KeV above 1 MeV. The total energy span of this group structure is from 10 -10 MeV to 20 MeV. The library has the SAND-II format, which implies that a special part of the library has to contain cover cross section data sets. These cross section data sets are required in the SAND-II program for taking into account the influence of special detector surroundings which may be used during an irradiation. 2 - Method of solution: The selection of the reactions from the evaluated nuclear data libraries was determined by various properties of the reactions for neutron metrology. For this reason all the well- known reactions of the ENDF/B-V dosimetry file are included but these data are supplemented with cross section sets for less well known metrology reactions which may become of interest
Evaluation of sodium-23 neutron capture cross section data for the ENDF/B V-III file
International Nuclear Information System (INIS)
Paik, N.C.; Pitterle, T.A.
1975-01-01
The evaluation of neutron cross sections of 23 Na, material number 1156, for the ENDF/B File is described. Cross sections were evaluated between 10 -5 eV and 15 MeV. Experimental data available up to March 1971 were included in the evaluation
The evaluated neutron cross sections and resonance integrals of fission products with Z = 57-62
International Nuclear Information System (INIS)
Fedorova, A.F.; Pisanko, Zh.I.; Novoselov, G.M.
1976-01-01
Neutron cross sections at a neutron velocity of V=2200 m/s, and resonance integrals for fission products with Z=57-71 are estimated. In obtaining the recommended values the results of the neutron cross sections and resonance integrals for elements used as references were normalized in accordance with the latest adjusted values. In the course of estimation, preference was given to the more accurate methods for obtaining the measured values and to the more recent investigations
The evaluated neutron cross sections and resonance integrals of fission products with Z=63-71
International Nuclear Information System (INIS)
Fedorova, A.F.; Pisanko, Zh.I.; Novoselov, G.M.
1976-01-01
Neutron cross sections at a neutron velocity of V=2200 m/s, and the resonance integrals for fission products with Z=63-71 are estimated. In obtaining the recommended values the results were normalized of the neutron cross sections and resonance integrals for elements used as references in accordance with the latest adjusted values. In the course of estimation, preference was given to the more accurate measuring methods and the more recent investigations. Scientific publications up to 1975 have been used
New evaluations of neutron cross sections for 14N and 16O
International Nuclear Information System (INIS)
Hale, G.M.; Young, P.G.; Chadwick, M.; Chen, Z.P.
1991-01-01
New evaluations of the neutron cross sections for 14 N and 16 O have been made for ENDF/B=VI. The evaluations are based at low energies on R-matrix analyses of reactions in the 15 N and 17 O systems, and at higher energies on GNASH calculations and experimental data evaluations, including covariance analyses. The 15 N system R-matrix analysis includes data from reactions among the channels n+ 14 N, p+ 14 C, and α+ 11 B at energies corresponding to excitations in 15 N below E x =13 MeV. The resonance structure of all cross sections in this energy range is fairly well reproduced. New data indicate a different J-value for the first resonance, however. Sub-threshold S-wave levels required to explain the large n+ 14 N total and elastic cross sections near zero energy give scattering lengths that differ significantly from the previous values. The R-matrix analysis of the 17 O system includes many new measurements of the n+ 16 O total cross section, done primarily at Oak Ridge and at Karlsruhe. The resonance structure of all the cross sections [total, (n,n), (n,α), and (α,α)] is well represented by the fit in the region below E n = 6.5 MeV. The new total cross section information gives different positions for some of the resonances and implies a different normalization for the (n,α) cross sections than that obtained in the ENDF/B-IV analysis. The evaluations at energies above the ranges of the R-matrix analyses incorporate results from a number of experiments performed since the previous ENDF/B evaluations. Especially important are new measurements of the total cross sections and differential elastic, and gamma-ray production cross sections
Evaluation of cross sections for 14 important neutron-dosimetry reactions
International Nuclear Information System (INIS)
Wagner, M.; Vonach, H.; Pavlik, A.; Strohmaier, B.; Tagesen, S.; Martinez-Rico, J.
1990-01-01
The evaluation of the cross sections for the neutron dosimetry reactions 24 Mg(n,p) 24 Na, 27 Al(n,α) 24 Na, 58 Ni(n,2n) 57 Ni, 64 Zn(n,p) 64 Cu, 90 Zr(n,2n) 89 Zr and 93 Nb(n,n') 93m Nb carried out at the IRK about ten years ago were updated taking into account recent experimental results. Besides, new evaluations were performed for four additional dosimetry reactions, namely 52 Cr(n,2n) 51 Cr, 59 Co(n,2n) 58 Co, 93 Nb(n,2n) 92m Nb and 197 Au(n,2n) 196 Au. The deadlines for the retrieval of data for the different reactions lay between March 1989 and February 1990. The evaluations comprise the neutron energy range from threshold to 20 MeV, in a few cases this range is extended up to 21 MeV or 30 MeV. Cross sections and their uncertainties were evaluated in energy groups with widths of 0.1 MeV to 2.0 MeV, and relative correlation matrices of the evaluated cross sections at different energies were derived. The results of the evaluations are compared to the previous ones and to other recent evaluations reported in the literature. The main results of our previous evaluations for the reactiosn 19 F(n,2n) 18 F, 31 P(n,p) 31 Si, 63 Cu(n,2n) 62 Cu and 103 Rh(n,n') 103m Rh which remain unchanged are also given for completeness. The evaluations reported in this work will be included in the new version of the IRDF (International Reactor Dosimetry File) of the IAEA in ENDF/B-VI format. (orig.)
New Standard Evaluated Neutron Cross Section Libraries for the GEANT4 Code and First Verification
Mendoza, Emilio; Koi, Tatsumi; Guerrero, Carlos
2014-01-01
The Monte Carlo simulation of the interaction of neutrons with matter relies on evaluated nuclear data libraries and models. The evaluated libraries are compilations of measured physical parameters (such as cross sections) combined with predictions of nuclear model calculations which have been adjusted to reproduce the experimental data. The results obtained from the simulations depend largely on the accuracy of the underlying nuclear data used, and thus it is important to have access to the nuclear data libraries available, either of general use or compiled for specific applications, and to perform exhaustive validations which cover the wide scope of application of the simulation code. In this paper we describe the work performed in order to extend the capabilities of the GEANT4 toolkit for the simulation of the interaction of neutrons with matter at neutron energies up to 20 MeV and a first verification of the results obtained. Such a work is of relevance for applications as diverse as the simulation of a n...
ZZ ENDL82, Evaluated Charged Particle, Neutron, Photon Cross-Section Library
International Nuclear Information System (INIS)
2001-01-01
Description of program or function: - Format: Described in the manual; - Number of groups: (energies between 100 eV and 100 MeV); - Nuclides: 94 (Z 1 to 99); - Origin: LLNL Evaluated Nuclear Data Library. ENDL82 is a collection of evaluated data for neutron-induced reactions, photon interactions with matter, and charged-particle-induced reactions. It is maintained in a computer-oriented system. All interpolable quantities for neutron-induced reactions are presented so that linear interpolation between successive entries yields values that are consistent with stated experimental errors, where experiments exist, or that adhere to an assumed law, such as 1/v energy dependence, within a small fraction (typically 1%). In the case of an assumed energy-dependence law for cross sections, this is accomplished by creating a large number of (energy, cross section) pairs by computer and subsequently thinning the points to a specified accuracy, using the subroutine THINER. All angular distributions are differential probabilities normalized to an integral of unity over the cosine of the scattering angle. All energy distributions of secondary particles are presented as normalized Legendre polynomial representations. The linear interpolation will construct an acceptable angular distribution at an intermediate energy
Evaluation and Compilation of Neutron Activation Cross Sections for Medical Isotope Production
International Nuclear Information System (INIS)
Binney, Stephen E.
2004-01-01
Calculational assessment and experimental verification of certain neutron cross sections that are related to widely needed new medical isotopes. Experiments were performed at the Oregon State University TRIGA Reactor and the High Flux Irradiation Reactor at Oak Ridge National Laboratory
Evaluation of neutron and gamma-ray-production cross-section data for lead
International Nuclear Information System (INIS)
Fu, C.Y.; Perey, F.G.
1975-01-01
A survey was made of the available information on neutron and gamma-ray-production cross-section measurements of lead. From these and from relevant nuclear-structure information on the Pb isotopes, recommended neutron cross-section data sets for lead covering the neutron energy range from 0.00001 eV to 20.0 MeV have been prepared. The cross sections are derived from experimental results available to February 1972 and from calculations based on optical-model, DWBA, and Hauser--Feshbach theories. Comparisons which show good agreement between theoretical and experimental values are displayed in a number of graphs. Also presented graphically are smoothed total cross sections, Legendre coefficients for angular distributions, and a representative energy distribution of gamma rays from resonance capture. 15 tables, 36 figures, 104 references
Evaluation of Neutron-induced Cross Sections and their Related Covariances with Physical Constraints
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.
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.
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
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.)
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.)
Statistical analysis of correlated experimental data and neutron cross section evaluation
International Nuclear Information System (INIS)
Badikov, S.A.
1998-01-01
The technique for evaluation of neutron cross sections on the basis of statistical analysis of correlated experimental data is presented. The most important stages of evaluation beginning from compilation of correlation matrix for measurement uncertainties till representation of the analysis results in the ENDF-6 format are described in details. Special attention is paid to restrictions (positive uncertainty) on covariation matrix of approximate parameters uncertainties generated within the method of least square fit which is derived from physical reasons. The requirements for source experimental data assuring satisfaction of the restrictions mentioned above are formulated. Correlation matrices of measurement uncertainties in particular should be also positively determined. Variants of modelling the positively determined correlation matrices of measurement uncertainties in a situation when their consequent calculation on the basis of experimental information is impossible are discussed. The technique described is used for creating the new generation of estimates of dosimetric reactions cross sections for the first version of the Russian dosimetric file (including nontrivial covariation information)
R-matrix evaluation of 16O neutron cross sections up to 6.3 MeV
International Nuclear Information System (INIS)
Sayer, R.O.; Leal, L.C.; Larson, N.M.; Spencer, R.R.; Wright, R.O.
2002-01-01
We have evaluated 16 O neutron cross sections in the resolved resonance region with the multilevel Reich-Moore code SAMMY. Resonance parameters were determined by a consistent analysis, including both Doppler and resolution broadening effects. To properly treat the α particle exit channel, an algorithm to calculate charged particle penetrabilities and shifts was incorporated into SAMMY. (author)
R-Matrix Evaluation of 16O neutron cross sections up to 6.3 MeV
International Nuclear Information System (INIS)
Sayer, R.O.; Leal, L.C.; Larson, N.M.; Spencer, R.R.; Wright, R.Q.
2000-01-01
In this paper the authors describe an evaluation of 16 O neutron cross sections in the resolved resonance region with the multilevel Reich-Moore R-matrix formalism. Resonance analyses were performed with the computer code SAMMY [LA98] which utilizes Bayes' method, a generalized least squares technique
International Nuclear Information System (INIS)
Simakov, S.P.; Fischer, U.; Moellendorff, U. von; Schmuck, I.; Konobeev, A.Yu.; Korovin, Yu.A.; Pereslavtsev, P.
2002-01-01
A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ 6,7 Li cross section data. A new code M c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M c DeLicious code was checked against available experimental data and calculation results of M c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M c DeLicious along with newly evaluated d+ 6,7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data
Simakov, S P; Moellendorff, U V; Schmuck, I; Konobeev, A Y; Korovin, Y A; Pereslavtsev, P
2002-01-01
A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ sup 6 sup , sup 7 Li cross section data. A new code M sup c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M sup c DeLicious code was checked against available experimental data and calculation results of M sup c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M sup c DeLicious along with newly evaluated d+ sup 6 sup , sup 7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data.
Energy Technology Data Exchange (ETDEWEB)
Lopez Jimenez, M.J. [CEA/DIF/DPTA/Service de Physique Nucleaire, B.P. 12, F-91680 Bruyeres-le-Chatel (France); Morillon, B. [CEA/DIF/DPTA/Service de Physique Nucleaire, B.P. 12, F-91680 Bruyeres-le-Chatel (France); Romain, P. [CEA/DIF/DPTA/Service de Physique Nucleaire, B.P. 12, F-91680 Bruyeres-le-Chatel (France)]. E-mail: pascal.romain@cea.fr
2005-01-15
A new neutron-induced cross-section evaluation of {sup 238}U from 1 keV up to 200 MeV has been performed using only nuclear reactions models. A new fission penetrability model taking into account a triple humped barrier has been developed. A clear improvement has been observed for K-effective validation tests (up to 30 MeV) with this new evaluation. This improvement is mainly due to a better treatment of the inelastic exit channel.
Evaluation of neutron- and proton-induced cross sections of {sup 27}Al up to 2 GeV
Energy Technology Data Exchange (ETDEWEB)
Lee, Young-Ouk; Chang, Jonghwa [Korea Atomic Energy Research Institute, Yusung, Taejon (Korea, Republic of); Fukahori, Tokio; Chiba, Satoshi
1999-03-01
We have evaluated neutron and proton nuclear data of {sup 27}Al for energies up to 2 GeV. The best set of optical model parameters were obtained above 20 MeV for neutron and above reaction threshold for proton up to 250 MeV with the phenomenological potential forms proposed by Chiba. The transmission coefficients for neutron and proton derived from the optical models are fed into the GNASH code system to calculate angle-energy correlated emission spectra for light ejectiles and gamma rays. For energies above 250 MeV and below 2 GeV, the total, reaction and elastic scattering cross sections were evaluated by an empirical fit and recent systematics. Emitted nucleon and pion were estimated by use of QMD + SDM (Quantum Molecular Dynamics + Statistical Decay Model). (author)
Neutron cross sections for fusion
International Nuclear Information System (INIS)
Haight, R.C.
1979-10-01
First generation fusion reactors will most likely be based on the 3 H(d,n) 4 He reaction, which produces 14-MeV neutrons. In these reactors, both the number of neutrons and the average neutron energy will be significantly higher than for fission reactors of the same power. Accurate neutron cross section data are therefore of great importance. They are needed in present conceptual designs to calculate neutron transport, energy deposition, nuclear transmutation including tritium breeding and activation, and radiation damage. They are also needed for the interpretation of radiation damage experiments, some of which use neutrons up to 40 MeV. In addition, certain diagnostic measurements of plasma experiments require nuclear cross sections. The quality of currently available data for these applications will be reviewed and current experimental programs will be outlined. The utility of nuclear models to provide these data also will be discussed. 65 references
International Nuclear Information System (INIS)
Pashchenko, A.B.; Wienke, H.
1998-01-01
This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron-nucleus interaction cross sections, photon production cross sections and photon-atom interaction cross sections for fusion applications. It is part of the evaluated nuclear data library for fusion applications FENDL-2. The data are available cost-free from the Nuclear Data Section upon request. The data can also be retrieved by the user via online access through international computer networks. (author)
Curves and tables of neutron cross sections
International Nuclear Information System (INIS)
Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi
1990-07-01
Neutron cross-section curves from the Japanese Evaluated Nuclear Data Library version 3, JENDL-3, are presented in both graphical and tabular form for users in a wide range of application areas in the nuclear energy field. The contents cover cross sections for all the main reactions induced by neutrons with an energy below 20 MeV including; total, elastic scattering, capture, and fission, (n,n'), (n,2n), (n,3n), (n,α), (n,p) reactions. The 2200 m/s cross-section values, resonance integrals, and Maxwellian- and fission-spectrum averaged cross sections are also tabulated. (author)
Energy Technology Data Exchange (ETDEWEB)
Pronayaev, V G [Institute of Physics and Power Engineering, Obninsk (Russian Federation)
1997-06-01
Evaluations were made of neutron inelastic scattering cross-sections with excitation of discrete levels of the residual nucleus, the (n,2n) reaction, secondary neutron emission spectra and secondary photons for reactions which contribute substantially to production of photons for {sup 127}I with the use of the theoretical model for neutrons with an initial energy of 60 keV-20 MeV. (author). 10 refs, 3 figs.
Evaluated cross-section libraries and kerma factors for neutrons up to 100 MeV on 12C
International Nuclear Information System (INIS)
Chadwick, M.B.; Blann, M.; Cox, L.; Young, P.G.; Meigooni, A.
1995-01-01
A program is being carried out at Lawrence Livermore National Laboratory to develop high-energy evaluated nuclear data libraries for use in Monte Carlo simulations of cancer radiation therapy. In this report we describe evaluated cross sections and kerma factors for neutrons with incident energies up to 100 MeV on 12 C. The aim of this effort is to incorporate advanced nuclear physics modeling methods, with new experimental measurements, to generate cross section libraries needed for an accurate simulation of dose deposition in fast neutron therapy. The evaluated libraries are based mainly on nuclear model calculations, benchmarked to experimental measurements where they exist. We use the GNASH code system, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms. The libraries tabulate elastic and nonelastic cross sections, angle-energy correlated production spectra for light ejectiles with A≤and kinetic energies given to light ejectiles and heavy recoil fragments. The major steps involved in this effort are: (1) development and validation of nuclear models for incident energies up to 100 MeV; (2) collation of experimental measurements, including new results from Louvain-la-Nueve and Los Alamos; (3) extension of the Livermore ENDL formats for representing high-energy data; (4) calculation and evaluation of nuclear data; and (5) validation of the libraries. We describe the evaluations in detail, with particular emphasis on our new high-energy modeling developments. Our evaluations agree well with experimental measurements of integrated and differential cross sections. We compare our results with the recent ENDF/B-VI evaluation which extends up to 32 MeV
Neutron capture cross section of ^243Am
Jandel, M.
2009-10-01
The Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos National Laboratory (LANL) was used for neutron capture cross section measurement on ^243Am. The high granularity of DANCE (160 BaF2 detectors in a 4π geometry) enables the efficient detection of prompt gamma-rays following neutron capture. DANCE is located on the 20.26 m neutron flight path 14 (FP14) at the Manuel Lujan Jr. Neutron Scattering Center at the Los Alamos Neutron Science Center (LANSCE). The methods and techniques established in [1] were used for the determination of the ^243Am neutron capture cross section. The cross sections were obtained in the range of neutron energies from 0.02 eV to 400 keV. The resonance region was analyzed using SAMMY7 and resonance parameters were extracted. The results will be compared to existing evaluations and calculations. Work was performed under the auspices of the U.S. Department of Energy at Los Alamos National Laboratory by the Los Alamos National Security, LLC under Contract No. DE-AC52-06NA25396 and at Lawrence Livermore National Laboratory by the Lawrence Livermore National Security, LLC under Contract No. DE-AC52-07NA27344. [4pt] [1] M. Jandel et al., Phys. Rev. C78, 034609 (2008)
Evaluation of neutron cross-sections for 242Cm to obtain a complete file
International Nuclear Information System (INIS)
Bakhanovich, L.A.; Klepetskij, A.B.; Maslov, V.M.; Porodzinskij, Yu.V.; Sukhovitskij, E.Sh.
1994-01-01
Experimental fission, capture, inelastic scattering, (n2n), (n3n) and other cross-sections are scarce or unavailable. As a consequence, theoretical models and parameters systematics have been used extensively in the calculation of these data. Data obtained in this work are compared with previous evaluations. Severe discrepancies were found. (author). 10 refs, 2 figs, 2 tabs
Evaluation of tritium production cross-section for neutron-interaction with 7Li
International Nuclear Information System (INIS)
Yu Baosheng; Cai Dunjiu
1987-01-01
The 7 Li(n, n't) α reaction cross-section has been evaluated and recommended. These experimental data were selected up to the end of 1986, in which main microscopic nuclear data and benchmark measurements were included. These data are retrieved from EXFOR master files of International Atomic Energy Agency, and new information is added in which IAE (the Chinese Institute of Atomic Energy) experimental results is considered
International Nuclear Information System (INIS)
Pashchenko, A.B.; Wienke, H.; Ganesan, S.; McLaughlin, P.K.
1996-01-01
This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron nuclear interaction cross-sections and photon production cross-sections and photon-atom interaction cross-sections for fusion applications. It is part of FENDL, the evaluated nuclear data library for fusion applications. The nuclear data are available cost-free for distribution to interested scientists upon request. The data can also be retrieved by the user via online access through international computer networks. (author). 11 refs, 1 tab
International Nuclear Information System (INIS)
Ahmad, Siraj-ul-Islam; Ahmad, Nasir; Aslam
2004-01-01
Several new WIMSD libraries based on recent cross-section evaluations such as IAEA, ENDFB-VI, JENDL, and JEF have been made available by IAEA. These libraries were used for the computation of multiplication factor and energy spectrum for Pakistan Research Reactor-1 (PARR-1). Methodology was validated for benchmark problems made available by IAEA and comparison with reference results. The value of effective multiplication factors for all newly released libraries are 1.8-3.2% less than that of 1981 WIMSD library. The effect of various cross-section libraries on neutron energy spectrum was also studied. Differences of about -10% to 12.5% were found in thermal flux using the newly released libraries as compared with that obtained using 1981 WIMSD library. From the analysis, it was found that the main source of the difference is the cross-sections of hydrogen bound in water. When these cross-sections of hydrogen (bound in water) from new libraries were used along with all other data in 1981 WIMSD library, the k eff obtained in this way has a difference of only 0.02-0.8% with that obtained from new libraries, while the flux spectrum agreed within 1% below 1 MeV with new libraries
Evaluated 182,183,184,186W Neutron Cross Sections and Covariances in the Resolved Resonance Region
International Nuclear Information System (INIS)
Pigni, Marco T; Leal, Luiz C
2015-01-01
Oak Ridge National Laboratory (ORNL) has recently completed the resonance parameter evaluation of four tungsten isotopes, i.e., 182,183,184,186 W, in the neutron energy range of thermal up to several keV. This nuclear data work was performed with support from the US Nuclear Criticality Safety Program (NCSP) in an effort to provide improved tungsten cross section and covariance data for criticality safety analyses. The evaluation methodology uses the Reich-Moore approximation of the R-matrix formalism of the code SAMMY to fit high-resolution measurements performed in 2010 and 2012 at the Geel linear accelerator facility (GELINA), as well as other experimental data sets on natural tungsten available in the EXFOR library. In the analyzed energy range, this work nearly doubles the resolved resonance region (RRR) present in the latest US nuclear data library ENDF/B-VII.1. In view of the interest in tungsten for distinct types of nuclear applications and the relatively homogeneous distribution of the isotopic tungsten - namely, 182 W(26.5%), 183 W(14.31%), 184 W(30.64%), and 186 W(28.43%) - the completion of these four evaluations represents a significant contribution to the improvement of the ENDF library. This paper presents an overview of the evaluated resonance parameters and related covariances for total and capture cross sections on the four tungsten isotopes.
Measurement of actinide neutron cross sections
International Nuclear Information System (INIS)
Firestone, Richard B.; Nitsche, Heino; Leung, Ka-Ngo; Perry, DaleL.; English, Gerald
2003-01-01
The maintenance of strong scientific expertise is critical to the U.S. nuclear attribution community. It is particularly important to train students in actinide chemistry and physics. Neutron cross-section data are vital components to strategies for detecting explosives and fissile materials, and these measurements require expertise in chemical separations, actinide target preparation, nuclear spectroscopy, and analytical chemistry. At the University of California, Berkeley and the Lawrence Berkeley National Laboratory we have trained students in actinide chemistry for many years. LBNL is a leader in nuclear data and has published the Table of Isotopes for over 60 years. Recently, LBNL led an international collaboration to measure thermal neutron capture radiative cross sections and prepared the Evaluated Gamma-ray Activation File (EGAF) in collaboration with the IAEA. This file of 35, 000 prompt and delayed gamma ray cross-sections for all elements from Z=1-92 is essential for the neutron interrogation of nuclear materials. LBNL has also developed new, high flux neutron generators and recently opened a 1010 n/s D+D neutron generator experimental facility
International Nuclear Information System (INIS)
Evain, B.P.; Smith, D.L.; Lucchese, P.
1985-04-01
Available 14-MeV experimental neutron activation cross sections are compiled and evaluated for the following reactions of interest for nuclear-energy technology applications: 27 Al(n,p) 27 Mg, Si(n,X) 28 Al, Ti(n,X) 46 Sc, Ti(n,X) 47 Sc, Ti(n,X) 48 Sc, 51 V(n,p) 51 Ti, 51 V(n,α) 48 Sc, Cr(n,X) 52 V, 55 Mn(n,α) 52 V, 55 Mn(n,2n) 54 Mn, Fe(n,X) 54 Mn, 54 Fe(n,α) 51 Cr, 59 Co(n,p) 59 Fe, 59 Co(n,α) 56 Mn, 59 Co(n,2n) 58 Co, 65 Cu(n,p) 65 Ni, Zn(n,X) 64 Cu, 64 Zn(n,2n) 63 Zn, 113 In(n,n')/sup 113m/In, 115 In(n,n') /sup 115m/In. The compiled values are listed and plotted for reference without adjustments. From these collected results those values for which adequate supplementary information on nuclear constants, standards and experimental errors is provided are selected for use in reaction-by-reaction evaluations. These data are adjusted as needed to account for recent revisions in the nuclear constants and cross section standards. The adjusted results are subsequently transformed to equivalent cross sections at 14.7 MeV for the evaluation process. The evaluations are performed utilizing a least-squares method which considers correlations between the experimental data. 440 refs., 41 figs., 46 tabs
Energy Technology Data Exchange (ETDEWEB)
Evain, B.P.; Smith, D.L.; Lucchese, P.
1985-04-01
Available 14-MeV experimental neutron activation cross sections are compiled and evaluated for the following reactions of interest for nuclear-energy technology applications: /sup 27/Al(n,p)/sup 27/Mg, Si(n,X)/sup 28/Al, Ti(n,X)/sup 46/Sc, Ti(n,X)/sup 47/Sc, Ti(n,X)/sup 48/Sc, /sup 51/V(n,p)/sup 51/Ti, /sup 51/V(n,..cap alpha..)/sup 48/Sc, Cr(n,X)/sup 52/V, /sup 55/Mn(n,..cap alpha..)/sup 52/V, /sup 55/Mn(n,2n)/sup 54/Mn, Fe(n,X)/sup 54/Mn, /sup 54/Fe(n,..cap alpha..)/sup 51/Cr, /sup 59/Co(n,p)/sup 59/Fe, /sup 59/Co(n,..cap alpha..)/sup 56/Mn, /sup 59/Co(n,2n)/sup 58/Co, /sup 65/Cu(n,p)/sup 65/Ni, Zn(n,X)/sup 64/Cu, /sup 64/Zn(n,2n)/sup 63/Zn, /sup 113/In(n,n')/sup 113m/In, /sup 115/In(n,n') /sup 115m/In. The compiled values are listed and plotted for reference without adjustments. From these collected results those values for which adequate supplementary information on nuclear constants, standards and experimental errors is provided are selected for use in reaction-by-reaction evaluations. These data are adjusted as needed to account for recent revisions in the nuclear constants and cross section standards. The adjusted results are subsequently transformed to equivalent cross sections at 14.7 MeV for the evaluation process. The evaluations are performed utilizing a least-squares method which considers correlations between the experimental data. 440 refs., 41 figs., 46 tabs.
Neutron scattering cross sections of uranium-238
International Nuclear Information System (INIS)
Beghian, L.E.; Kegel, G.H.R.; Marcella, T.V.; Barnes, B.K.; Couchell, G.P.; Egan, J.J.; Mittler, A.; Pullen, D.J.; Schier, W.A.
1979-01-01
The University of Lowell high-resolution time-of-flight spectrometer was used to measure angular distributions and 90-deg excitation functions for neutrons scattered from 238 U in the energy range from 0.9 to 3.1 MeV. This study was limited to the elastic and the first two inelastic groups, corresponding to states of 238 U at 45 keV (2 + ) and 148 keV (4 + ). Angular distributions were measured at primary neutron energies of 1.1, 1.9, 2.5, and 3.1 MeV for the same three neutron groups. Whereas the elastic data are in fair agreement with the evaluation in the ENDF/B-IV file, there is substantial disagreement between the inelastic measurements and the evaluated cross sections. 12 figures
ISSUES IN NEUTRON CROSS SECTION COVARIANCES
Energy Technology Data Exchange (ETDEWEB)
Mattoon, C.M.; Oblozinsky,P.
2010-04-30
We review neutron cross section covariances in both the resonance and fast neutron regions with the goal to identify existing issues in evaluation methods and their impact on covariances. We also outline ideas for suitable covariance quality assurance procedures.We show that the topic of covariance data remains controversial, the evaluation methodologies are not fully established and covariances produced by different approaches have unacceptable spread. The main controversy is in very low uncertainties generated by rigorous evaluation methods and much larger uncertainties based on simple estimates from experimental data. Since the evaluators tend to trust the former, while the users tend to trust the latter, this controversy has considerable practical implications. Dedicated effort is needed to arrive at covariance evaluation methods that would resolve this issue and produce results accepted internationally both by evaluators and users.
Evaluation of neutron monitor cross sections for 59Co(n,x)56,57,58Co, 52,54,56Mn, 59Fe reactions
International Nuclear Information System (INIS)
Yu Baosheng; Shen Qingbiao; Cai Dunjiu
1996-01-01
The neutron monitor cross sections for 59 Co(n,x) 56,57,58 Co, 52,54,56 Mn, 59 Fe reactions were evaluated based on recent experimental data and theoretical calculations from threshold energy to 100 MeV. (8 figs.)
Neutron capture cross section standards for BNL 325, Fourth Edition
International Nuclear Information System (INIS)
Holden, N.E.
1981-01-01
This report evaluates the experimental data and recommends values for the thermal neutron cross sections and resonance integrals for the neutron capture reactions: 55 Mn(n,γ), 59 Co(n,γ) and 197 Au(n,γ). The failure of lithium and boron as standards due to the natural variation of the absorption cross sections of these elements is discussed. The Westcott convention, which describes the neutron spectrum as a thermal Maxwellian distribution with an epithermal component, is also discussed
International Nuclear Information System (INIS)
Trofimov, Yu.N.
1991-01-01
The radiative neutron capture cross-section of nuclei has been derived as a function of neutron excess on the basis of the exponential dependence of the cross-section on the reaction energy. It is shown that unknown cross-sections of stable and radioactive nuclei may be evaluated by using the isotonic and isotopic dependence together with available reference cross-section measurements. (author). 4 refs, 3 figs
[Fast neutron cross section measurements
International Nuclear Information System (INIS)
Knoll, G.F.
1992-01-01
From its inception, the Nuclear Data Project at the University of Michigan has concentrated on two major objectives: (1) to carry out carefully controlled nuclear measurements of the highest possible reliability in support of the national nuclear data program, and (2) to provide an educational opportunity for students with interests in experimental nuclear science. The project has undergone a successful transition from a primary dependence on our photoneutron laboratory to one in which our current research is entirely based on a unique pulsed 14 MeV fast neutron facility. The new experimental facility is unique in its ability to provide nanosecond bursts of 14 MeV neutrons under conditions that are ''clean'' and as scatter-free as possible, and is the only one of its type currently in operation in the United States. It has been designed and put into operation primarily by graduate students, and has met or exceeded all of its important initial performance goals. We have reached the point of its routine operation, and most of the data are now in hand that will serve as the basis for the first two doctoral dissertations to be written by participating graduate students. Our initial results on double differential neutron cross sections will be presented at the May 1993 Fusion Reactor Technology Workshop. We are pleased to report that, after investing several years in equipment assembly and optimization, the project has now entered its ''data production'' phase
Evaluation of the neutron cross sections of 235U in the thermal energy region. Final report
International Nuclear Information System (INIS)
Leonard, B.R. Jr.; Kottwitz, D.A.; Thompson, J.K.
1976-02-01
The objective of this work has been to improve the knowledge of the thermal cross sections of the fissile nuclei as a step toward providing a standard data base for the nuclear industry. The methodology uses a form of the Adler-Adler multilevel-fission theory and Breit-Wigner multilevel-scattering theory. It incorporates these theories in a general nonlinear least-squares (LSQ) fitting program SIGLEARNThe analysis methodology in this work was applied to the thermal data on 235 U. A reference data file has been developed which includes most of the known data of interest. The first important result of this work is the assessment of the shape uncertainties of the partial cross sections. The results of our studies lead to the following values and error estimates for 235 U g factors in a thermal (20.44 0 C) energy spectrum: g/sub f/ = 0.97751 (+-0.11%); g/sub γ/ = 0.98230 (+-0.14%). A second important result of this study is the development of a recommended set of 2200 m/s (0.0253 eV) values of the parameters and the probable range of further adjustment which might be made. The analysis also provides the result of a common interpretation of energy-dependent absolute cross-section data of different measurements to yield a consistent set of experimental 0.0253 eV values with rigorous error estimates. It also provides normalization factors for relative fission and capture cross sections on a common basis with rigorous error estimates. The results of these analyses provide a basis for deciding what new measurements would be most beneficial. The most important of these would be improved direct capture data in the thermal region
International Nuclear Information System (INIS)
Smith, D.L.
1992-01-01
Neutron energy distributions (spectra) in differential neutron cross section measurements are represented conventionally by giving the energies (not well defined) and full widths at half maximum of the spectra. In most modern experiments, it is possible for investigators to provide a much more thorough representation of neutron spectra. However, reporting such information would entail more numerical detail than appears practical to document. Therefore, it is suggested that such spectra be represented in publications and files of experimental data by providing the following four moments of each distribution: mean value (energy), standard deviation (resolution), skewness (asymmetry), and kurtosis (sharpness). Implementation of such a standard for data reporting would offer evaluators a much more complete and rational basis for comparing reported values and performing evaluations than previously possible. Some examples are provided to illustrate the concepts
Database studies for the evaluation of the neutron cross section standards
International Nuclear Information System (INIS)
Carlson, A.D.
2003-01-01
Studies are being made of experiments under consideration for the standards database. For each experiment the documentation is checked for corrections that may need to be made and possible errors or missing information. The procedure leads to estimates of the uncertainties and correlations within an experiment and correlations with other experiments. This information is used to form covariance matrices for the measurements so that a full covariance analysis can be performed. Recent work and data which will have an important impact on the evaluation will be the focus of this status report. (author)
Neutron cross sections: Book of curves
International Nuclear Information System (INIS)
McLane, V.; Dunford, C.L.; Rose, P.F.
1988-01-01
Neuton Cross Sections: Book of Curves represents the fourth edition of what was previously known as BNL-325, Neutron Cross Sections, Volume 2, CURVES. Data is presented only for (i.e., intergrated) reaction cross sections (and related fission parameters) as a function of incident-neutron energy for the energy range 0.01 eV to 200 MeV. For the first time, isometric state production cross sections have been included. 11 refs., 4 figs
[Fast neutron cross section measurements
International Nuclear Information System (INIS)
1991-01-01
In the 14 MeV Neutron Laboratory, we have continued the development of a facility that is now the only one of its kind in operation in the United States. We have refined the klystron bunching system described in last year's report to the point that 1.2 nanosecond pulses have been directly measured. We have tested the pulse shape discrimination capability of our primary NE 213 neutron detector. We have converted the RF sweeper section of the beamline to a frequency of 1 MHz to replace the function of the high voltage pulser described in last year's report which proved to be difficult to maintain and unreliable in its operation. We have also overcome several other significant experimental difficulties, including a major problem with a vacuum leak in the main accelerator column. We have completed additional testing to prove the remainder of the generation and measurement systems, but overcoming some of these experimental difficulties has delayed the start of actual data taking. We are now in a position to begin our first series of ring geometry elastic scattering measurements, and these will be underway before the end of the current contract year. As part of our longer term planning, we are continuing the conceptual analysis of several schemes to improve the intensity of our current pulsed beam. These include the provision of a duoplasmatron ion source and/or the provision of preacceleration bunching. Additional details are given later in this report. A series of measurements were carried out at the Tandem Dynamatron Facility involving the irradiation of a series of yttrium foils and the determination of activation cross sections using absolute counting techniques. The experimental work has been completed, and final analysis of the cross section data will be completed within several months
Neutron cross section libraries for analysis of fusion neutronics experiments
International Nuclear Information System (INIS)
Kosako, Kazuaki; Oyama, Yukio; Maekawa, Hiroshi; Nakamura, Tomoo
1988-03-01
We have prepared two computer code systems producing neutron cross section libraries to analyse fusion neutronics experiments. First system produces the neutron cross section library in ANISN format, i.e., the multi-group constants in group independent format. This library can be obtained by using the multi-group constant processing code system MACS-N and the ANISN format cross section compiling code CROKAS. Second system is for the continuous energy cross section library for the MCNP code. This library can be obtained by the nuclear data processing system NJOY which generates pointwise energy cross sections and the cross section compiling code MACROS for the MCNP library. In this report, we describe the production procedures for both types of the cross section libraries, and show six libraries with different conditions in ANISN format and a library for the MCNP code. (author)
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.
Evaluation of n+Mn-55 cross section data up to 150 MeV neutron energy
International Nuclear Information System (INIS)
Pereslavtsev, P.; Fischer, U.
2008-01-01
A new evaluation was performed for the reaction system n + 55 Mn in the neutron energy range from 0.001 to 150 MeV. The evaluation is based on the use of the ECIS96 and GNASH codes. A good description of available experimental data was achieved. A very limited set of existing evaluated data from the available international nuclear data libraries was used for the evaluation. The final general purpose data file was prepared in standard ENDF-6 format and was verified with standard format checking utilities. The data file will undergo benchmark testing and will be finally integrated into the JEFF-3.2 data library. (authors)
Evaluation of the Neutron Data Standards
Carlson, A. D.; Pronyaev, V. G.; Capote, R.; Hale, G. M.; Chen, Z.-P.; Duran, I.; Hambsch, F.-J.; Kunieda, S.; Mannhart, W.; Marcinkevicius, B.; Nelson, R. O.; Neudecker, D.; Noguere, G.; Paris, M.; Simakov, S. P.; Schillebeeckx, P.; Smith, D. L.; Tao, X.; Trkov, A.; Wallner, A.; Wang, W.
2018-02-01
With the need for improving existing nuclear data evaluations, (e.g., ENDF/B-VIII.0 and JEFF-3.3 releases) the first step was to evaluate the standards for use in such a library. This new standards evaluation made use of improved experimental data and some developments in the methodology of analysis and evaluation. In addition to the work on the traditional standards, this work produced the extension of some energy ranges and includes new reactions that are called reference cross sections. Since the effort extends beyond the traditional standards, it is called the neutron data standards evaluation. This international effort has produced new evaluations of the following cross section standards: the H(n,n), 6Li(n,t), 10B(n,α), 10B(n,α1 γ), natC(n,n), Au(n,γ), 235U(n,f) and 238U(n,f). Also in the evaluation process the 238U(n,γ) and 239Pu(n,f) cross sections that are not standards were evaluated. Evaluations were also obtained for data that are not traditional standards: the Maxwellian spectrum averaged cross section for the Au(n,γ) cross section at 30 keV; reference cross sections for prompt γ-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 252Cf spontaneous fission neutron spectrum and the 235U prompt fission neutron spectrum induced by thermal incident neutrons; and the thermal neutron constants. The data and covariance matrices of the uncertainties were obtained directly from the evaluation procedure.
Energy Technology Data Exchange (ETDEWEB)
Chadwick, M.B.; Young, P.G.
1995-07-01
We present evaluations of the interaction of 20 to 100 MeV neutrons with oxygen and nitrogen nuclei, which follows on from our previous work on carbon. Our aim is to accurately represent integrated cross sections, inclusive emission spectra, and kerma factors, in a data library which can be used in radiation transport calculations. We apply the FKK-GNASH nuclear model code, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms, and use experimental measurements to optimize the calculations. We determine total, elastic, and nonelastic cross sections, angle-energy correlated emission spectra, for light ejectiles with A{<=}4 and gamma-rays, and average energy depositions. Our results for charged-particle emission spectra agree well with the measurements of Subramanian et al.. We compare kerma factors derived from our evaluated cross sections with experimental data, providing an integral benchmarking of our work. The evaluated data libraries are available as electronic files.
Energy Technology Data Exchange (ETDEWEB)
Sirakov, I. [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Capote, R.; Trkov, A. [International Atomic Energy Agency, NAPC-Nuclear Data Section, Vienna (Austria); Gritzay, O. [Institute for Nuclear Research, Kyiv (Ukraine); Kim, H.I. [Korea Atomic Energy Research Institute, Nuclear Data Center, Daejeon (Korea, Republic of); Kopecky, S.; Paradela, C.; Schillebeeckx, P. [European Commission, Joint Research Centre, Geel (Belgium); Kos, B. [Jozef Stefan Institute, Ljubljana (Slovenia); Pronyaev, V.G. [Rosatom State Corporation, Atomsrandart, Moscow (Russian Federation)
2017-10-15
Cross sections for neutron interactions with {sup 238}U in the energy region from 5 keV to 150 keV have been evaluated. Average total and capture cross sections have been derived from a least squares analysis using experimental data reported in the literature. The resulting cross sections have been parameterised in terms of average resonance parameters maintaining full consistency with results of optical model calculations by using a dispersive coupled channel optical model potential. The average compound partial cross sections have been expressed in terms of transmission coefficients by applying the Hauser-Feshbach statistical reaction theory including width-fluctuations. A generalized single-level representation compatible with the energy-dependent options of the ENDF-6 format has been applied using standard boundary conditions. The results have been transferred into a full ENDF-6 compatible data file. (orig.)
Total neutron cross section for 181Ta
Directory of Open Access Journals (Sweden)
Schilling K.-D.
2010-10-01
Full Text Available The neutron time of flight facility nELBE, produces fast neutrons in the energy range from 0.1 MeV to 10 MeV by impinging a pulsed relativistic electron beam on a liquid lead circuit [1]. The short beam pulses (∼10 ps and a small radiator volume give an energy resolution better than 1% at 1 MeV using a short flight path of about 6 m, for neutron TOF measurements. The present neutron source provides 2 ⋅ 104 n/cm2s at the target position using an electron charge of 77 pC and 100 kHz pulse repetition rate. This neutron intensity enables to measure neutron total cross section with a 2%–5% statistical uncertainty within a few days. In February 2008, neutron radiator, plastic detector [2] and data acquisition system were tested by measurements of the neutron total cross section for 181Ta and 27Al. Measurement of 181Ta was chosen because lack of high quality data in an anergy region below 700 keV. The total neutron cross – section for 27Al was measured as a control target, since there exists data for 27Al with high resolution and low statistical error [3].
Energy Technology Data Exchange (ETDEWEB)
Chadwick, M.B. [California Univ., Livermor, CA (United States). Lawrence Livermore National Lab.; Young, P.G.
1997-03-01
We present evaluations of the interaction of neutrons with energies between 20 and 100 MeV with oxygen and nitrogen nuclei, which follows on from our previous work on carbon. Our aim is to accurately represent integrated cross sections, inclusive emission spectra, and kerma factors, in a data library which can be used in radiation transport calculations. We apply the FKK-GNASH nuclear model code, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms, and use experimental measurements to optimize the calculations. We determine total, elastic, and nonelastic cross sections, angle-energy correlated emission spectra for light ejectiles with A {<=} 4 and gamma-rays, and average energy depositions. Our results for charged-particle emission spectra agree well with the measurements of Subramanian et al. We compare kerma factors derived from our evaluated cross sections with experimental data, providing an integral benchmarking of our work. (author). 52 refs.
Evaluation of the 237Np neutron cross sections in the energy range from 10-5 eV to 5 MeV
International Nuclear Information System (INIS)
Derrien, H.; Fort, E.
1979-01-01
The 237 Np neutron cross-sections have been evaluated in the energy range from thermal to 5 MeV. A set of resonance parameters including a negative level, is recommanded after examination of the available experimental data. This set is used 1) to calculate the cross-sections from the thermal region to 150 ev, and 2) to provide the statistical parameters suitable to the calculations in the unresolved region. At higher energies, the transmission coefficients Te are calculated by the coupled channel optical model code ECIS. They are then used as input in the statistical model code FISINGA. The optical model parameters, including the deformation parameters, are those used by Lagrange for the Pu isotopes, slightly modified to reproduce at 40 KeV the total cross-sections obtained from the pure statistical parameters. The recommendations of Lynn concerning the level density parameters have been used. In this paper we describe the various steps of the evaluation
Energy Technology Data Exchange (ETDEWEB)
Dossantos-Uzarralde, P.; Le Luel, C.; Bauge, E. [CEA Bruyeres le Chatel, 91 (France). Dept. de Physique Theorique et Appliquee
2004-07-01
This paper presents neutron cross section evaluations of Europium isotopes. The cross sections are evaluated in 1 keV - 30 MeV energy range for the isotopes {sup 146}Eu, {sup 147}Eu, {sup 148}Eu, {sup 149}Eu, {sup 150}Eu, {sup 151}Eu, {sup 152}Eu, {sup 153}Eu, {sup 154}Eu in their ground state. This evaluation includes cross section productions of the long life isomeric states. Special attention is put on the options used for the description of the files written in ENDF-6 format. The final issue is a proposal of a new breed of ENDF-6 formatted neutron activation file. (authors)
International Nuclear Information System (INIS)
Cullen, D.E.
1979-01-01
A code, SIGMAL, to Doppler-broaden evaluated cross sections in the ENDL format was designed. This code can Doppler-broaden cross sections that result from neutrons, protons, deuterons, tritons, 3 He, or alpha particles incident on any target nuclei. The code allows broadening to up to 100 final temperatures, either directly from the initial temperature or by bootstrapping to successively higher temperatures. 6 figures, 2 tables
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)
Evaluation of cross sections and calculation of kerma factors for neutrons up to 80 MeV on {sup 12}C
Energy Technology Data Exchange (ETDEWEB)
Harada, M.; Watanabe, Y. [Kyushu Univ., Fukuoka (Japan); Chiba, S.; Fukahori, T.
1997-03-01
We have evaluated the cross sections for neutrons with incident energies from 20 to 80 MeV on {sup 12}C for the JENDL high-energy file. The total cross sections were determined by a generalized least-squares method with available experimental data. The cross sections of elastic and inelastic scattering to the first 2{sup +} were evaluated with the theoretical calculations. The optical potentials necessary for these calculations were derived using a microscopic approach by Jeukenne-Lejeune-Mahaux. For the evaluation of double differential emission cross sections (DDXs), we have developed a code system SCINFUL/DDX in which total 35 reactions including the 3-body simultaneous breakup process (n+{sup 12}C {yields} n+{alpha}+{sup 8}Be) can be taken into consideration in terms of a Monte Carlo method, and have calculated the DDXs of all light-emissions (A{<=}4) and heavier reaction products. The results for protons, deuterons, and alphas showed overall good agreement with experimental data. The code is also applicable for calculations of total and partial kerma factors. Total kerma factors calculated for energies from 20 to 80 MeV were compared with the measurements and the other latest evaluations from the viewpoints of medical application and nuclear heating estimation. (author)
Neutron standard cross sections in reactor physics - Need and status
International Nuclear Information System (INIS)
Carlson, A.D.
1990-01-01
The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community
Status of neutron cross sections for reactor dosimetry
International Nuclear Information System (INIS)
Vlasov, M.F.; Fabry, A.; McElroy, W.N.
1977-03-01
The status of current international efforts to develop standardized sets of evaluated energy-dependent (differential) neutron cross sections for reactor dosimetry is reviewed. The status and availability of differential data are considered, some recent results of the data testing of the ENDF/B-IV dosimetry file using 252 Cf and 235 U benchmark reference neutron fields are presented, and a brief review is given of the current efforts to characterize and identify dosimetry benchmark radiation fields
ACTIV87 Fast neutron activation cross section file 1987
International Nuclear Information System (INIS)
Manokhin, V.N.; Pashchenko, A.B.; Plyaskin, V.I.; Bychkov, V.M.; Pronyaev, V.G.; Schwerer, O.
1989-10-01
This document summarizes the content of the Fast Neutron Activation Cross Section File based on data from different evaluated data libraries and individual evaluations in ENDF/B-5 format. The entire file or selective retrievals from it are available on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author)
Measurements of neutron capture cross sections
International Nuclear Information System (INIS)
Nakajima, Yutaka
1984-01-01
A review of measurement techniques for the neutron capture cross sections is presented. Sell transmission method, activation method, and prompt gamma-ray detection method are described using examples of capture cross section measurements. The capture cross section of 238 U measured by three different prompt gamma-ray detection methods (large liquid scintillator, Moxon-Rae detector, and pulse height weighting method) are compared and their discrepancies are resolved. A method how to derive the covariance is described. (author)
Neutron cross section standards and instrumentation: Annual report
International Nuclear Information System (INIS)
1987-01-01
This annual report from the National Bureau of Standards contains a summary of the results of the Neutron Cross Section Standards and Instrumentation Program. The technical measurements for the past year are given along with the proposed program and budget needs for the next three years. The neutron standards measurements have concentrated on the most important 235 U(n,f) cross section in the thermal to 20 MeV energy range along with the development of neutron detectors required for these measurements. The NBS measurements have made a significant contribution to the improvement in the understanding of this reaction. Measurements were performed with numerous neutron detectors at overlapping energies and at different neutron sources in order to reduce the systematic errors to achieve the required accuracy in this important neutron standard. Significant progress was also made in the development of a detector to utilize the 3 He(n,p) reaction as a standard in the eV to MeV energy region. Improvements in data acquisition systems as well as additional studies of advanced neutron sources were accomplished. Contacts with private industry were maintained and coordination of the neutron standards evaluation was continued. The report also includes biographical listings of the research staff along with copies of a few of our recent publications. 13 figs., 1 tab
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.
Accurate measurements of neutron activation cross sections
International Nuclear Information System (INIS)
Semkova, V.
1999-01-01
The applications of some recent achievements of neutron activation method on high intensity neutron sources are considered from the view point of associated errors of cross sections data for neutron induced reaction. The important corrections in -y-spectrometry insuring precise determination of the induced radioactivity, methods for accurate determination of the energy and flux density of neutrons, produced by different sources, and investigations of deuterium beam composition are considered as factors determining the precision of the experimental data. The influence of the ion beam composition on the mean energy of neutrons has been investigated by measurement of the energy of neutrons induced by different magnetically analysed deuterium ion groups. Zr/Nb method for experimental determination of the neutron energy in the 13-15 MeV energy range allows to measure energy of neutrons from D-T reaction with uncertainty of 50 keV. Flux density spectra from D(d,n) E d = 9.53 MeV and Be(d,n) E d = 9.72 MeV are measured by PHRS and foil activation method. Future applications of the activation method on NG-12 are discussed. (author)
International Nuclear Information System (INIS)
Badikov, S.A.; Ignatyuk, A.V.; Zolotarev, K.I.; Pashchenko, A.B.
1993-11-01
The reaction cross-sections of 46 Ti(n,2n) 45 Ti and 54 Fe(n,2n) 53m+g Fe, which are important for fusion reactor neutron dosimetry, were evaluated using a generalized least squares method. The experimental cross-section data of all measurements performed up to January 1993, were critically reviewed. The evaluated cross-section data are presented in analytical form and in ENDF-6 format, including covariance data. (author)
Blokhin, A I; Manokhin, V N; Mikhajlyukova, M V; Nasyrova, S M; Skripova, M V
2001-01-01
The present paper analyses and evaluates the cross-sections of threshold reactions leading to the production of long-lived radionuclides during the irradiation, by thermonuclear spectrum neutrons, of steels containing V, Ti, Cr, Fe and Ni. On the basis of empirical systematics. a new evaluation of the (n,2n), (n,p), (n,np), (n,alpha) and (n,n alpha) excitation functions is made for all isotopes of V, Ti, Cr, Fe and Ni and for intermediate isotopes produced in the chain from irradiated isotopes up to production of the long-lived radionuclides sup 3 sup 9 Ar, sup 4 sup 2 Ar, sup 4 sup 1 Ca, sup 5 sup 3 Mn, sup 6 sup 0 Fe, sup 6 sup 0 Co, sup 5 sup 9 Ni and sup 6 sup 3 Ni. A comparison is made with the experimental and other evaluated data.
International Nuclear Information System (INIS)
Blokhin, A.I.; Buleeva, N.N.; Manokhin, V.N.; Mikhajlyukova, M.V.; Nasyrova, S.M.; Skripova, M.V.
2002-01-01
The present paper analyses and evaluates the cross-sections of threshold reactions leading to the production of long-lived radionuclides during the irradiation, by thermonuclear spectrum neutrons, of steels containing V, Ti, Cr, Fe and Ni. On the basis of empirical systematics. a new evaluation of the (n,2n), (n,p), (n,np), (n,α) and (n,nα) excitation functions is made for all isotopes of V, Ti, Cr, Fe and Ni and for intermediate isotopes produced in the chain from irradiated isotopes up to production of the long-lived radionuclides 39 Ar, 42 Ar, 41 Ca, 53 Mn, 60 Fe, 60 Co, 59 Ni and 63 Ni. A comparison is made with the experimental and other evaluated data. (author)
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
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)
New neutron cross sections for fusion materials studies
International Nuclear Information System (INIS)
Greenwood, L.R.; Smither, R.K.
1985-01-01
Neutron cross sections are being developed for a variety of fusion-related applications including neutron dosimetry, fusion plasma diagnostics, the activation of very long-lived isotopes, and high-energy accelerator neutron sources
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)
Total neutron cross section of lead
International Nuclear Information System (INIS)
Kanda, K.; Aizawa, O.
1976-01-01
The total thermal-neutron cross section of natural lead under various physical conditions was measured by the transmission method. It became clear that the total cross section at room temperature previously reported is lower than the present data. The total cross section at 400, 500, and 600 0 C, above the melting point of lead, 327 0 C, was also measured, and the changes in the cross section as a function of temperature were examined, especially near and below the melting point. The data obtained for the randomly oriented polycrystalline state at room temperature were in reasonable agreement with the theoretical values calculated by the THRUSH and UNCLE-TOM codes
Neutron Cross Sections for Aluminium
Energy Technology Data Exchange (ETDEWEB)
Forsberg, Leif
1963-08-15
Total, elastic, inelastic, (n, 2n), (n, {alpha}), (n, p), and (n, {gamma}) cross sections for aluminium have been compiled from thermal to 100 MeV based upon literature search and theoretical interpolations and estimates. Differential elastic cross sections in the centre of mass system are represented by the Legendre coefficients. This method was chosen in order to obtain the best description of the energy dependence of the anisotropy.
Neutron data evaluation for natural niobium
International Nuclear Information System (INIS)
Ma Gonggui; Zou Yiming; Wang Shiming
1992-01-01
The complete neutron nuclear data for natural niobium were evaluated based on both experimental data measured up to 1989 and calculated data with program MUP2 and AUJP. The present work was done for CENDL-2 and supersedes the CENDL-1 (MAT = 1411) evaluation. The following neutron data are given for Nb in the energy range 10 -5 eV to 20 MeV (MAT = 2411): total, elastic, nonelastic, total inelastic, inelastic cross sections to 13 discrete levels, inelastic continuum, (n,2n), (n,3n), (n,n'α) + (n,αn'), (n,n'p) + (n,pn'),(n,n'd) + (n,dn'), (n,p), (n,d), (n,t), (n,α) and capture cross sections. Derived data for MT = 251, 252 and 253 are also included. Angular distributions and energy spectra of secondary neutron are also given inelastic continuum,
Evaluated neutronic file for indium
International Nuclear Information System (INIS)
Smith, A.B.; Chiba, S.; Smith, D.L.; Meadows, J.W.; Guenther, P.T.; Lawson, R.D.; Howerton, R.J.
1990-01-01
A comprehensive evaluated neutronic data file for elemental indium is documented. This file, extending from 10 -5 eV to 20 MeV, is presented in the ENDF/B-VI format, and contains all neutron-induced processes necessary for the vast majority of neutronic applications. In addition, an evaluation of the 115 In(n,n') 116m In dosimetry reaction is presented as a separate file. Attention is given in quantitative values, with corresponding uncertainty information. These files have been submitted for consideration as a part of the ENDF/B-VI national evaluated-file system. 144 refs., 10 figs., 4 tabs
Neutron capture cross sections of Kr
Directory of Open Access Journals (Sweden)
Fiebiger Stefan
2017-01-01
Full Text Available Neutron capture and β− -decay are competing branches of the s-process nucleosynthesis path at 85Kr [1], which makes it an important branching point. The knowledge of its neutron capture cross section is therefore essential to constrain stellar models of nucleosynthesis. Despite its importance for different fields, no direct measurement of the cross section of 85Kr in the keV-regime has been performed. The currently reported uncertainties are still in the order of 50% [2, 3]. Neutron capture cross section measurements on a 4% enriched 85Kr gas enclosed in a stainless steel cylinder were performed at Los Alamos National Laboratory (LANL using the Detector for Advanced Neutron Capture Experiments (DANCE. 85Kr is radioactive isotope with a half life of 10.8 years. As this was a low-enrichment sample, the main contaminants, the stable krypton isotopes 83Kr and 86Kr, were also investigated. The material was highly enriched and contained in pressurized stainless steel spheres.
Measurement of thermal neutron capture cross section
International Nuclear Information System (INIS)
Huang Xiaolong; Han Xiaogang; Yu Weixiang; Lu Hanlin; Zhao Wenrong
2001-01-01
The thermal neutron capture cross sections of 71 Ga(n, γ) 72 Ga, 94 Zr(n, γ) 95 Zr and 191 Ir(n, γ) 192 Ir m1+g,m2 reactions were measured by using activation method and compared with other measured data. Meanwhile the half-life of 72 Ga was also measured. The samples were irradiated with the neutron in the thermal column of heavy water reactor of China Institute of Atomic Energy. The activities of the reaction products were measured by well-calibrated Ge(Li) detector
Evaluation of Neutron shielding efficiency of Metal hydrides
Energy Technology Data Exchange (ETDEWEB)
Jeon, Sang Hwan; Chae, San; Kim, Yong Soo [Hanyang University, Seoul (Korea, Republic of)
2012-05-15
Neutron shielding is achieved of interaction with material by moderation and absorption. Material that contains large amounts hydrogen atoms which are almost same neutron atomic weight is suited for fast neutron shielding material. Therefore, polymers containing high density hydrogen atom are being used for fast neutron shielding. On the other hand, composite materials containing high thermal neutron absorption cross section atom (Li, B, etc) are being used for thermal neutron shielding. However, these materials have low fast neutron absorption cross section. Therefore, these materials are not suited for fast neutron shielding. Hydrogen which has outstanding neutron energy reduction ability has very low thermal neutron absorption cross section, almost cannot be used for thermal neutron shielding. In this case, a large atomic number material (Pb, U, etc.) has been used. Thus, metal hydrides are considered as complement to concrete shielding material. Because metal hydrides contain high hydrogen density and elements with high atomic number. In this research neutron shielding performance and characteristic of nuclear about metal hydrides ((TiH{sub 2}, ZrH{sub 2}, HfH{sub 2}) is evaluated by experiment and MCNPX using {sup 252}Cf neutron source as purpose development shielding material to developed shielding material
International Nuclear Information System (INIS)
Ilieva, K.; Belousov, S.; Apostolov, T.
1998-01-01
The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)
Neutron cross section standards and instrumentation. Annual report
Energy Technology Data Exchange (ETDEWEB)
Wasson, O.A.
1993-07-01
The objective of this interagency program is to provide accurate neutron interaction measurements for the US Department of Energy nuclear programs which include waste disposal, fusion, safeguards, defense, fission, and personnel protection. These measurements are also useful to other energy programs which indirectly use the unique properties of the neutron for diagnostic and analytical purposes. The work includes the measurement of reference cross sections and related neutron data employing unique facilities and capabilities at NIST and other laboratories as required; leadership and participation in international intercomparisons and collaborations; the preservation of standard reference deposits and the development of improved neutron detectors and measurement methods. A related and essential element of the program is critical evaluation of neutron interaction data including international coordinations. Data testing of critical data for important applications is included. The program is jointly supported by the Department of Energy and the National Institute of Standards and Technology. This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the third year of this three-year interagency agreement. The proposed program and required budget for the following three years are also presented. The program continues the shifts in priority instituted in order to broaden the program base.
Neutron cross section standards and instrumentation. Annual report
International Nuclear Information System (INIS)
Wasson, O.A.
1993-01-01
The objective of this interagency program is to provide accurate neutron interaction measurements for the US Department of Energy nuclear programs which include waste disposal, fusion, safeguards, defense, fission, and personnel protection. These measurements are also useful to other energy programs which indirectly use the unique properties of the neutron for diagnostic and analytical purposes. The work includes the measurement of reference cross sections and related neutron data employing unique facilities and capabilities at NIST and other laboratories as required; leadership and participation in international intercomparisons and collaborations; the preservation of standard reference deposits and the development of improved neutron detectors and measurement methods. A related and essential element of the program is critical evaluation of neutron interaction data including international coordinations. Data testing of critical data for important applications is included. The program is jointly supported by the Department of Energy and the National Institute of Standards and Technology. This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the third year of this three-year interagency agreement. The proposed program and required budget for the following three years are also presented. The program continues the shifts in priority instituted in order to broaden the program base
Numerical estimates of multiple reaction corrections in neutron cross-section measurements
International Nuclear Information System (INIS)
Magnusson, G.
1979-04-01
A method to evaluate the effect of secondary neutrons in 14-15 MeV neutron cross-section measurements is presented. The emission spectra of secondary neutrons are calculated by means of the preequilibrium and statistical models. An expression for the collision probability in a homogenous body has been utilized in the calculations. (author)
Neutron cross-sections of deuterium in the energy range 0.0001eV-15MeV
International Nuclear Information System (INIS)
Bazazyants, N.O.; Zabrodskaya, A.S.; Larina, A.F.; Nikolaev, M.N.
1978-08-01
The paper describes the evaluation of deuterium neutron cross-sections, the spectra of neutrons from the reaction D(n,2n)P and the angular distributions of neutrons from this reaction and of neutrons elastically scattered on deuterium. The evaluation results are presented in the SOCRATOR format. The 26-group system of constants for deuterium is also presented. (author)
Measurements of neutron spallation cross section. 2
Energy Technology Data Exchange (ETDEWEB)
Kim, E.; Nakamura, T. [Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center; Imamura, M.; Nakao, N.; Shibata, S.; Uwamino, Y.; Nakanishi, N.; Tanaka, Su.
1997-03-01
Neutron spallation cross section of {sup 59}Co(n,xn){sup 60-x}Co, {sup nat}Cu(n,sp){sup 56}Mn, {sup nat}Cu(n,sp){sup 58}Co, {sup nat}Cu(n,xn){sup 60}Cu, {sup nat}Cu(n,xn){sup 61}Cu and {sup nat}Cu(n,sp){sup 65}Ni was measured in the quasi-monoenergetic p-Li neutron fields in the energy range above 40 MeV which have been established at three AVF cyclotron facilities of (1) INS of Univ. of Tokyo, (2) TIARA of JAERI and (3) RIKEN. Our experimental data were compared with the ENDF/B-VI high energy file data by Fukahori and the calculated cross section data by Odano. (author)
International Nuclear Information System (INIS)
Boulaich, Y.; Bardouni, C.; Elyounoussi, C.; Elbakkari, H.; Boukhal, H.; Erradi, L.; Nacir, B.
2011-01-01
Full text: In this work, we present our analysis of the CREOLE experiment on the parameter by using the three-dimensional continuous energy code (MCNPS) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA-Cadarache, was dedicated to studies for both UO2 and UO2-PuO2 PWR type lattices covering the whole temperature range from 20 0 C to 300 0 C. We have developed an accurate model of the EOLE reactor to be used by the MCNP5 Monte Carlo code. This model guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII. JEFF3.1, JENDL3.3 and JENDL4 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the reactivity parameter were analyzed and the results have shown that the JENDL evaluations give the most consistent values. In order to specify the source of the relatively large difference between experiment and calculation due to ENDF-BVII nuclear data evaluation, the discrepancy in reactivity between ENDF-BVII and JENDL evaluations was decomposed using sensitivity and uncertainty analysis technique
Mizuyama, Kazuhito; Iwamoto, Nobuyuki; Iwamoto, Osamu; Hasemi, Hiroyuki; Kino, Koichi; Kimura, Atsushi; Kiyanagi, Yoshiaki
2017-09-01
In order to analyze the experimental data measured by the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Japan Proton Accelerator Research Complex/Materials and Life Science Experimental Facility (J-PARC/MLF), it is necessary to take into account the double-bunch structure of the neutron pulse and the energy resolution function for the operational condition of the J-PARC/MLF. The modified REFIT code has been developed to treat the double-bunch neutron pulse and the energy resolution function for J-PARC/MLF. In this study, we applied the modified REFIT code to analyze the new data of the neutron capture cross section of 155Gd and 157Gd recently measured by ANNRI in the J-PARC/MLF, and obtained the resonance parameters of two Gd isotopes. We discussed the differences between the our obtained results and the other libraries.
Energy Technology Data Exchange (ETDEWEB)
Baosheng, Yu; Qingbiao, Shen; Dunjiu, Cai [Chinese Nuclear Data Center, Beijing, BJ (China)
1996-06-01
The neutron monitor cross sections for {sup 59}Co(n,x){sup 56,57,58}Co, {sup 52,54,56}Mn, {sup 59}Fe reactions were evaluated based on recent experimental data and theoretical calculations from threshold energy to 100 MeV. (8 figs.).
International Nuclear Information System (INIS)
Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.
1994-01-01
The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values
Neutron data evaluation of 232U
International Nuclear Information System (INIS)
Maslov, V.M.; Porodzinskij, Yu.V.; Tetereva, N.A.; Kagalenko, A.B.; Kornilov, N.V.; Baba, M.; Hasegawa, A.
2003-03-01
Consistent evaluation of 232 U measured data base is performed. Hauser-Feshbach- Moldauer theory, coupled channel model and double-humped fission barrier model are employed. Total, differential scattering, fission and (n,xn) data are consistently reproduced as a major constraint for inelastic scattering cross section estimate. The direct excitation of ground state and higher band levels is calculated within rigid rotator and soft (deformable) rotator model, respectively. Prompt fission neutron spectra data are described. Average resonance parameters are provided, which reproduce evaluated cross sections in the range of 10-150 keV. (author)
The total neutron cross sections for 14N and 24Mg
International Nuclear Information System (INIS)
Bommer, J.
This report contains tables of the total neutron cross sections of 14 N and 24 Mg as determined in a recent measurement for neutron energies between 1 and 5.3 MeV. Graphic representations and details on the evaluation of the cross sections are included. (orig.) [de
Neutron cross section measurements at ORELA
International Nuclear Information System (INIS)
Dabbs, J.W.T.
1979-01-01
ORELA (Oak Ridge Electron Linear Accelerator) has been for the last decade the most powerful and useful pulsed neutron time-of-flight facility in the world, particularly in the broad midrange of neutron energies (10 eV to 1 MeV). This position will be enhanced with the addition of a pulse narrowing prebuncher, recently installed and now under test. Neutron capture, fission, scattering, and total cross sections are measured by members of the Physics and Engineering Physics Divisions of ORNL, and by numerous guests and visitors. Several fundamental and applied measurements are described, with some emphasis on instrumentation used. The facility comprises the accelerator and its target(s), 10 evacuated neutron flight paths having 18 measurement stations at flight path distances 8.9 to 200 meters, and a complex 4-computer data acquisition system capable of handling some 17,000 32-bit events/s from a total of 12 data input ports. The system provides a total of 2.08 x 10 6 words of data storage on 3 fast disk units. In addition, a dedicated PDP-10 timesharing system with a 250-megabyte disk system and 4 PDP-15 graphic display satellites permits on-site data reduction and analysis. More than 10 man-years of application software development supports the system, which is used directly by individual experiments. 12 figures, 1 table
Fast-neutron total and scattering cross sections of niobium
Energy Technology Data Exchange (ETDEWEB)
Smith, A.B.; Guenther, P.T.; Whalen, J.F.
1982-07-01
Neutron total cross sections of niobium were measured from approx. = 0.7 to 4.5 MeV at intervals of less than or equal to 50 keV with broad resolution. Differential-elastic-scattering cross sections were measured from approx. = 1.5 to 4.0 MeV at intervals of 0.1 to 0.2 MeV and at 10 to 20 scattering angles distributed between approx. = 20 and 160 degrees. Inelastically-scattered neutrons, corresponding to the excitation of levels at: 788 +- 23, 982 +- 17, 1088 +- 27, 1335 +- 35, 1504 +- 30, 1697 +- 19, 1971 +- 22, 2176 +- 28, 2456 +- (.), and 2581 +- (.) keV, were observed. An optical-statistical model, giving a good description of the observables, was deduced from the measured differential-elastic-scattering cross sections. The experimental-results were compared with the respective evaluated quantities given in ENDF/B-V.
Fast-neutron total and scattering cross sections of niobium
International Nuclear Information System (INIS)
Smith, A.B.; Guenther, P.T.; Whalen, J.F.
1982-07-01
Neutron total cross sections of niobium were measured from approx. = 0.7 to 4.5 MeV at intervals of less than or equal to 50 keV with broad resolution. Differential-elastic-scattering cross sections were measured from approx. = 1.5 to 4.0 MeV at intervals of 0.1 to 0.2 MeV and at 10 to 20 scattering angles distributed between approx. = 20 and 160 degrees. Inelastically-scattered neutrons, corresponding to the excitation of levels at: 788 +- 23, 982 +- 17, 1088 +- 27, 1335 +- 35, 1504 +- 30, 1697 +- 19, 1971 +- 22, 2176 +- 28, 2456 +- (.), and 2581 +- (.) keV, were observed. An optical-statistical model, giving a good description of the observables, was deduced from the measured differential-elastic-scattering cross sections. The experimental-results were compared with the respective evaluated quantities given in ENDF/B-V
Curves and tables of neutron cross sections of fission product nuclei in JENDL-3
Energy Technology Data Exchange (ETDEWEB)
Nakagawa, Tsuneo [ed.
1992-06-15
Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10{sup {minus}5} eV to 20 MeV. Almost of the cross section data reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in order tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum.
Curves and tables of neutron cross sections of fission product nuclei in JENDL-3
International Nuclear Information System (INIS)
Nakagawa, Tsuneo
1992-06-01
Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10 -5 eV to 20 MeV. Almost all the cross section data are reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in other tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum. (author)
International Nuclear Information System (INIS)
Pronyaev, V.; Carlson, A.D.; Capote Noy, R.; Wallner, A.
2011-03-01
The meeting participants have considered the progress in the measurement and evaluation of neutron cross sections and spectra which can be used as standard or reference data. This includes extension of the 197 Au(n,γ) standard to the energy range below 200 keV, 235 U(n th ,f) prompt fission neutron spectrum and neutron induced gamma-production cross sections. The work on this data development project for next two years has been agreed. (author)
International Nuclear Information System (INIS)
Keinert, J.
1982-06-01
In providing THERM-126 with cross section matrices for deuterium bound in heavy water the IKE phonon spectrum was reevaluated. The changes are modifications in the acoustic part and in the frequency of the second oscillator. Contrary to the phonon spectrum model for D in D 2 O in ENDF/B-IV the broad band of hindered rotations is assumed to be temperature dependent taking into account the diffusive motion of the molecule. With the new model scattering law data S (α, β) are generated in the temperature range 293.6 K-673.6 K. The THERM-126 scattering cross section matrices are calculated up to P 3 . As a validity check a lot of differential and integral cross sections are compared to experiments and benchmarks are recalculated. (orig.) [de
Influence of cross-section structure on unfolded neutron spectra
International Nuclear Information System (INIS)
Ertek, C.; Vlasov, M.F.; Cross, B.; Smith, P.M.
1979-01-01
The influence of cross-section structure on neutron spectra unfolded by multiple foil activation technique, SAND-II case, has been studied. For three reactions with evident structure in neutron cross-section above threshold: 27Al(n,α)24Na, 31P(n,p)31Si and 32S(n,p)32P, two remarkably different sets of evaluated data were selected from the available evaluations; one set of data was ''smooth'', the structure having been averaged over by a smooth curve; the other set was ''sharp'' with structure given in detail. These data were used in unfolding procedure together with other reactions, the same in both cases (as well as input spectra and measured reaction rates). It was found that during unfolding calculations less iteration steps were needed to unfold the neutron flux spectrum with the set of ''sharp'' data. In case of ''smooth'' data it was difficult to obtain an agreement between measured and calculated activity values even by increasing the number of iteration steps. Contrary to expectations, considerable deformation of unfolded neutron flux spectrum has been observed in the case of the ''smooth'' data set. (author)
Neutron capture cross section of $^{93}$Zr
We propose to measure the neutron capture cross section of the radioactive isotope $^{93}$Zr. This project aims at the substantial improvement of existing results for applications in nuclear astrophysics and emerging nuclear technologies. In particular, the superior quality of the data that can be obtained at n_TOF will allow on one side a better characterization of s-process nucleosynthesis and on the other side a more accurate material balance in systems for transmutation of nuclear waste, given that this radioactive isotope is widely present in fission products.
Measurement of neutron total cross-sections for {sup nat}Dy at Pohang Neutron Facility
Energy Technology Data Exchange (ETDEWEB)
Shin, S. G.; Kye, Y. U.; Shvetsov, Valery; Cho, M. H. [POSTECH, Pohang (Korea, Republic of); Namkung, W.; Cho, M. H. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Kim, G. N. [Kyungpook National Univ., Daegu (Korea, Republic of); Lee, M. W. [Dongnam Inst. of radiological and Medical Science, Busan (Korea, Republic of)
2013-05-15
There are few measurements for Dy below 100 eV. Moreover, there exist discrepancies among the measurements. In the present work, the total neutron cross-sections for {sup nat}Dy were measured by using the time-of-flight (TOF) method at the Pohang Neutron Facility (PNF). The PNF consists of an electron linac, a water-cooled Ta target, and an 11-m-long TOF path. The characteristics of PNF are described elsewhere. We also briefly discuss the future plan to verify our experimental result. We have measured the total neutron cross-sections of {sup nat}Dy in the neutron energy region from 0.1 eV to 100 eV with the TOF method at the Po hang Neutron Facility. The present result is in good agreement with the previous data and the evaluated data in ENDF/B-VI. We would like to get resonance parameters by using SAMMY or REFIT codes.
[Fast neutron cross section measurements]: Progress report
International Nuclear Information System (INIS)
1988-01-01
As projected in our previous proposal, the past year on the cross section project at the University of Michigan has been one primarily of construction and assembly of our 14 MeV pulsed Neutron Facility. All the components of the system have now been either purchased or fabricated in our shop facilities and have been assembled in their final configuration. We are now in the process of testing the rf components that have been designed to deliver voltage to both the pulser and buncher stages. We expect that the system will be operational by the end of the current contract year. We have also accomplished the design and construction of several other major pieces of equipment that are needed to begin fast neutron time-of-flight measurements. These include the primary proton recoil detector, and a californium fission chamber needed in the efficiency calibration of the primary detector. We have also added considerable concrete shielding designed to lower the neutron background in the experimental area. 10 figs., 5 tabs
An evaluation of neutron nuclear data for natural lead
International Nuclear Information System (INIS)
Ma Gonggui; Zhou Yiming; Wang Shiming
1990-04-01
Complete neutron nuclear data for natural lead were evaluated based on both experimental data measured up to December 1985 and theoretical data calculated with program MUP2. The experimental data evaluation is mainly based on CENDL-1, except (n, 2n) cross section, for which there are newly measured data available. The data, in the neutron energy from 10 -5 eV to 20 MeV, contain cross section, angular distributions and energy spectra of secondary neutrons, and are in ENDF/B-4 format
International Nuclear Information System (INIS)
Boulaich, Y.; El Bardouni, T.; Erradi, L.; Chakir, E.; Boukhal, H.; Nacir, B.; El Younoussi, C.; El Bakkari, B.; Merroun, O.; Zoubair, M.
2011-01-01
Highlights: → In the present work, we have analyzed the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. → Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values. → In order to specify the source of the relatively large discrepancy in the case of ENDF-BVII nuclear data evaluation, the k eff discrepancy between ENDF-BVII and JENDL3.3 was decomposed by using sensitivity and uncertainty analysis technique. - Abstract: In the present work, we analyze the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA/Cadarache, was mainly dedicated to the RTC studies for both UO 2 and UO 2 -PuO 2 PWR type lattices covering the whole temperature range from 20 deg. C to 300 deg. C. We have developed an accurate 3D model of the EOLE reactor by using the MCNP5 Monte Carlo code which guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII, JEFF3.1 and JENDL3.3 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values; the discrepancy is
Measurement of the neutron capture cross-section of 232Th using the neutron activation technique
International Nuclear Information System (INIS)
Naik, H.; Singh, Sarbjit; Goswami, A.; Manchanda, V.K.; Prajapati, P.M.; Surayanarayana, S.V.; Nayak, B.K.; Sharma, S.C.; Jagadeesan, K.C.; Thakare, S.V.; Raj, D.; Ganesan, S.; Mulik, V.K.; Sivashankar, B.S.; Mukherjee, S.
2011-01-01
The 232 Th(n, γ) reaction cross-section at average neutron energies of 3.7±0.3 MeV and 9.85±0.38 MeV from the 7 Li(p, n) reaction has been determined for the first time using activation and off-line γ -ray spectrometric technique. The 232 Th(n, 2n) reaction cross-section at the average neutron energy of 9.85±0.38 MeV has been also determined using the same technique. The experimentally determined 232 Th(n, γ) and 232 Th(n, 2n) reaction cross-sections were compared with the evaluated data of ENDF/B-VII, JENDL-4.0 and JEFF-3.1 and were found to be in good agreement. The present data along with literature data in a wide range of neutron energies were interpreted in terms of competition between different reaction channels including fission. The 232 Th(n, γ) and 232 Th(n, 2n) reaction cross-sections were also calculated theoretically using the TALYS 1.2 computer code and were found to be slightly higher than the experimental data. (orig.)
Neutron Data Evaluation of 237Np
International Nuclear Information System (INIS)
Maslov, V.M.; Tetereva, N.A.; Kolesov, A.M.; Pronyaev, V.G.; Zolotarev, K.I.; Granier, T.; Hambsch, F.-J.
2010-11-01
The diverse measured data base of n+ 237 Np was evaluated using a statistical theory and generalized least squares codes. Consistent description of the total, fission and partial inelastic scattering data in 1-3 MeV energy range provides an important constraint for the absorption cross section, which is quite important for the robust estimate of the capture cross section in the 0.5-500 keV energy range. Important constraints for the measured capture cross section come from the average radiative S0 and S1 strength functions. The evaluated inelastic cross sections of available evaluations are in severe disagreement with measured data on the inelastic scattering of neutrons with excitation of specific groups of levels. A change of the inelastic data shape at E n ∼1.5 MeV might be explained by the sharp increase of the level density of the residual odd-even nuclide 237 Np due to the onset of three-quasi-particle excitations. The influence of exclusive (n, xnf) pre-fission neutrons on prompt fission neutron spectra (PFNS) and (n, xn) spectra is modeled. Contributions of emissive/non-emissive fission and exclusive spectra of (n, xnf) reactions are defined by a consistent description of the 237 Np (n, F), 237 Np (n, 2n) 236s Np reactions and the ratio of the yields of short-lived (1 - ) and long-lived (6 - ) 236 Np states measured at 14 MeV. Excited levels of 236 Np are modeled using predicted Gallher-Moshkowski doublets. This work is performed under the project Agreement B-1604 with the International Science and Technology Center (Moscow). The financing party is EU. Partial support of International Atomic Energy Agency under Research Contract 14809 is acknowledged by JINER. (author)
Sensitivity of neutron air transport to nitrogen cross section uncertainties
International Nuclear Information System (INIS)
Niiler, A.; Beverly, W.B.; Banks, N.E.
1975-01-01
The sensitivity of the transport of 14-MeV neutrons in sea level air to uncertainties in the ENDF/B-III values of the various Nitrogen cross sections has been calculated using the correlated sampling Monte Carlo neutron transport code SAMCEP. The source consisted of a 14.0- to 14.9-MeV band of isotropic neutrons and the fluences (0.5 to 15.0 MeV) were calculated at radii from 50 to 1500 metres. The maximum perturbations, assigned to the ENDF/B-III or base cross section set in the 6.0- to 14.5-MeV energy range were; (1) 2 percent to the total, (2) 10 percent to the total elastic, (3) 40 percent to the inelastic and absorption and (4) 20 percent to the first Legendre coefficient and 10 percent to the second Legendre coefficient of the elastic angular distribtuions. Transport calculations were carried out using various physically realistic sets of perturbed cross sections, bounded by evaluator-assigned uncertainties, as well as the base set. Results show that in some energy intervals at 1500 metres, the differential fluence level with a perturbed set differed by almost a factor of two from the differential fluence level with the base set. 5 figures
Summary of the Workshop on Neutron Cross Section Covariances
International Nuclear Information System (INIS)
Smith, Donald L.
2008-01-01
A Workshop on Neutron Cross Section Covariances was held from June 24-27, 2008, in Port Jefferson, New York. This Workshop was organized by the National Nuclear Data Center, Brookhaven National Laboratory, to provide a forum for reporting on the status of the growing field of neutron cross section covariances for applications and for discussing future directions of the work in this field. The Workshop focused on the following four major topical areas: covariance methodology, recent covariance evaluations, covariance applications, and user perspectives. Attention was given to the entire spectrum of neutron cross section covariance concerns ranging from light nuclei to the actinides, and from the thermal energy region to 20 MeV. The papers presented at this conference explored topics ranging from fundamental nuclear physics concerns to very specific applications in advanced reactor design and nuclear criticality safety. This paper provides a summary of this workshop. Brief comments on the highlights of each Workshop contribution are provided. In addition, a perspective on the achievements and shortcomings of the Workshop as well as on the future direction of research in this field is offered
International Nuclear Information System (INIS)
Wagner, M.; Vonach, H.
1990-01-01
These proceedings of a specialists' meeting on neutron activation cross sections for fission and fusion energy applications are divided into 4 sessions bearing on: - data needs: 4 conferences - experimental work: 11 conferences - theoretical work: 4 conferences - evaluation work: 5 conferences
Energy Technology Data Exchange (ETDEWEB)
Stankunas, Gediminas, E-mail: gediminas.stankunas@lei.lt [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Via E. Fermi, 45, 00044 Frascati, Rome (Italy); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Sjöstrand, Henrik; Conroy, Sean [Department of Physics and Astronomy, Uppsala University, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)
2015-07-11
The neutron activation technique is routinely used in fusion experiments to measure the neutron yields. This paper investigates the uncertainty on these measurements as due to the uncertainties on dosimetry and activation reactions. For this purpose, activation cross-sections were taken from the International Reactor Dosimetry and Fusion File (IRDFF-v1.05) in 640 groups ENDF-6 format for several reactions of interest for both 2.5 and 14 MeV neutrons. Activation coefficients (reaction rates) have been calculated using the neutron flux spectra at JET vacuum vessel, both for DD and DT plasmas, calculated by MCNP in the required 640-energy group format. The related uncertainties for the JET neutron spectra are evaluated as well using the covariance data available in the library. These uncertainties are in general small, but not negligible when high accuracy is required in the determination of the fusion neutron yields.
Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi
2012-01-01
The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. Copyright © 2011 Elsevier Ltd. All rights reserved.
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
A new evaluation of the neutron data standards
Directory of Open Access Journals (Sweden)
Carlson A.D.
2017-01-01
Full Text Available Evaluations are being done for the H(n,n, 6Li(n,t, 10B(n,αγ, 10B(n,α, C(n,n, Au(n,γ, 235U(n,f and 238U(n,f standard cross sections. Evaluations are also being done for data that are not traditional standards including: the Au(n,γ cross section at energies below where it is considered a standard; reference cross sections for prompt gamma-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 235U thermal neutron fission spectrum and the 252Cf spontaneous fission neutron spectrum and the thermal constants.
International Nuclear Information System (INIS)
Holden, N.E.
1994-01-01
A re-evaluation of all thermal neutron cross sections and neutron resonance integrals has been performed, utilizing the previous database of the ''Barn Book'' and all of the more recently published experiments. Only significant changes or previously undetermined values are recorded in this report. The source for each value is also recorded in the accompanying table
Evaluation of neutron nuclear data for 7Li
International Nuclear Information System (INIS)
Yu Baosheng; Cai Dunjiu
1993-01-01
New complete neutron data for 7 Li have been evaluated for CENDL-2. It updates our work in 1978 for CENDL-1 to extend incident neutron energy range from 10 -5 eV to 20 MeV. Combining new experimental data and benchmark testing, the tritium production cross section were improved. The comparison of present evaluation with ENDF/B-6 and JENDL-3 has been made
Neutron cross section measurements for the Fast Breeder Program
International Nuclear Information System (INIS)
Block, R.C.
1979-06-01
This research was concerned with the measurement of neutron cross sections of importance to the Fast Breeder Reactor. The capture and total cross sections of fission products ( 101 102 104 Ru, 143 145 Nd, 149 Sm, 95 97 Mo, Cs, Pr, Pd, 107 Pd, 99 Tc) and tag gases (Kr, 78 80 Kr) were measured up to 100 keV. Filtered neutron beams were used to measure the capture cross section of 238 U (with an Fe filter) and the total cross section of Na (with a Na filter). A radioactive neutron capture detector was developed. A list of publications is included
Evaluation for ENDF/B-IV of the neutron cross sections for 235U from 82 eV to 25 keV
International Nuclear Information System (INIS)
Peelle, R.W.
1976-05-01
Capture and fission cross sections for 235 U in the ''unresolved resonance'' energy region were evaluated to permit determination of local-average resonance parameters for the ENDF/B-IV cross section file. Microscopic data were examined for infinitely dilute average fission and capture cross sections and also for intermediate structure unlikely to be reproduced by statistical fluctuations of resonance widths and spacings within known laws. Evaluated cross sections, averaged over lethargy intervals greater than 0.1, were obtained as an average over selected data sets after appropriate renormalization. Estimated uncertainties are given for these evaluated average cross sections. The ''intermediate'' structure fluctuations common to a few independent data sets were approximated by straight lines joining successive cross sections at 120 selected energy points; the cross sections at the vertices were adjusted to reproduce the evaluated average cross sections over the broad energy regions. Data sources and methods are reviewed, output values are tabulated, and some modified procedures are suggested for future evaluations. Evaluated fission and capture integrals for the resolved resonance region are also tabulated. These are not in agreement with integrals based on the resonance parameters of ENDF/B versions III and IV. 8 tables, 5 figures
Energy Technology Data Exchange (ETDEWEB)
Baosheng, Yu; Qingbiao, Shen; Dunjiu, Cai [Chinese Nuclear Data Center, Beijing, BJ (China)
1996-06-01
The cross sections for {sup 90}Zr(n,x){sup 89,88}Zr and {sup 90}Zr(n,x){sup 88,87,86}Y reactions in intermediate energy region are useful in neutron field monitor, safety and material damage research. Below 20 MeV, the evaluated cross sections for {sup 90}Zr(n,2n){sup 89}Zr reaction are recommended based on the recent experimental data, including the new measured results in CIAE (Above 20 MeV). The measured cross sections are still insufficient to do evaluation. So the evaluation for {sup 90}Zr(n,x){sup 89,88}Zr and {sup 90}Zr(n,x){sup 88,87,86}Y reactions from threshold to 100 MeV are based on experimental and calculated data. (2 figs.).
Phenomenological dirac optical potential for neutron cross sections
Energy Technology Data Exchange (ETDEWEB)
Maruyama, Shin-ichi; Kitsuki, Hirohiko; Shigyo, Nobuhiro; Ishibashi, Kenji [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering
1997-03-01
Because of limitation on neutron-incident data, it is difficult to obtain global optical model potential for neutrons. In contrast, there are some global optical model potentials for proton in detail. It is interesting to convert the proton-incident global optical potentials into neutron-incident ones. In this study we introduce (N-Z)/A dependent symmetry potential terms into the global proton-incident optical potentials, and then obtain neutron-incident ones. The neutron potentials reproduce total cross sections in an acceptable degree. However, a comparison with potentials proposed by other authors brings about a confused situation in the sign of the symmetry terms. (author)
Evaluation of neutron nuclear data for 249Cf
International Nuclear Information System (INIS)
Zhou Delin; Yu Baosheng; Yuan Hanrong; Liu Tong; Zhang Jin; Su Zongdi; Yan Shiwei; Wang Cuilan; Zhang Jingshang
1992-01-01
A complete set of neutron nuclear data from 10 -5 eV to 20 MeV in ENDF/B-6 format has been evaluated for 249 Cf based on measured data, systematics predications and model theory calculations. The evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) reaction cross sections, the resolved and unresolved resonance parameters, the angular and energy distributions of emitted neutrons and the average number of neutrons emitted per fission. The numerical data are available in ENDF/B-6 format
Neutron Scattering Differential Cross Sections for 12C
Byrd, Stephen T.; Hicks, S. F.; Nickel, M. T.; Block, S. G.; Peters, E. E.; Ramirez, A. P. D.; Mukhopadhyay, S.; McEllistrem, M. T.; Yates, S. W.; Vanhoy, J. R.
2016-09-01
Because of the prevalence of its use in the nuclear energy industry and for our overall understanding of the interactions of neutrons with matter, accurately determining the effects of fast neutrons scattering from 12C is important. Previously measured 12C inelastic neutron scattering differential cross sections found in the National Nuclear Data Center (NNDC) show significant discrepancies (>30%). Seeking to resolve these discrepancies, neutron inelastic and elastic scattering differential cross sections for 12C were measured at the University of Kentucky Acceleratory Laboratory for incident neutron energies of 5.58, 5.83, and 6.04 MeV. Quasi mono-energetic neutrons were scattered off an enriched 12C target (>99.99%) and detected by a C6D6 liquid scintillation detector. Time-of-flight (TOF) techniques were used to determine scattered neutron energies and allowed for elastic/inelastic scattering distinction. Relative detector efficiencies were determined through direct measurements of neutrons produced by the 2H(d,n) and 3H(p,n) source reactions, and absolute normalization factors were found by comparing 1H scattering measurements to accepted NNDC values. This experimental procedure has been successfully used for prior neutron scattering measurements and seems well-suited to our current objective. Significant challenges were encountered, however, with measuring the neutron detector efficiency over the broad incident neutron energy range required for these measurements. Funding for this research was provided by the National Nuclear Security Administration (NNSA).
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
Neutron total scattering cross sections of elemental antimony
Energy Technology Data Exchange (ETDEWEB)
Smith, A.B.; Guenther, P.T.; Whalen, J.F.
1982-11-01
Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V.
Neutron total scattering cross sections of elemental antimony
International Nuclear Information System (INIS)
Smith, A.B.; Guenther, P.T.; Whalen, J.F.
1982-11-01
Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V
Recent progress in fast neutron activation cross section data
International Nuclear Information System (INIS)
Michaelis, W.
A brief review is given of some significant investigations performed during the past few years in the area of fast neutron activation cross sections that may be relevant for the use of nuclear techniques in the exploration of mineral resources, in process and quality control in industry as well as for general analytical purposes. Differential capture cross sections are considered for the natural elements or isotopes of Fe, Cu, Se, Y, Nb, Cd, In, Gd, W, Os and Au. Some of the data are compared with statistical model calculations. Experimental and evaluated average cross sections for capture and threshold reactions in the spontaneous fission neutron field of 252 Cf are reviewed taking into account the elements or isotopes of Mg, Al, Si, S, Ti, V, Mn, Fe, Co, Ni, Cu, Zn, Sr, Zr, Nb, Cd, In, Ba, Ta and Au. A summary of recent studies of differential cross sections for threshold reactions comprises data on Al, Si, S, Ti, Fe, Co, Ni, Cu, Zn, Zr, Nb, Ta, W and Au. Besides experimental investigations, evaluations and theoretical model calculations are considered. Cross sections at 14 MeV and in the region around this energy are reviewed for Na, Mg, Al, Cl, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Br, Sr, Zr, Nb, In, Er, Yb, Ta, W, Os, Ir, Au and Pb. Particular emphasis is laid on (n,p), (n,2n) and (n,α) reactions. (n,n') reactions are allowed for if the half-life of the metastable state excited permits elemental analyses by common experimental techniques. (orig.)
Neutron capture cross section standards for BNL-325
International Nuclear Information System (INIS)
Holden, N.E.
1980-01-01
The most common cross section standards for capture reactions in the thermal neutron energy region are gold, cobalt, and manganese. In preparation for the fourth edition of BNL-325, data on the thermal cross section and resonance integral were evaluated for these three standards. For gold, only measurements below the Bragg scattering cutoff were used and extrapolated to a neutron velocity of 2200 meters/second. A non 1/v correction due to the 4.9 eV resonance was made. The resonance integral is based on Jirlow's integral measurement and Tellier's parameters. The resonance integrals for cobalt and manganese are based solely on integral measurements because the capture widths of the first major resonance either vary by 20% in various measurements (cobalt), or have never been measured (manganese). Recommended thermal cross sections and resonance integrals are respectively gold: 98.65/plus or minus/0.9 barns, 1550/plus or minus/28 barns; cobalt: 37.18/plus or minus/0.06 barns, 74.2/plus or minus/2.0 barns and manganese: 13.3/plus or minus/0.2 barns, and 14.0/plus or minus/0.3 barns. 72 refs
International Nuclear Information System (INIS)
Van'kov, A.A.
1994-01-01
The contradiction between a measured integral neutron absorption cross-section averaged over a fast reactor spectrum and the corresponding value which was calculated with the use of evaluated microscopic cross-sections and a theoretical neutron spectrum has been investigated. The possible systematic error of a correction factor which takes into account multiple resonance neutron scattering in samples used in the measurement of the absorption cross-section is investigated. It is proposed that this error may be one of the main reason for the contradiction mentioned above which arises in the measurement of the 236 U neutron absorption cross-section. (author). 13 refs, 3 figs
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)
Cross section for inelastic neutron acceleration by 178Hfm2
International Nuclear Information System (INIS)
Karamyan, S.A.; Carroll, J.J.
2009-01-01
The scattering of thermal neutrons from isomeric nuclei may include events in which the outgoing neutrons have increased kinetic energy. This process has been called Inelastic Neutron Acceleration (INNA) and occurs when the final nucleus after emission of the neutron is left in a state with lower energy than that of the isomer. The result, therefore, is an induced depletion of the isomeric population to the ground state. A cascade of several gammas must accompany the neutron emission to release the high angular momentum of the initial isomeric state. INNA was previously observed in a few cases and the associated cross sections were only in modest agreement with theoretical estimates. The most recent measurement of an INNA cross section was σ INNA = (258 ± 58) b for neutron scattering by 177 Lu m . In the present work, an INNA cross section of σ INNA = 152 -36 +51 b was deduced from measurements of the total burn-up of the high-spin, four-quasiparticle isomer 178 Hf m2 during irradiation by thermal neutrons. Statistical estimates for the probability of different reaction channels past neutron absorption were used in the analysis, and the deduced σ INNA is compared to the theoretically predicted cross section
Evaluation of neutron nuclear data for 12C
International Nuclear Information System (INIS)
Shibata, Keiichi
1983-12-01
Neutron nuclear data of 12 C have been evaluated for JENDL-3 in the energy range from 10 -5 eV to 20 MeV. Evaluated quantities are the total, elastic and inelastic scattering, radiative capture, photon-production, (n,p), (n,d) and (n,α) reaction cross sections and the angular or energy distribution of neutrons and photons. The total cross section below the threshold energy of the inelastic scattering has been calculated on the basis of the R-matrix theory. Three discrete levels have been taken into account for the inelastic scattering. (author)
International Nuclear Information System (INIS)
Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi
2012-01-01
The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. - Highlights: ► The effects of silicon cross section and neutron spectrum on the radial uniformity in NTD were experimentally investigated. ► The numerical results using silicon single crystal cross section reveal good agreements. ► The radial uniformity in hard neutron spectrum was more flat than that in soft spectrum. ► The silicon single crystal cross section and hard neutron spectrum are recommended for numerical analyses and radial uniformity flattening in NTD, respectively.
The spin-spin effect in the total neutron cross section of polarized neutrons on polarized 165Ho
International Nuclear Information System (INIS)
Fasoli, U.; Galeazzi, G.; Pavan, P.; Toniolo, D.; Zago, G.; Zannoni, R.
1978-01-01
The spin-spin effect in the total neutron cross section of polarized neutrons on polarized 165 Ho has been measured in the energy interval 0.4 to 2.5 MeV, in perpendicular geometry. The results are consistent with zero effect. The spin-spin cross section sigmasub(ss) has been theoretically evaluated by a non-adiabatic coupled-channel calculation. From the comparison between the experimental and theoretical results a value Vsub(ss) = 9+-77 keV for the strength of the spin-spin potential has been obtained. Compound-nucleus effects do not seem to be relevant. (Auth.)
Expected anomalies of the neutron cross section near the liquid-glass transition
International Nuclear Information System (INIS)
Gotze, W.
1987-01-01
In the frameworks of a microscopic theory the anomalies of the neutron cross section near the liquid-glass transition are discussed. The central concept of the theory is the correlation function for density fluctuations of wave vector q and frequency ω. Its absorptive part is proportional to the dynamical structure factor S(q, ω), this is the scattering law for coherent neutron scattering. Tagged particle motion is evaluated as well and it yields the incoherent neutron scattering cross section S i (q, ω) in. The predictions of the theory for S(q, ω) and Si (q, ω) a q-ω domain are given
Thermal neutron capture cross sections of tellurium isotopes
International Nuclear Information System (INIS)
Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.
2003-01-01
New values for thermal neutron capture cross sections of the tellurium isotopes 122 Te, 124 Te, 125 Te, 126 Te, 128 Te, and 130 Te are reported. These values are based on a combination of newly determined partial γ-ray cross sections obtained from experiments on targets contained natural Te and γ intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given
Thermal neutron capture cross sections of tellurium isotopes
International Nuclear Information System (INIS)
Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.
2004-01-01
New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given
Thermal neutron capture cross sections of tellurium isotopes
Energy Technology Data Exchange (ETDEWEB)
Tomandl, I.; Honzatko, J.; von Egidy, T.; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.
2004-03-01
New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given.
Use of pulsed neutron logging to evaluate perforation washing
International Nuclear Information System (INIS)
Dimon, C.A.
1986-01-01
This invention relates to the use of pulsed neutron logging techniques before and after perforation washing operations are performed to evaluate the degree of success of the perforation washing operations. Well logging operations of a type designed to respond to the difference between a formation immediately behind the well sheath and voids in the formation are performed both before and after the perforation washing operation. differences between the two resulting logs are then indicative of voids created by perforation washing. In a preferred embodiment, pulsed neutron logging is used as the logging technique, while a weighted brine having a high absorption cross section to pulsed neutrons is used as the perforation washing fluid
Measurement of reaction cross sections of fission products induced by DT neutrons
Energy Technology Data Exchange (ETDEWEB)
Nakano, Daisuke; Murata, Isao; Takahashi, Akito [Osaka Univ., Suita (Japan)
1998-03-01
With the view of future application of fusion reactor to incineration of fission products, we have measured the {sup 129}I(n,2n){sup 128}I reaction cross section by DT neutrons with the activation method. The measured cross section was compared with the evaluated nuclear data of JENDL-3.2. From the result, it was confirmed that the evaluation overestimated the cross section by about 20-40%. (author)
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 ...
Neutron capture cross section measurements: case of lutetium isotopes
International Nuclear Information System (INIS)
Roig, O.; Meot, V.; Belier, G.
2011-01-01
The neutron radiative capture is a nuclear reaction that occurs in the presence of neutrons on all isotopes and on a wide energy range. The neutron capture range on Lutetium isotopes, presented here, illustrates the variety of measurements leading to the determination of cross sections. These measurements provide valuable fundamental data needed for the stockpile stewardship program, as well as for nuclear astrophysics and nuclear structure. Measurements, made in France or in United-States, involving complex detectors associated with very rare targets have significantly improved the international databases and validated models of nuclear reactions. We present results concerning the measurement of neutron radiative capture on Lu 173 , Lu 175 , Lu 176 and Lu 177m , the measurement of the probability of gamma emission in the substitution reaction Yb 174 (He 3 ,pγ)Lu 176 . The measurement of neutron cross sections on Lu 177m have permitted to highlight the process of super-elastic scattering
Personnel neutron dosimeter evaluation and upgrade program
International Nuclear Information System (INIS)
Fix, J.J.; Brackenbush, L.W.; McDonald, J.C.; Roberson, P.L.; Holbrook, K.L.; Endres, G.W.R.; Faust, L.G.
1983-01-01
Evaluation of neutron dosimeters from twelve DOE laboratories involved about 2500 dosimeter irradiations at both PNL and the National Bureau of Standards (NBS) using neutrons of several energies and doses and several irradiations for good statistical analysis. The data and their analyses will be published later. The information evaluates accuracy, precision, lower dose detection, and energy response of dosimeters
Computer codes for neutron data evaluation
International Nuclear Information System (INIS)
Nakagawa, Tsuneo
1979-01-01
Data compilation codes such as NESTOR and REPSTOR, and NDES (Neutron Data Evaluation System) are mainly discussed. NDES is a code for neutron data evaluation using a TSS terminal, TEKTRONIX 4014. Users of NDES can perform plotting of data and calculation with nuclear models under conversational mode. (author)
Vagelatos, Nicholas; Steinman, Donald K.; John, Joseph; Young, Jack C.
1981-01-01
A nuclear method and apparatus determines the temperature of a medium by injecting fast neutrons into the medium and detecting returning slow neutrons in three first energy ranges by producing three respective detection signals. The detection signals are combined to produce three derived indicia each systematically related to the population of slow neutrons returning from the medium in a respective one of three second energy ranges, specifically exclusively epithermal neutrons, exclusively substantially all thermal neutrons and exclusively a portion of the thermal neutron spectrum. The derived indicia are compared with calibration indicia similarly systematically related to the population of slow neutrons in the same three second energy ranges returning from similarly irradiated calibration media for which the relationships temperature, neutron absorption cross section and neutron moderating power to such calibration indicia are known. The comparison indicates the temperature at which the calibration indicia correspond to the derived indicia and consequently the temperature of the medium. The neutron absorption cross section and moderating power of the medium can be identified at the same time.
International Nuclear Information System (INIS)
Pokotilovskij, Yu.N.; Novopol'tsev, M.I.; Geltenbort, P.; Brenner, T.
2003-01-01
Some results of the test of the time-of-flight neutron spectrometers in the energy range (0.05-2.5)μeV are described. The measurements of total and differential cross sections were performed for several substances relevant to the experiments in the physics of ultracold neutrons: Zr, Al, polyethylene and liquid fluoropolymers
Estimation of 242Cm neutron cross sections for total file creation
International Nuclear Information System (INIS)
Bakhanovich, L.A.; Klepatskij, A.B.; Maslov, V.M.; Porodzinskij, Yu.V.; Sukhovitskij, E.Sh.
1989-01-01
Description of evaluation of 242 Cm neutron cross sections in 10 -5 eV-20 MeV energy range is given. Due to the lack of experimental data the evaluation is largely based on the application of theoretical models and systematics. The data obtained are compared to evaluations by other authors. 10 refs.; 2 figs.; 2 tabs
Re/Os cosmochronometer: measurement of neutron cross sections
International Nuclear Information System (INIS)
Mosconi, M.
2007-01-01
This experimental work is devoted to the improved assessment of the Re/Os cosmochronometer. The dating technique is based on the decay of 187 Re (t 1/2 =41.2 Gyr) into 187 Os and determines the age of the universe by the time of onset of nucleosynthesis. The nucleosynthesis mechanisms, which are responsible for the 187 Re/ 187 Os pair, provide the possibility to identify the radiogenic fraction of 187 Os exclusively by nuclear physics considerations. Apart from its radiogenic component, 187 Os can be synthesized otherwise only by the s process, which means that this missing fraction can be reliably determined and subtracted by proper s-process modeling. On the other hand, 187 Re is almost completely produced by the r process. The only information needed for the interpretation as a cosmic clock is the production rate of 187 Re as a function of time. The accuracy of the s-process calculations that are needed to determine the nucleosynthetic abundance of 187 Os depends on the quality of the neutron capture cross sections averaged over the thermal neutron spectrum at the s-process sites. Laboratory measurements of these cross sections have to be corrected for the effect of nuclear levels, which can be significantly populated at the high stellar temperatures during the s process. The neutron capture cross sections of 186 Os, 187 Os and 188 Os have been measured at the CERN n TOF facility in the range between 0.7 eV and 1 MeV. From these data, Maxwellian averaged cross sections have been determined for thermal energies from 5 to 100 keV with an accuracy around 4%, 3%, and 5% for 186 Os, 187 Os, and 188 Os, respectively. Since, the first excited state in 187 Os occurs at 9.75 keV, the cross section of this isotope requires a substantial correction for thermal population of low lying nuclear levels. This effect has been evaluated on the basis of resonance data derived in the (n, γ) experiments and by an improved measurements of the inelastic scattering cross section for
Re/Os cosmochronometer: measurement of neutron cross sections
Energy Technology Data Exchange (ETDEWEB)
Mosconi, M.
2007-12-21
This experimental work is devoted to the improved assessment of the Re/Os cosmochronometer. The dating technique is based on the decay of {sup 187}Re (t{sub 1/2}=41.2 Gyr) into {sup 187}Os and determines the age of the universe by the time of onset of nucleosynthesis. The nucleosynthesis mechanisms, which are responsible for the {sup 187}Re/{sup 187}Os pair, provide the possibility to identify the radiogenic fraction of {sup 187}Os exclusively by nuclear physics considerations. Apart from its radiogenic component, {sup 187}Os can be synthesized otherwise only by the s process, which means that this missing fraction can be reliably determined and subtracted by proper s-process modeling. On the other hand, {sup 187}Re is almost completely produced by the r process. The only information needed for the interpretation as a cosmic clock is the production rate of {sup 187}Re as a function of time. The accuracy of the s-process calculations that are needed to determine the nucleosynthetic abundance of {sup 187}Os depends on the quality of the neutron capture cross sections averaged over the thermal neutron spectrum at the s-process sites. Laboratory measurements of these cross sections have to be corrected for the effect of nuclear levels, which can be significantly populated at the high stellar temperatures during the s process. The neutron capture cross sections of {sup 186}Os, {sup 187}Os and {sup 188}Os have been measured at the CERN n TOF facility in the range between 0.7 eV and 1 MeV. From these data, Maxwellian averaged cross sections have been determined for thermal energies from 5 to 100 keV with an accuracy around 4%, 3%, and 5% for {sup 186}Os, {sup 187}Os, and {sup 188}Os, respectively. Since, the first excited state in {sup 187}Os occurs at 9.75 keV, the cross section of this isotope requires a substantial correction for thermal population of low lying nuclear levels. This effect has been evaluated on the basis of resonance data derived in the (n, {gamma
Secondary standard neutron detector for measuring total reaction cross sections
International Nuclear Information System (INIS)
Sekharan, K.K.; Laumer, H.; Gabbard, F.
1975-01-01
A neutron detector has been constructed and calibrated for the accurate measurement of total neutron-production cross sections. The detector consists of a polyethylene sphere of 24'' diameter in which 8- 10 BF 3 counters have been installed radially. The relative efficiency of this detector has been determined for average neutron energies, from 30 keV to 1.5 MeV by counting neutrons from 7 Li(p,n) 7 Be. By adjusting the radial positions of the BF 3 counters in the polyethylene sphere the efficiency for neutron detection was made nearly constant for this energy range. Measurement of absolute efficiency for the same neutron energy range has been done by counting the neutrons from 51 V(p,n) 51 Cr and 57 Fe(p,n) 57 Co reactions and determining the absolute number of residual nuclei produced during the measurement of neutron yield. Details of absolute efficiency measurements and the use of the detector for measurement of total neutron yields from neutron producing reactions such as 23 Na(p,n) 23 Mg are given
Update of the evaluation of the cross section of the neutron dosimetry reaction 103Rh(n,n')103mRh
International Nuclear Information System (INIS)
Pavlik, A.; Miah, M.M.H.; Strohmaier, B.; Vonach, H.
1995-10-01
On the occasion of a new measurement of the excitation function of the reaction 103 Rh(n,n') 103m Rh in the energy range between 5.69 and 12.0 MeV performed at the present institute in collaboration wit the PTB Braunschweig, the cross section of this reaction, which is part of the International Reactor Dosimetry Field (IRDF-90), was re-evaluated. Whereas the energy range of the evaluation, namely from threshold to 20 MeV, was kept unchanged with respect to IRDF-90, the underlying data base was extended by the experiment mentioned as well as by another measurement, and revised with regard to judgement and normalization of older data in the light of recent information. Based on the experimental data upgraded in this way, new model calculations were carried out, which in the energy region 14 - 20 MeV served to supplement the experimental cross sections for this evaluation. The cross sections and their uncertainties were evaluated in energy groups with widths of 0.2 to 1.0 MeV, and the relative correlation matrix of the evaluated cross sections at the different energies was calculated. The results presented here supersede the corresponding values published in Physics Data 13-5 and included to the IRDF-90. (author). 26 refs, 4 figs, 6 tabs
Principles and problems in neutron nuclear data evaluation
International Nuclear Information System (INIS)
Schmidt, J.J.
1967-01-01
The history of neutron nuclear data evaluation is briefly summarized. The physical problems involved in nuclear data evaluation, such as discrepancies and inconsistencies between different experimental data sets and gaps in experimental information, are discussed. The discrepancies in the capture cross-section data for molybdenum and iron are chosen to illustrate the great difficulties in systematizing and automatizing the evaluation process. The technical problems of data evaluation, such as computer storage and the establishment of nuclear data files, are not discussed. (author)
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
Energy Technology Data Exchange (ETDEWEB)
Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).
An evaluated neutronic data file for elemental cobalt
Energy Technology Data Exchange (ETDEWEB)
Guenther, P.; Lawson, R.; Meadows, J.; Sugimoto, M.; Smith, A.; Smith, D.; Howerton, R.
1988-08-01
A comprehensive evaluated neutronic data file for elemental cobalt is described. The experimental data base, the calculational methods, the evaluation techniques and judgments, and the physical content are outlined. The file contains: neutron total and scattering cross sections and associated properties, (n,2n) and (n,3n) processes, neutron radiative capture processes, charged-particle-emission processes, and photon-production processes. The file extends from 10/sup /minus/5/ eV to 20 MeV, and is presented in the ENDF/B-VI format. Detailed attention is given to the uncertainties and correlations associated with the prominent neutron-induced processes. The numerical contents of the file have been transmitted to the National Nuclear Data Center, Brookhaven National Laboratory. 143 refs., 16 figs., 5 tabs.
Porosity effects in the neutron total cross section of graphite
International Nuclear Information System (INIS)
Santisteban, J. R; Dawidowski, J; Petriw, S. N
2009-01-01
Graphite has been used in nuclear reactors since the birth of the nuclear industry due to its good performance as a neutron moderator material. Graphite is still an option as moderator for generation IV reactors due to its good mechanical and thermal properties at high operation temperatures. So, there has been renewed interest in a revision of the computer libraries used to describe the neutron cross section of graphite. For sub-thermal neutron energies, polycrystalline graphite shows a larger total cross section (between 4 and 8 barns) than predicted by existing theoretical models (0.2 barns). In order to investigate the origin of this discrepancy we measured the total cross section of graphite samples of three different origins, in the energy range from 0.001 eV to 10 eV. Different experimental arrangements and sample treatments were explored, to identify the effect of various experimental parameters on the total cross section measurement. The experiments showed that the increase in total cross section is due to neutrons scattered around the forward direction. We associate these small-angle scattered neutrons (SANS) to the porous structure of graphite, and formulate a very simple model to compute its contribution to the total cross section of the material. This results in an analytic expression that explicitly depends on the density and mean size of the pores, which can be easily incorporated in nuclear library codes. [es
A neutron detector for measurement of total neutron production cross sections
Energy Technology Data Exchange (ETDEWEB)
Sekharan, K K; Laumer, H; Kern, B D; Gabbard, F [Kentucky Univ., Lexington (USA). Dept. of Physics and Astronomy
1976-03-01
A neutron detector has been constructed and calibrated for the accurate measurement of total neutron production cross sections. The detector consists of a polyethylene sphere of 60 cm diameter in which eight /sup 10/BF/sub 3/ counters have been installed radially. The relative efficiency of this detector has been determined for average neutron energies from 30 keV to 1.5 MeV by counting neutrons from /sup 7/Li(p, n)/sup 7/Be. By adjusting the radial positions of the BF/sub 3/ counters in the polyethylene sphere the efficiency for neutron detection was made nearly constant for this energy range. Measurement of absolute efficiency for the same neutron energy range has been done by counting the neutrons from /sup 51/V(p, n)/sup 51/Cr and /sup 57/Fe(p, n)/sup 57/Co reactions and determining the absolute number of residual nuclei produced during the measurement of neutron yield. Details of absolute efficiency measurements and the use of the detector for determination of neutron production cross sections are given.
A neutron detector for measurement of total neutron production cross sections
International Nuclear Information System (INIS)
Sekharan, K.K.; Laumer, H.; Kern, B.D.; Gabbard, F.
1976-01-01
A neutron detector has been constructed and calibrated for the accurate measurement of total neutron production cross sections. The detector consists of a polyethylene sphere of 60 cm diameter in which eight 10 BF 3 counters have been installed radially. The relative efficiency of this detector has been determined for average neutron energies from 30 keV to 1.5 MeV by counting neutrons from 7 Li(p, n) 7 Be. By adjusting the radial positions of the BF 3 counters in the polyethylene sphere the efficiency for neutron detection was made nearly constant for this energy range. Measurement of absolute efficiency for the same neutron energy range has been done by counting the neutrons from 51 V(p, n) 51 Cr and 57 Fe(p, n) 57 Co reactions and determining the absolute number of residual nuclei produced during the measurement of neutron yield. Details of absolute efficiency measurements and the use of the detector for determination of neutron production cross sections are given. (Auth.)
Graphs of neutron cross section data for fusion reactor development
International Nuclear Information System (INIS)
Asami, Tetsuo; Tanaka, Shigeya
1979-03-01
Graphs of neutron cross section data relevant to fusion reactor development are presented. Nuclides and reaction types in the present compilation are based on a WRENDA request list from Japan for fusion reactor development. The compilation contains various partial cross sections for 55 nuclides from 6 Li to 237 Np in the energy range up to 20 MeV. (author)
Status report and measurement of total cross-sections at the Pohang Neutron Facility
International Nuclear Information System (INIS)
Kim, G.N.; Meaze, A.K.M.M.H.; Ahmed, H.
2004-01-01
We report the status of the Pohang Neutron Facility which consists of an electron linear accelerator, a water-cooled Ta target, and an 11-m time-of-flight path. It has been equipped with a four-position sample changer controlled remotely by a CAMAC data acquisition system, which allows simultaneous accumulation of the neutron time of flight spectra from 4 different detectors. It is possible to measure the neutron total cross-sections in the neutron energy range from 0.1 eV to 100 eV by using the neutron time of flight method. A 6 LiZnS(Ag) glass scintillator was used as a neutron detector. The neutron flight path from the water-cooled Ta target to the neutron detector was 10.81±0.02 m. The background level was determined by using notch-filters of Co, In, Ta, and Cd sheets. In order to reduce the gamma rays from Bremsstrahlung and those from neutron capture, we employed a neutron-gamma separation system based on their different pulse shapes. The present measurements are in general agreement with the evaluated data in ENDF/B-VI. The resonance parameters were extracted from the transmission data from the SAMMY fitting and compared with the previous ones. (author)
Evaluation of neutron nuclear data for 16O
International Nuclear Information System (INIS)
Shibata, Keiichi; Chiba, Satoshi; Nakajima, Yutaka; Asami, Tetsuo; Murata, Toru; Kanda, Yukinori; Tanaka, Shigeya.
1990-02-01
Neutron nuclear data of 16 O have been evaluated for JENDL-3 in the energy range from 10 -5 eV to 20 MeV. Evaluated quantities are the total, elastic and inelastic scattering, (n,2n), (n,γ), (n,p), (n,d), (n,α), (n,np) and (n,nα) reaction cross sections and the angular and energy distributions of emitted neutrons and gamma-rays. The total cross section below 3 MeV was calculated on the basis of the R-matrix theory. The inelastic scattering, (n,np) and (n,nα) reaction cross sections were obtained from the statistical model calculation. The gamma-ray production cross section was also calculated with the statistical model. (author)
Energy Technology Data Exchange (ETDEWEB)
Cullen, D.E.
1979-03-17
A code, SIGMAL, to Doppler-broaden evaluated cross sections in the ENDL format was designed. This code can Doppler-broaden cross sections that result from neutrons, protons, deuterons, tritons, /sup 3/He, or alpha particles incident on any target nuclei. The code allows broadening to up to 100 final temperatures, either directly from the initial temperature or by bootstrapping to successively higher temperatures. 6 figures, 2 tables.
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
Library of files of evaluated neutron data
International Nuclear Information System (INIS)
Blokhin, A.I.; Ignatyuk, A.V.; Koshcheev, V.N.; Kuz'minov, B.D.; Manokhin, V.N.; Manturov, G.N.; Nikolaev, M.N.
1988-01-01
It is reported about development of the evaluated neutron data files library which was recommended by the GKAE Nuclear Data Commission as the base of improving constant systems in neutron engeneering calculations. A short description of the library content is given and status of the library is pointed out
Differences between cross-section libraries for neutron dosimetry
International Nuclear Information System (INIS)
Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H.
2013-01-01
Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)
Average cross sections for the 252Cf neutron spectrum
International Nuclear Information System (INIS)
Dezso, Z.; Csikai, J.
1977-01-01
A number of average cross sections have been measured for 252 Cf neutrons in (n, γ), (n,p), (n,2n), (n,α) reactions by the activation method and for fission by fission chamber. Cross sections have been determined for 19 elements and 45 reactions. The (n,γ) cross section values lie in the interval from 0.3 to 200 mb. The data as a function of target neutron number increases up to about N=60 with minimum near to dosed shells. The values lie between 0.3 mb and 113 mb. These cross sections decrease significantly with increasing the threshold energy. The values are below 20 mb. The data do not exceed 10 mb. Average (n,p) cross sections as a function of the threshold energy and average fission cross sections as a function of Zsup(4/3)/A are shown. The results obtained are summarized in tables
Grammer, K. B.; Alarcon, R.; Barrón-Palos, L.; Blyth, D.; Bowman, J. D.; Calarco, J.; Crawford, C.; Craycraft, K.; Evans, D.; Fomin, N.; Fry, J.; Gericke, M.; Gillis, R. C.; Greene, G. L.; Hamblen, J.; Hayes, C.; Kucuker, S.; Mahurin, R.; Maldonado-Velázquez, M.; Martin, E.; McCrea, M.; Mueller, P. E.; Musgrave, M.; Nann, H.; Penttilä, S. I.; Snow, W. M.; Tang, Z.; Wilburn, W. S.
2015-05-01
Liquid hydrogen is a dense Bose fluid whose equilibrium properties are both calculable from first principles using various theoretical approaches and of interest for the understanding of a wide range of questions in many-body physics. Unfortunately, the pair correlation function g (r ) inferred from neutron scattering measurements of the differential cross section d/σ d Ω from different measurements reported in the literature are inconsistent. We have measured the energy dependence of the total cross section and the scattering cross section for slow neutrons with energies between 0.43 and 16.1 meV on liquid hydrogen at 15.6 K (which is dominated by the parahydrogen component) using neutron transmission measurements on the hydrogen target of the NPDGamma collaboration at the Spallation Neutron Source at Oak Ridge National Laboratory. The relationship between the neutron transmission measurement we perform and the total cross section is unambiguous, and the energy range accesses length scales where the pair correlation function is rapidly varying. At 1 meV our measurement is a factor of 3 below the data from previous work. We present evidence that these previous measurements of the hydrogen cross section, which assumed that the equilibrium value for the ratio of orthohydrogen and parahydrogen has been reached in the target liquid, were in fact contaminated with an extra nonequilibrium component of orthohydrogen. Liquid parahydrogen is also a widely used neutron moderator medium, and an accurate knowledge of its slow neutron cross section is essential for the design and optimization of intense slow neutron sources. We describe our measurements and compare them with previous work.
The determination of thermal neutron cross section of 81Br
International Nuclear Information System (INIS)
Kovacs, Luciana; Zamboni, Cibele B.; Dalaqua Junior, Leonardo
2009-01-01
In this investigation several standard materials were used to determine the thermal neutron cross section of 81 Br. This nuclear parameter is an important data to perform several quantitative investigations, mainly in medical area. In other to confirm and to reduce the uncertainty, a new measurement was preformed using thermal neutron at IEA-R1 nuclear reactor of IPEN/CNEN-SP. The result obtained is compatible with the tabulated value and present small uncertainly. (author)
Fast neutron capture cross section facility at Cadarache
International Nuclear Information System (INIS)
Le Rigoleur, C.; Arnaud, A.
1975-01-01
The total energy weighting technique has been applied to measure absolute fast neutron capture cross section at Cadarache. We use a non hydrogeneous liquid scintillator to detect the gamma from the cascade. The neutron flux is measured with a B 10 INa(Tl) detector or Li 6 glass scintillator of well known efficiency. Time of flight technique is used with on line digital computer data processing. (orig.) [de
Removal cross section for 14 mev neutrons in constructional materials
International Nuclear Information System (INIS)
Vasvary, L.; Divos, F.; Peto, G.; Csikai, J.; Mumba, N.K.
1985-01-01
Using flight time difference the direct and scattered neutrons and gammas produced in the target head and samples were separated. With this method the attenuation of primary neutrons and gammas originating from the target head has been studied. Thickness dependence of the secondary gamma yield from extended samples of Al, Fe, Pb, paraffin and reinforced concrete was also measured. Results indicate a geometry dependence of the removal cross sections
Nuclear Astrophysics and Neutron Cross Section Measurements Using the ORELA
Energy Technology Data Exchange (ETDEWEB)
Winters, R. R.
2000-08-25
This is the final report for a research program which has been continuously supported by the AEC, ERDA, or USDOE since 1973. The neutron total and capture cross sections for n + {sup 88}Sr have been measured over the neutron energy range 100 eV to 1 MeV. The report briefly summaries our results and the importance of this work for nucleosynthesis and the optical model.
Nuclear Astrophysics and Neutron Cross Section Measurements Using the ORELA
International Nuclear Information System (INIS)
Winters, R. R.
2000-01-01
This is the final report for a research program which has been continuously supported by the AEC, ERDA, or USDOE since 1973. The neutron total and capture cross sections for n + 88 Sr have been measured over the neutron energy range 100 eV to 1 MeV. The report briefly summaries our results and the importance of this work for nucleosynthesis and the optical model
Neutron-absorption cross section of sodium-22
International Nuclear Information System (INIS)
Rundberg, R.; Elgart, M.F.; Finston, H.L.; Williams, E.T.; Bond, A.H. Jr.
1975-01-01
A simple method for determining the neutron-absorption cross sections for radionuclides produced and consumed in a reactor-neutron flux is described. Data were obtained for 22 Na which, through application of Westcott's procedure, yielded the following: sigma 0 = 51.5 +- 3.1 kbarns, s 0 = 2.3 +- 0.1, and Σ' = 100 +- 10 kbarns. (3 tables) (U.S.)
Damage energy and displacement cross sections: survey and sensitivity. [Neutrons
Energy Technology Data Exchange (ETDEWEB)
Doran, D.G.; Parkin, D.M.; Robinson, M.T.
1976-10-01
Calculations of damage energy and displacement cross sections using the recommendations of a 1972 IAEA Specialists' Meeting are reviewed. The sensitivity of the results to assumptions about electronic energy losses in cascade development and to different choices respecting the nuclear cross sections is indicated. For many metals, relative uncertainties and sensitivities in these areas are sufficiently small that adoption of standard displacement cross sections for neutron irradiations can be recommended.
Microscopic cross-section measurements by thermal neutron activation
International Nuclear Information System (INIS)
Avila L, J.
1987-08-01
Microscopic cross sections measured by thermal neutron activation using RP-0 reactor at the Peruvian Nuclear Energy Institute. The method consists in measuring microscopic cross section ratios through activated samples, requiring being corrected in thermal and epithermal energetic range by Westcott formalism. Furthermore, the comptage ratios measured for each photopeak to its decay fraction should be normalized from interrelation between both processes above, activation microscopic cross sections are obtained
Precise measurements of neutron capture cross sections for FP
International Nuclear Information System (INIS)
Nakamura, Shoji; Harada, Hideo; Katoh, Toshio
2000-01-01
The thermal neutron capture cross sections (σ 0 ) and the resonance integrals (I 0 ) of some fission products (FP), such as 137 Cs, 90 Sr, 99 Tc, 129 I and 135 Cs, were measured by the activation and γ-ray spectroscopic methods. Moreover, the cross section measurements were done for other FP elements, such as 127 I, 133 Cs and 134 Cs. This paper provides the summary of the FP cross section measurements, which have been performed by authors. (author)
Neutron halo in 14B studied via reaction cross sections
International Nuclear Information System (INIS)
Fukuda, M.; Tanaka, M.; Iwamoto, K.; Wakabayashi, S.; Yaguchi, M.; Ohno, J.; Morita, Y.; Kamisho, Y.; Mihara, M.; Matsuta, K.; Nishimura, D.; Suzuki, S.; Nagashima, M.; Ohtsubo, T.; Ogura, T.; Abe, K.; Kikukawa, N.; Sakai, T.; Sera, D.; Takechi, M.; Izumikawa, T.; Suzuki, T.; Yamaguchi, T.; Sato, K.; Furuki, H.; Miyazawa, S.; Ichihashi, N.; Kohno, J.; Yamaki, S.; Kitagawa, A.; Sato, S.; Fukuda, S.
2014-01-01
Reaction cross sections (σ R ) for the neutron-rich nucleus 14 B on Be, C, and Al targets have been measured at several energies in the intermediate energy range of 45-120 MeV/nucleon. The present experimental σ R show a significant enhancement relative to the systematics of stable nuclei. The nucleon density distribution was deduced through the fitting procedure with the modified Glauber calculation. The necessity of a long tail in the density distribution was found, which is consistent with the valence neutron in 2s 1/2 orbital with the small empirical one-neutron separation energy in 14 B. (authors)
Nuclear fission and neutron-induced fission cross-sections
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
Theory of neutron resonance cross sections for safety applications
International Nuclear Information System (INIS)
Froehner, F.H.
1992-09-01
Neutron resonances exert a strong influence on the behaviour of nuclear reactors, especially on their response to the temperature changes accompanying power excursions, and also on the efficiency of shielding materials. The relevant theory of neutron resonance cross sections including the practically important approximations is reviewed, both for the resolved and the unresolved resonance region. Numerical techniques for Doppler broadening of resonances are presented, and the construction of group constants and especially of self-shielding factors for neutronics calculations is outlined. (orig.) [de
International Nuclear Information System (INIS)
Zalan, T.A.
1988-01-01
Multi-energy-group neutron diffusion theory is used to numerically evaluate the utility of two different dual-detector neutron porosity logging devices, a 14 MeV (accelerator) neutron source - epithermal neutron detector device and a 4 MeV neutron source - capture gamma-ray detector device, relative to the traditional 4 MeV neutron source - thermal neutron detector device. Fast and epithermal neutron diffusion parameters are calculated using Monte Carlo - derived neutron flux distributions. Thermal parameters are calculated from tabulated cross sections. An existing analytical method to describe the transport of gamma-rays through common earth materials is modified in order to accommodate the modeling of the 4 MeV neutron - capture gamma-ray device. The 14 MeV neutron - epithermal neutron device is found to be less sensitive to porosity than the 4 MeV neutron - capture gamma-ray device, which in turn is found to be less sensitive to porosity than the traditional 4 MeV neutron - thermal neutron device. Salinity effects are found to be comparable for the 4 MeV neutron - capture gamma-ray and 4 MeV neutron - thermal neutron devices. The 4 MeV neutron capture gamma-ray measurement is found to be deepest investigating
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.
International evaluated neutron nuclear data libraries
International Nuclear Information System (INIS)
Liu Tingjin
2001-01-01
The current status of five major evaluated neutron nuclear data libraries in the world are introduced. They are ENDF/B-6 (U. S. A.), JENDL-3.2 (Japan), JEF-2.2 (Europe), CENDL-2.1 (China), BROND-2 (Russia). The developing trend of the international neutron evaluated nuclear data library is discussed. How to get and use these data for the domestic users is given
Thermal neutron absorption cross section of small samples
International Nuclear Information System (INIS)
Nghiep, T.D.; Vinh, T.T.; Son, N.N.; Vuong, T.V.; Hung, N.T.
1989-01-01
A modified steady method for determining the macroscopic thermal neutron absorption cross section of small samples 500 cm 3 in volume is described. The method uses a moderating block of paraffin, Pu-Be neutron source emitting 1.1x10 6 n.s. -1 , SNM-14 counter and ordinary counting equipment. The interval of cross section from 2.6 to 1.3x10 4 (10 -3 cm 2 g -1 ) was measured. The experimental data are described by calculation formulae. 7 refs.; 4 figs
Determination of the fast-neutron-induced fission cross-section of 242Pu at nELBE
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.
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.
Neutron data evaluation of {sup 238}U
Energy Technology Data Exchange (ETDEWEB)
Maslov, V.M.; Porodzinskij, Y.V.; Hasegawa, Akira; Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-08-01
Cross sections for neutron-induced reactions on {sup 238}U are calculated by using the Hauser-Feshbach-Moldauer theory, the coupled channel model and the double-humped fission barrier model. The direct excitation of ground state band levels is calculated with a rigid-rotator model. The direct excitation of vibrational octupole and K = 2{sup +} quadrupole bands is included using a soft (deformable) rotator model. The competition of inelastic scattering to fission reaction is shown to be sensitive to the target nucleus level density at excitations above the pairing gap. As for fission, (n,2n), (n,3n), and (n,4n) reactions, secondary neutron spectra data are consistently reproduced. Pre-equilibrium emission of first neutron is included. Shell effects in the level densities are shown to be important for estimation of energy dependence of non-emissive fission cross section. (author). 105 refs.
Neutron total and scattering cross sections of 6Li in the few MeV region
International Nuclear Information System (INIS)
Smith, A.; Guenther, P.; Whalen, J.
1980-02-01
Neutron total cross sections of 6 Li are measured from approx. 0.5 to approx. 4.8 MeV at intervals of approx. 10 scattering angles and at incident-neutron intervals of approx.< 100 keV. Neutron differential inelastic-scattering cross sections are measured in the incident-energy range 3.5 to 4.0 MeV. The experimental results are extended to lower energies using measured neutron total cross sections recently reported elsewhere by the authors. The composite experimental data (total cross sections from 0.1 to 4.8 MeV and scattering cross sections from 0.22 to 4.0 MeV) are interpreted in terms of a simple two-level R-matrix model which describes the observed cross sections and implies the reaction cross section in unobserved channels; notably the (n;α)t reaction (Q = 4.783 MeV). The experimental and calculational results are compared with previously reported results as summarized in the ENDF/B-V evaluated nuclear data file
International Nuclear Information System (INIS)
Zolotarev, K.I.
2009-04-01
Evaluations of cross sections and their associated covariance matrices have been carried out for five dosimetry reactions: - excitation functions were re-evaluated for the 27 Al(n,α) 24 Na, 55 Mn(n,2n) 54 Mn and 90 Zr(n,2n) 89m+g Zr reactions over the neutron energy range from threshold to 40 MeV; - excitation functions were re-evaluated for the 59 Co(n,p) 59 Fe and 59 Co(n,2n) 58m+g Co reactions over the neutron energy range from threshold to 60 MeV. Uncertainties in the cross sections for all of those reactions were also derived in the form of relative covariance matrices. Benchmark calculations performed for 235 U thermal fission and 252 Cf spontaneous fission neutron spectra show that the integral cross sections calculated from the newly evaluated excitation functions exhibit improved agreement with related experimental data when compared with the equivalent data from the IRDF-2002 library. (author)
Neutron cross section covariances in the resonance region: 52Cr, 56Fe, 58Ni
Energy Technology Data Exchange (ETDEWEB)
Oblozinsky, P.; Cho, Y.-S.; Mattoon, C.M.; Mughabghab, S.F.
2010-08-03
We evaluated covariances for neutron capture and elastic scattering cross sections on major structural materials, {sup 52}Cr, {sup 56}Fe and {sup 58}Ni, in the resonance region which extends beyond 800 keV for each of them. Use was made of the recently developed covariance formalism based on kernel approximation along with data in the Atlas of Neutron Resonances. The data of most interest for AFCI applications, elastic scattering cross section uncertainties at energies above about few hundred keV, are on the level of about 12% for {sup 52}Cr, 7-8% for {sup 56}Fe and 5-6% for {sup 58}Ni.
Status of neutron dosimetry cross sections
International Nuclear Information System (INIS)
Griffin, P.J.; Kelly, J.G.
1992-01-01
Several new cross section libraries, such as ENDF/B-VI(release 2), IRDF-90,JEF-2.2, and JENDL-3 Dosimetry, have recently been made available to the dosimetry community. the Sandia National Laboratories (SNL) Radiation Metrology Laboratory (RML) has worked with these libraries since pre-release versions were available. this paper summarizes the results of the intercomparison and testing of dosimetry cross sections. As a result of this analysis, a compendium of the best dosimetry cross sections was assembled from the available libraries for use within the SNL RML. this library, referred to as the SNLRML Library, contains 66 general dosimetry sensors and 3 special dosimeters unique to the RML sensor inventory. The SNLRML cross sections have been put into a format compatible with commonly used spectrum determination codes
Cross sections for fast-neutron interaction with ytterbium isotopes
International Nuclear Information System (INIS)
Luo, Junhua; Liu, Rong; Jiang, Li; Ge, Suhong; Liu, Zhenlai; Sun, Guihua
2013-01-01
Highlights: ► The cross sections for the (n,x) reactions on ytterbium isotopes have been measured. ► Mono-energetic neutron beams using the D + T reaction; Energies: 13.5 and 14.8 MeV. ► Neutron cross-section measurements by means of the activation technique. ► Reference reactions 93 Nb(n,2n) 92m Nb and 27 (n,α) 24 Na. ► Data for 172 Yb(n,p) 172 Tm and 176 Yb(n,d * ) 175 Tm are reported for the first time. - Abstract: Measurements of (n,2n), (n,p), and (n,d * ) (The expression (n,d * ) cross section used in this work includes a sum of (n,d), (n,np) and (n,pn) cross sections.) reaction cross-sections on ytterbium isotopes have been carried out in the range of 13.5–14.8 MeV using the activation technique. The monoenergetic neutron beams were produced via the 3 H(d,n) 3 He reaction. The neutron energies of different directions were determined using the Nb/Zr method. Samples were activated along with along with Nb and Al monitor foils to determine the incident neutron flux. Data are reported for the following reactions: 168 Yb(n,2n) 167 Yb, 170 Yb(n,2n) 169m+g Yb, 176 Yb(n,2n) 175m+g Yb, 172 Yb(n,p) 172 Tm, 173 Yb(n,p) 173 Tm, 176 Yb(n,d * ) 175 Tm, 174 Yb(n,p) 174 Tm, and 176 Yb(n,p) 176 Tm. The experimentally deduced cross-sections are compared with the existing experimental data. Furthermore, theoretical statistical model, based on the Hauser–Feshbach formalism, have been carried out using the HFTT
Neutron total cross section measurements on 249Cf
International Nuclear Information System (INIS)
Carlton, R.F.; Harvey, J.A.; Hill, N.W.; Pandey, M.S.; Benjamin, R.W.
1979-01-01
Neutron total cross section measurements were performed on a sample of 249 Cf (5.65 mg total weight) with the ORELA as a source of pulsed neutrons. The sample, the inverse thickness of which was 1542 barns/atom, consisted of 85.3% 249 Cf and 14.4% 249 Bk, and was cooled to liquid nitrogen temperature. Analyses were also made of data from a thin sample (l/n = 17430) of 65% 249 Cf in the region of the large fission resonance at 0.7 eV. Fifty-five resonances in 249 Cf were observed and analyzed over the energy range 0.1 eV to 90 eV by use of an R-matrix multilevel formalism. The resonance parameters obtained were used to determine the level spacing and the s-wave neutron and fission strength functions. Thermal total cross section measurements were also performed. 5 figures, 3 tables
Neutron cross section measurement using the Oak Ridge Electron Linear Accelerator
International Nuclear Information System (INIS)
Winters, R.R.
1991-08-01
This report discusses: argon-40 -- neutron reaction total cross sections from 6.9 kev to 50 kev; The maxwellian averaged neutron capture cross section of oxygen-16; r-matrix parameter analysis of the lead-208 -- neutron reaction cross section measurement; r-matrix parameter analysis of the ORELA neutron transmission zirconium-90 low energy measurement; porting computer codes from the HP9000 to the IBM RISC/6000;and measurements of neutron reactions with strontium-88, zirconium-90, and calcium-40
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).
International Nuclear Information System (INIS)
Erradi, L.; Karouani, K.
1994-01-01
Many multigroup neutron cross section libraries have been processed from basic evaluated nuclear data for use in neutron dosimetry, reactor shielding calculation and in the development of fusion reactors. Most of these libraries have been tested only for fission spectra and were not validated for fusion spectra. Fifteen of these libraries such as DOSCROS84, IRDF85 and ENDFB5 have been used along with the neutron spectra unfolding code SAND II to evaluate about fifteen threshold detector saturated activities. The comparison between these computed activities and the measured ones of a set of foils placed in different places along the axis of a paraffin cylinder and irradiated by 14 MeV neutrons generated by a D-T source, hence giving rise to complex spectra, leads to different types of discrepancies. The analysis of these discrepancies allows to select from these libraries the ones that can be recommended. 1 fig., 4 refs. (author)
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)
International Nuclear Information System (INIS)
Genreith, Christoph
2015-01-01
Nuclear waste needs to be characterized for its safe handling and storage. In particular long-lived actinides render the waste characterization challenging. The results described in this thesis demonstrate that Prompt Gamma Neutron Activation Analysis (PGAA) with cold neutrons is a reliable tool for the non-destructive analysis of actinides. Nuclear data required for an accurate identification and quantification of actinides was acquired. Therefore, a sample design suitable for accurate and precise measurements of prompt γ-ray energies and partial cross sections of long-lived actinides at existing PGAA facilities was presented. Using the developed sample design the fundamental prompt γ-ray data on 237 Np, 241 Am and 242 Pu were measured. The data were validated by repetitive analysis of different samples at two individual irradiation and counting facilities - the BRR in Budapest and the FRM II in Garching near Munich. Employing cold neutrons, resonance neutron capture by low energetic resonances was avoided during the experiments. This is an improvement over older neutron activation based works at thermal reactor neutron energies. 152 prompt γ-rays of 237 Np were identified, as well as 19 of 241 Am, and 127 prompt γ-rays of 242 Pu. In all cases, both high and lower energetic prompt γ-rays were identified. The most intense line of 237 Np was observed at an energy of E γ =182.82(10) keV associated with a partial capture cross section of σ γ =22.06(39) b. The most intense prompt γ-ray lines of 241 Am and of 242 Pu were observed at E γ =154.72(7) keV with σ γ =72.80(252) b and E γ =287.69(8) keV with σ γ =7.07(12) b, respectively. The measurements described in this thesis provide the first reported quantifications on partial radiative capture cross sections for 237 Np, 241 Am and 242 Pu measured simultaneously over the large energy range from 45 keV to 12 MeV. Detailed uncertainty assessments were performed and the validity of the given uncertainties was
AFCI-2.0 Neutron Cross Section Covariance Library
Energy Technology Data Exchange (ETDEWEB)
Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.
2011-03-01
The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural
AFCI-2.0 Neutron Cross Section Covariance Library
International Nuclear Information System (INIS)
Herman, M.; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.
2011-01-01
The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R and D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78
Neutron, Proton, and Photonuclear Cross Sections for Radiation Therapy and Radiation Protection
International Nuclear Information System (INIS)
Chadwick, M.B.
1998-01-01
The authors review recent work at Los Alamos to evaluate neutron, proton, and photonuclear cross section up to 150 MeV (to 250 MeV for protons), based on experimental data and nuclear model calculations. These data are represented in the ENDF format and can be used in computer codes to simulate radiation transport. They permit calculations of absorbed dose in the body from therapy beams, and through use of kerma coefficients allow absorbed dose to be estimated for a given neutron energy distribution. For radiation protection, these data can be used to determine shielding requirements in accelerator environments, and to calculate neutron, proton, gamma-ray, and radionuclide production. Illustrative comparisons of the evaluated cross section and kerma coefficient data with measurements are given
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)
Inelastic neutron scattering cross-section measurements on 7Li and 63,65Cu
Nyman, Markus; Belloni, Francesca; Ichinkhorloo, Dagvadorj; Pirovano, Elisa; Plompen, Arjan; Rouki, Chariklia
2017-09-01
The γ-ray production cross section for the 477.6-keV transition in 7Li following inelastic neutron scattering has been measured from the reaction threshold up to 18 MeV. This cross section is interesting as a possible standard for other inelastic scattering measurements. The experiment was conducted at the Geel Electron LINear Accelerator (GELINA) pulsed white neutron source with the Gamma Array for Inelastic Neutron Scattering (GAINS) spectrometer. Previous measurements of this cross section are reviewed and compared with our results. Recently, this cross section has also been calculated using the continuum discretized coupled-channels (CDCC) method. Experiments for studying neutrinoless double-β decay (2β0ν) or other very rare processes require greatly reducing the background radiation level (both intrinsic and external). Copper is a common shielding and structural material, used extensively in experiments such as COBRA, CUORE, EXO, GERDA, and MAJORANA. Understanding the background contribution arising from neutron interactions in Cu is important when searching for very weak experimental signals. Neutron inelastic scattering on natCu was investigated with GAINS. The results are compared with previous experimental data and evaluated nuclear data libraries.
Inelastic neutron scattering cross-section measurements on 7Li and 63,65Cu
Directory of Open Access Journals (Sweden)
Nyman Markus
2017-01-01
Full Text Available The γ-ray production cross section for the 477.6-keV transition in 7Li following inelastic neutron scattering has been measured from the reaction threshold up to 18 MeV. This cross section is interesting as a possible standard for other inelastic scattering measurements. The experiment was conducted at the Geel Electron LINear Accelerator (GELINA pulsed white neutron source with the Gamma Array for Inelastic Neutron Scattering (GAINS spectrometer. Previous measurements of this cross section are reviewed and compared with our results. Recently, this cross section has also been calculated using the continuum discretized coupled-channels (CDCC method. Experiments for studying neutrinoless double-β decay (2β0ν or other very rare processes require greatly reducing the background radiation level (both intrinsic and external. Copper is a common shielding and structural material, used extensively in experiments such as COBRA, CUORE, EXO, GERDA, and MAJORANA. Understanding the background contribution arising from neutron interactions in Cu is important when searching for very weak experimental signals. Neutron inelastic scattering on natCu was investigated with GAINS. The results are compared with previous experimental data and evaluated nuclear data libraries.
Neutron Cross Section Libraries for Cryogenic Aromatic Moderator Materials
International Nuclear Information System (INIS)
Cantargi, Florencia; Granada, J.R.; Sbaffoni, Maria Monica
2008-01-01
The dynamics of a set of aromatic hydrocarbons, such as benzene, toluene, mesitylene and a 3:2 mixture (by volume) of mesitylene and toluene, all of them in solid phase, was studied as potential moderator materials for cold neutron sources. Cross section libraries were generated for hydrogen bounded in those materials, at several temperatures in ACE format, and they were used in MCNP calculations to analyze their neutron production compared with traditional materials like solid methane and liquid hydrogen. In particular, cross section libraries were generated at 20 K, which is the operating temperature of the majority of the existing cold neutron sources. Although solid methane is the best moderator in terms of cold neutron production, it has very poor radiation resistance, causing spontaneous burping even at fairly low doses. Such effect is considerably reduced in the aromatic hydrocarbons. On the other hand, all of them show a similar and significant neutron production, with the exception of benzene. Even though those aromatic materials are very easy to handle, the solid phases that produce an enhanced flux of cold neutrons correspond to amorphous structures rich in low-energy excitations, and they can be created through lengthy cooling processes requiring in many cases additional annealing stages. The 3:2 mesitylene-toluene mixture, that forms in a simple and direct manner the appropriate disordered structure, constitutes an excellent cryogenic moderator material, as it is able to produce an intense flux of cold neutrons while presenting high resistance to radiation, thus conforming a new and advantageous alternative to traditional moderator materials. (authors)
International Nuclear Information System (INIS)
Baba, Mamoru; Itoh, Nobuo; Maeda, Kazuto; Hirakawa, Naohiro; Wakabayashi, Hidetaka.
1989-10-01
This report presents the summary of experimental studies of prompt fission neutron spectra and double-differential neutron inelastic-scattering cross sections of 238 U and 232 Th. The experiments were performed at Tohoku University Fast Neutron Laboratory employing a time-of-flight technique and Dynamitron accelerator as the pulsed neutron generator. From the experiments, we obtained the following data for both nuclei; 1. prompt fission neutron spectrum for 2 MeV neutrons, 2. double-differential neutron inelastic-scattering cross sections for 1.2, 2.0, 4.2, 6.1 and 14.1 MeV incident neutrons. Both in experiments and data processing, cares were taken to obtain reliable data by avoiding systematic uncertainty. The experimental data were compared with those by other experiments, evaluations and model calculations. Through the data comparison, some fundamental problems were found in the experiments by previous authors and the evaluations. The present data will provide useful data base for refinement of the evaluated data and theoretical models. (author)
International Nuclear Information System (INIS)
Tran Tuan Anh; Dang Lanh; Nguyen Canh Hai; Nguyen Xuan Hai; Pham Kien; Nguyen Thuy Nham; Pham Ngoc Son; Ho Huu Thang
2007-01-01
Determination of total neutron cross sections and average resonance parameters in the energy range from tens keV to hundreds keV is important for fast reactors calculations and designs because this energy range gives the most output of all neutron induced reactions in the spectrum of fast reactors. Besides, the total neutron cross section measurement is also one of the methods for determination of s, p and d-wave neutron strength functions. The purpose of this project is to develop a method for measurement of total neutron cross sections based on the neutron transmission technique using a He-3 counter. The average total neutron cross sections of 238 U were obtained from neutron transmission measurements on filtered neutron beams of 55 keV and 144 keV at the horizontal channel No.4 of the Dalat research reactor. The present results have been compared with the previous measurements, and the evaluated data from ENDF/B-6.8 library. (author)
Summary report of technical meeting on neutron cross section covariances
International Nuclear Information System (INIS)
Trkov, A.; Smith, D.L.; Capote Noy, R.
2011-01-01
A summary is given of the Technical Meeting on Neutron Cross Section Covariances. The meeting goal was to assess covariance data needs and recommend appropriate methodologies to address those needs. Discussions on covariance data focused on three general topics: 1) Resonance and unresolved resonance regions; 2) Fast neutron region; and 3) Users' perspective: benchmarks' uncertainty and reactor dosimetry. A number of recommendations for further work were generated and the important work that remains to be done in the field of covariances was identified. (author)
Absolute measurements of neutron cross sections. Progress report
International Nuclear Information System (INIS)
1984-11-01
In the photoneutron laboratory, we have completed a major refurbishing of experimental facilities and begun work on measurements of the capture cross section in thorium and U-238. In the 14 MeV neutron experimental bay, work continues on the measurement of 14 MeV neutron induced reactions of interest as standards or because of their technological importance. First results have been obtained over the past year, and we are extending these measurements along the lines outlined in our proposal of a year ago
Kitis, G; Wiescher, M; Dahlfors, M; Soares, J
2002-01-01
We propose to measure the neutron capture cross sections of $^{139}$La, of $^{93}$Zr (t$_{1/2}$)=1.5 10$^{6}$ yr), and of all the stable Zr isotopes at n_TOF. The aim of these measurements is to improve the accuracy of existing results by at least a factor of three in order to meet the quality required for using the s-process nucleosynthesis as a diagnostic tool for neutron exposure and neutron flux during the He burning stages of stellar evolution. Combining these results with a wealth of recent information coming from high-resolution stellar spectroscopy and from the detailed analysis of presolar dust grains will shed new light on the chemical history of the universe. The investigated cross sections are also needed for technological applications, in particular since $^{93}$Zr is one of the major long-lived fission products.
A compact fast-neutron producing target for high resolution cross section measurements
Flaska, M.
2006-01-01
A proper knowledge of neutron cross sections is very important for the operation safety of various nuclear facilities. Reducing uncertainties in the neutron cross sections can lead to an enhanced safety of present and future nuclear power systems. Accurate neutron cross sections also play a relevant
Thermal neutron scattering cross sections of beryllium and magnesium oxides
International Nuclear Information System (INIS)
Al-Qasir, Iyad; Jisrawi, Najeh; Gillette, Victor; Qteish, Abdallah
2016-01-01
Highlights: • Neutron thermalization in BeO and MgO was studied using Ab initio lattice dynamics. • The BeO phonon density of states used to generate the current ENDF library has issues. • The BeO cross sections can provide a more accurate ENDF library than the current one. • For MgO an ENDF library is lacking: a new accurate one can be built from our results. • BeO is a better filter than MgO, especially when cooled down to 77 K. - Abstract: Alkaline-earth beryllium and magnesium oxides are fundamental materials in nuclear industry and thermal neutron scattering applications. The calculation of the thermal neutron scattering cross sections requires a detailed knowledge of the lattice dynamics of the scattering medium. The vibrational properties of BeO and MgO are studied using first-principles calculations within the frame work of the density functional perturbation theory. Excellent agreement between the calculated phonon dispersion relations and the experimental data have been obtained. The phonon densities of states are utilized to calculate the scattering laws using the incoherent approximation. For BeO, there are concerns about the accuracy of the phonon density of states used to generate the current ENDF/B-VII.1 libraries. These concerns are identified, and their influences on the scattering law and inelastic scattering cross section are analyzed. For MgO, no up to date thermal neutron scattering cross section ENDF library is available, and our results represent a potential one for use in different applications. Moreover, the BeO and MgO efficiencies as neutron filters at different temperatures are investigated. BeO is found to be a better filter than MgO, especially when cooled down, and cooling MgO below 77 K does not significantly improve the filter’s efficiency.
Cross section measurement for (n,n{alpha}) reactions by 14 MeV neutrons
Energy Technology Data Exchange (ETDEWEB)
Kasugai, Y.; Ikeda, Y.; Uno, Y. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Yamamoto, H.; Kawade, K.
1997-03-01
Nine (n,n{alpha}) cross sections for (n,n{alpha}) reactions induced by 13.5-14.9 MeV neutrons were measured for {sup 51}V, {sup 65}Cu, {sup 71}Ga, {sup 76}Ge, {sup 87}Rb, {sup 91}Zr, {sup 93}Nb, {sup 96}Zr and {sup 109}Ag isotopes by using Fusion Neutronics Source (FNS) at JAERI. The reactions for 91Zr and 96Zr were measured for the first time. The evaluated data of JENDL-3 and ENDF/B-VI were compared with the present data. Some of the evaluated values are much different from our data by a factor more than ten. (author)
Experimental evaluation of the primary damage process: neutron energy effects
Energy Technology Data Exchange (ETDEWEB)
Goland, A.N.
1979-01-01
Experimental evaluation of the neutron energy dependnece of the primary damage stage depends upon a number of theoretical concepts. This state can only be observed after low- or perhaps ambient-temperature, low-fluence irradiations. The primary recoil energy spectrum, which determines the character of the displacement cascades, can be calculated if dosimetry has provided an accurate neutron spectrum. A review of experimental results relating neutron-energy effects shows that damage energy or damage energy cross section has often been a reliable correlation parameter for primary damage state experiments. However, the forthcoming emphasis on higher irradiation temperatures, more complex alloys and microstructural evolution has fostered a search for additional meaningful correlation parameters.
Evaluated cross section libraries
International Nuclear Information System (INIS)
Maqurno, B.A.
1976-01-01
The dosimetry tape (ENDF/B-IV tape 412) was issued in a general CSEWG distribution, August 1974. The pointwise cross section data file was tested with specified reference spectra. A group averaged cross section data file (620 groups based on tape 412) was tested with the above spectra and the results are presented in this report
Resonance structure of 32S+n from measurements of neutron total and capture cross sections
International Nuclear Information System (INIS)
Halperin, J.; Johnson, C.H.; Winters, R.R.; Macklin, R.L.
1980-01-01
Neutron total and capture cross sections of 32 S have been measured up to 1100 keV neutron energy [E/sub exc/( 33 S) =9700 keV]. Spin and parity assignments have been made for 28 of the 64 resonances found in this region. Values of total radiation widths, reduced neutron widths, level spacings, and neutron strength functions have been evaluated for s/sub 1/2/, p/sub 1/2/, p/sub 3/2/, and d/sub 5/2/ levels. Single particle contributions using the valency model account for a significant portion of the total radiation width only for the p/sub 1/2/-wave resonances. A significant number of resonances can be identified with reported levels excited in 32 S(d,p) and 29 Si(α,n) reactions. A calculation of the Maxwellian average cross section appropriate to stellar interiors indicates an average capture cross section at 30 keV, sigma-bar approx. = 4.2(2) mb, a result that is relatively insensitive to the assumed stellar temperature. Direct (potential) capture and the s-wave resonance capture contributions to the thermal capture cross section do not fully account for the reported thermal cross section (530 +- 40 mb) and a bound state is invoked to account for the discrepancy
Benchmark of neutron production cross sections with Monte Carlo codes
Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun
2018-02-01
Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first
International Nuclear Information System (INIS)
Konobeyev, A.Yu.; Fischer, U.; Simakov, S.P.
2016-01-01
The goal of this work is the evaluation of uncertainties of calculated atomic displacement cross sections for iron and tungsten irradiated with neutrons. Uncertainties were analysed for neutron incident energies above 0.1 MeV, which make the main contribution to the value of radiation damage rate for different types of nuclear or fusion reactors and neutron sources
Tables of RCN-2 fission-product cross section evaluation
International Nuclear Information System (INIS)
Gruppelaar, H.
1979-05-01
This report (continuation of ECN-13 and ECN-33) describes the third part of the RCN-2 evaluation of neutron cross sections for fission product nuclides in KEDAK format. It contains evaluated data for nine nuclides, i.e. 142 Nd, 143 Nd, 144 Nd, 145 Nd, 146 Nd, 147 Nd, 148 Nd, 150 Nd and 147 Pm. Most emphasis has been given to the evaluation of the radiative capture cross section, in order to provide a data base for adjustment calculations using results of integral measurements. Short evaluation reports are given for this cross section. The evaluated capture cross sections are compared with recent experimental differential and integral data. Graphs are given of the capture cross sections at neutron energies above 1 keV, in which also adjusted point cross sections, based upon integral STEK and CFRMF data have been plotted. Moreover, the results are compared with those of the well-known ENDF/B-IV evaluation for fission product nucleides. Finally, evaluation summaries are given, which include tables of other important neutron cross sections, such as the total, elastic scattering and inelastic scattering cross sections
International Nuclear Information System (INIS)
Cullen, D.E.
1977-01-01
A code, SIGMA1, has been designed to Doppler broaden evaluated cross sections in the ENDF/B format. The code can only be applied to tabulated data that vary linearly in energy and cross section between tabulated points. This report describes the methods used in the code and serves as a user's guide to the code
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.
Fission neutron spectrum averaged cross sections for threshold reactions on arsenic
International Nuclear Information System (INIS)
Dorval, E.L.; Arribere, M.A.; Kestelman, A.J.; Comision Nacional de Energia Atomica, Cuyo Nacional Univ., Bariloche; Ribeiro Guevara, S.; Cohen, I.M.; Ohaco, R.A.; Segovia, M.S.; Yunes, A.N.; Arrondo, M.; Comision Nacional de Energia Atomica, Buenos Aires
2006-01-01
We have measured the cross sections, averaged over a 235 U fission neutron spectrum, for the two high threshold reactions: 75 As(n,p) 75 mGe and 75 As(n,2n) 74 As. The measured averaged cross sections are 0.292±0.022 mb, referred to the 3.95±0.20 mb standard for the 27 Al(n,p) 27 Mg averaged cross section, and 0.371±0.032 mb referred to the 111±3 mb standard for the 58 Ni(n,p) 58m+g Co averaged cross section, respectively. The measured averaged cross sections were also evaluated semi-empirically by numerically integrating experimental differential cross section data extracted for both reactions from the current literature. The calculations were performed for four different representations of the thermal-neutron-induced 235 U fission neutron spectrum. The calculated cross sections, though depending on analytical representation of the flux, agree with the measured values within the estimated uncertainties. (author)
Cross sections for D-T neutron interaction with neodymium isotopes
International Nuclear Information System (INIS)
Luo, Junhua; An, Li; Jiang, Li; He, Long
2015-01-01
The cross-sections for (n, x) reactions with neodymium isotopes were measured at (D-T) neutron energies around 14 MeV with the activation technique. Samples were activated along with Nb and Al monitor foils to determine the incident neutron flux. Data are reported for the following reactions: 142 Nd(n,2n) 141 Nd, 148 Nd(n,2n) 147 Nd, 150 Nd(n,2n) 149 Nd, 142 Nd(n,p) 142 Pr, 146 Nd(n,α) 143 Ce, and 146 Nd(n,p) 146 Pr. Theoretical calculations of excitation functions were performed with the TALYS-1.6 nuclear model code, at neutron energies varying from the reaction threshold to 20 MeV. The results were discussed and compared with experimental data found in the literature, and with the comprehensive evaluation data in ENDF/B-VII.1, JENDL-4.0, and CENDL-3 libraries. - Highlights: • The cross sections for the (n,x) reactions on Neodymium have been measured. • Mono-energetic neutron beams using the D-T reaction; Energies: 13.5–14.8 MeV. • Neutron cross-section measurements by means of the activation technique. • Reference reactions 93 Nb(n,2n) 92m Nb and 27 (n, α) 24 Na were used as the monitor. • Nuclear reaction code TALYS-1.6 was used
Evaluation of neutron nuclear data for 243Am
International Nuclear Information System (INIS)
Igarasi, Sin-iti; Nakagawa, Tsuneo
1977-06-01
Evaluation of neutron nuclear data for 243 Am was performed below 16 MeV. The energy region above 250 eV was separated from the lower region where the resonance parameters were given. Evaluation was made to select suitable resonance parameters, and thermal values of the capture and fission cross sections were obtained with the adopted resonance parameters. An average fission width was assumed to bridge the cross sections at 0.0253 eV and above 250 eV. Using a semi-empirical formula, the fission cross section was reproduced above 250 eV. Optical and statistical model calculations were made in order to obtain the total, capture, inelastic and elastic scattering, and (n,2n) reaction cross sections. (auth.)
Method of measuring neutron spectra in JMTR exclusively used for irradiation and their evaluation
International Nuclear Information System (INIS)
Sakurai, Kiyoshi
1983-01-01
In the core of the Japan Materials Testing Reactor, about 60 capsules are irradiated. These are the material capsules for irradiating reactor materials, the fuel capsules for irradiating reactor fuel, the RI capsules for producing radioisotopes and so on. In the irradiation experiment using a reactor, the information on the neutron fluence is indispensable, and the neutron fluence in the irradiated specimen part is evaluated with a dosimeter or the nuclear calculation for the core of the JMTR. At the time of irradiating reactor materials, the dosimeter Fe-54 (n,p) Mn-54 is generally used for evaluating the neutron fluence more than 1 MeV. In the case of fuel irradiation, the thermal neutron fluence is evaluated with the dosimeter Co-59 (n,γ) Co-60. It is important to examine in detail neutron spectra by both calculation and experiment in the reactors exclusively used for irradiation such as the JMTR. The neutron irradiation field in the JMTR, neutron spectrum measuring experiment, the neutron flux monitors for standardizing data, the measurement of X-ray and gamma ray, neutron guess spectrum, the compilation of neutron cross section for SAND 2, and the unfolding of neutron spectra are reported. The degree of agreement of the neutron fluence more than 1 MeV by measurement and calculation was +- 10 to 20 %. (Kako, I.)
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%
Fast-neutron scattering cross sections of elemental silver
International Nuclear Information System (INIS)
Smith, A.B.; Guenther, P.T.
1982-05-01
Differential neutron elastic- and inelastic-scattering cross sections of elemental silver are measured from 1.5 to 4.0 MeV at intervals of less than or equal to 200 keV and at 10 to 20 scattering angles distributed between 20 and 160 0 . Inelastically-scattered neutron groups are observed corresponding to the excitation of levels at; 328 +- 13, 419 +- 50, 748 +- 25, 908 +- 26, 1150 +- 38, 1286 +- 25, 1507 +- 20, 1623 +- 30, 1835 +- 20 and 1944 +- 26 keV. The experimental results are used to derive an optical-statistical model that provides a good description of the observed cross sections. The measured values are compared with corresponding quantities given in ENDF/B-V
Evaluation of neutron data for plutonium-242
International Nuclear Information System (INIS)
Maslov, V.M.; Porodzinkij, Yu.V.; Sukhovitskij, E.Sh.; Morogovskij, G.B.
1997-11-01
The evaluation of neutron data for 242 Pu is made in the energy region from 10 -5 eV up to 20 MeV. The results of the evaluation are compiled in the ENDF/B-VI format. This work is performed under the Project Agreement CIS-03-95 with the International Science and Technology Center (Moscow). The Financing Party for the Project is Japan. The evaluation was requested by Japan Nuclear Data Center (JAERI). (author)
Evaluation of neutron data for americium-241
Energy Technology Data Exchange (ETDEWEB)
Maslov, V.M.; Sukhovitskij, E.Sh.; Porodzinskij, Yu.V.; Klepatskij, A.B.; Morogovskij, G.B. [Radiation Physics and Chemistry Problems Inst., Minsk-Sosny (Belarus)
1997-03-01
The evaluation of neutron data for {sup 241}Am is made in the energy region from 10{sup -5} eV up to 20 MeV. The results of the evaluation are compiled in the ENDF/B-VI format. This work is performed under the Project Agreement CIS-03-95 with the International Science and Technology Center (Moscow). The Financing Party for the Project is Japan. The evaluation was requested by Y. Kikuchi (JAERI). (author). 60 refs.
International Nuclear Information System (INIS)
Meadows, J.W.; Smith, D.L.
1980-05-01
Experimental methods related to the production of neutrons for cross section studies at the Argonne Fast-Neutron Generator are reviewed. Target assemblies commonly employed in these measurements are described, and some of the relevant physical properties of the neutron source reactions are discussed. Various measurements have been performed to ascertain knowledge about these source reaction that is required for cross section data analysis purposes. Some results from these studies are presented, and a few specific examples of neutron-source-related corrections to cross section data are provided. 16 figures, 3 tables
Evaluation of Neutron Nuclear Data for Natural Titanium and Antimony
Institute of Scientific and Technical Information of China (English)
1989-01-01
<正> The evaluations of neutron data of natural titanium and antimony are performedin the energy range from 10-5 eV to 20 MeV.The following data are included:resonance parameters,cross sections of total,elastic seattering,nonelastic sea-ttering,inelastic seattering,（n,2n）,（n,3n）,and（n,X）（X=p,d,t,3He,α）reactions etc.The angular distributions and energy distributions of secondaryneutrons are also given.The reproduced cross sections from resolved and unresolved
BROND-2. USSR Evaluated Neutron Data Library
International Nuclear Information System (INIS)
Manokhin, V.N.
1992-01-01
BROND-2, the USSR computerized data library for evaluated neutron reaction data was released in 1992. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. The Library BROND-2 reported in this document supersedes the earlier versions ''BROND-NDS1'' plus three supplements ''BROND-NDS2'', ''BROND-NDS3'', ''BROND-NDS4''. 16 refs
BROND: USSR evaluated neutron data library
International Nuclear Information System (INIS)
Manokhin, V.N.
1991-03-01
BROND, the USSR computerized data library for evaluated neutron reaction data was released in 1987/1990. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. This document describes the contents of this library in the four supplements BROND-NDS1, BROND-NDS2, BROND-NDS3, BROND-NDS4. (author). 16 refs
BROND. USSR Evaluated Neutron Data Library
International Nuclear Information System (INIS)
Manokhin, V.N.
1990-09-01
BROND, the USSR computerized data library for evaluated neutron reaction data was released in 1987/1990. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. This document describes the contents of this library in the versions BROND-NDS1 and BROND-NDS2, BROND-NDS3, BROND-NDS4. (author). 16 refs
BROND: USSR Evaluated Neutron Data Library
International Nuclear Information System (INIS)
Manokhin, V.N.; Lemmel, H.D.; McLaughlin, P.K.
1989-10-01
BROND, the USSR computerized data library for evaluated neutron reaction data was released in 1987/1989. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. This document describes the contents of this library in the versions BROND-NDS1 and BROND-NDS2. (author). 16 refs
Fast-neutron scattering cross sections of elemental zirconium
International Nuclear Information System (INIS)
Smith, A.B.; Guenther, P.T.
1982-12-01
Differential neturon-elastic-scattering cross sections of elemental zirconium are measured from 1.5 to 4.0 MeV at intervals of less than or equal to 200 keV. Inelastic-neutron-scattering cross sections corresponding to the excitation of levels at observed energies of: 914 +- 25, 1476 +- 37, 1787 +- 23, 2101 +- 26, 2221 +- 17, 2363 +- 14, 2791 +- 15 and 3101 +- 25 keV are determined. The experimental results are interpreted in terms of the optical-statistical model and are compared with corresponding quantities given in ENDF/B-V
Directory of Open Access Journals (Sweden)
О. О. Gritzay
2016-12-01
Full Text Available Development of the technique for determination of the total neutron cross sections from the measurements of sample transmission by filtered neutrons, scattered on hydrogen is described. One of the methods of the transmission determination TH52Cr from the measurements of 52Cr sample, using average energy shift method for filtered neutron beam is presented. Using two methods of the experimental data processing, one of which is presented in this paper (another in [1], there is presented a set of transmissions, obtained for different samples and for different measurement angles. Two methods are fundamentally different; therefore, we can consider the obtained processing results, using these methods as independent. In future, obtained set of transmissions is planned to be used for determination of the parameters E0, Гn and R/ of the resonance 52Cr at the energy of 50 keV.
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
Contribution to the study of the unresolved resonance range of the neutrons cross sections
International Nuclear Information System (INIS)
Noguere, Gilles
2014-01-01
This document presents the statistical description of neutron cross sections in the unresolved resonance range. The modeling of the total cross section and of the 'shape - elastic' cross section is based on the 'average R-Matrix' formalism. The partial cross sections describing the radiative capture, elastic scattering, inelastic scattering and fission process are calculated using the Hauser-Feshbach formalism with width fluctuation corrections. In the unresolved resonance range, these models depend on the average resonance parameters (neutron strength function Sc, mean level spacing D c , average partial reaction widths Γ c , channel radius a c , effective radius R' and distant level parameter R-bar c ∞ ). The codes (NJOY, CALENDF...) dedicated to the processing of nuclear data libraries (JEFF, ENDF/B, JENDL, CENDL, BROND... ) use the average parameters to take into account the self-shielding phenomenon for the simulation of the neutron transport in Monte-Carlo (MCNP, TRIPOLI... ) and deterministic (APOLLO, ERANOS...) codes. The evaluation work consists in establishing a consistent set of average parameters as a function of the total angular momentum J of the system and of the orbital moment of the incident neutron l. The work presented in this paper aims to describe the links between the S-Matrix and the 'average R-Matrix' formalism for the calculation of Sc, R-bar c ∞ , ac and R'. (author) [fr
FENDL/A-2.0. Neutron activation cross section data library for fusion applications
International Nuclear Information System (INIS)
Pashchenko, A.B.; Wienke, H.; Kopecky, J.; Sublet, J.C. Sublet; Forrest, R.A.
1997-01-01
This document describes the contents of a comprehensive neutron cross section data library for 13,006 neutron activation reactions with 739 target nuclides from H (A=1,Z=1) to Cm (A=248,Z=96), in the incident energy range up to 20 MeV. FENDL/A-2 is a sublibrary of FENDL-2, the second revision of the evaluated nuclear data library for fusion applications. It is supplemented by a decay data library FENDL/D-2 in ENDF-6 format for 1867 nuclides. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape upon request. (author)
Integral test on activation cross section of tag gas nuclides using fast neutron spectrum fields
Energy Technology Data Exchange (ETDEWEB)
Aoyama, Takafumi; Suzuki, Soju [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1997-03-01
Activation cross sections of tag gas nuclides, which will be used for the failed fuel detection and location in FBR plants, were evaluated by the irradiation tests in the fast neutron spectrum fields in JOYO and YAYOI. The comparison of their measured radioactivities and the calculated values using the JENDL-3.2 cross section set showed that the C/E values ranged from 0.8 to 2.8 for the calibration tests in YAYOI and that the present accuracies of these cross sections were confirmed. (author)
Evaluation of the D(γ,n) reaction cross section
International Nuclear Information System (INIS)
Murata, T.
1994-01-01
Evaluation was performed for the cross section of photo-disintegration of Deuteron in the photon energy range between the threshold energy of the reaction (2.224 MeV) and pion production threshold (140 MeV). Angular distributions of the emitted neutrons were also evaluated. (author)
Neutron displacement damage cross sections for SiC
International Nuclear Information System (INIS)
Huang Hanchen; Ghoniem, N.
1993-01-01
Calculations of neutron displacement damage cross sections for SiC are presented. We use Biersack and Haggmark's empirical formula in constructing the electronic stopping power, which combines Lindhard's model at low PKA energies and Bethe-Bloch's model at high PKA energies. The electronic stopping power for polyatomic materials is computed on the basis of Bragg's Additivity Rule. A continuous form of the inverse power law potential is used for nuclear scattering. Coupled integro-differential equations for the number of displaced atoms j, caused by PKA i, are then derived. The procedure outlined above gives partial displacement cross sections, displacement cross sections for each specie of the lattice, and for each PKA type. The corresponding damage rates for several fusion and fission neutron spectra are calculated. The stoichiometry of the irradiated material is investigated by finding the ratio of displacements among various atomic species. The role of each specie in displacing atoms is also investigated by calculating the fraction of displacements caused by each PKA type. The study shows that neutron displacement damage rates of SiC in typical magnetic fusion reactor first walls will be ∝10-15 dpa MW -1 m 2 ; in typical lead-protected inertial confinement fusion reactor first walls they will be ∝15-20 dpa MW -1 m 2 . For fission spectra, we find that the neutron displacement damage rate of SiC is ∝74 dpa per 10 27 n/m 2 in FFTF, ∝39 dpa per 10 27 n/m 2 in HFIR, and 25 dpa per 10 27 n/m 2 in NRU. Approximately 80% of displacement atoms are shown to be of the carbon-type. (orig.)
Evaluation of the fusion-related neutron nuclear data for JENDL-3
International Nuclear Information System (INIS)
Chiba, Satoshi
1988-01-01
Status of the neutron nuclear data evaluations for JENDL-3 will be described for nuclides important in the development of D-T fusion reactors. In this article, however, only explanation of the evaluations for the very light mass region will be presented to avoid overlapping with what are given in another papers submitted to this seminar. Emphases are placed on the tritium production cross sections, inelastic scattering cross sections including the double-differential neutron emission spectrum (DDX), threshold reaction cross sections and photon production cross sections. The methods employed to prepare JENDL-3T library and their results will be summarized. (author)
Measurement of the neutron total cross section of sodium
International Nuclear Information System (INIS)
Larson, D.C.; Harvey, J.A.; Hill, M.W.
1976-01-01
The transmission of neutrons through a sample of pure sodium was measured in the energy range 40 keV to 20 MeV. The measurement points out several areas for improvement in the sodium evaluation for ENDF/B-V, the most important being the broadening of the minimum at 300 keV
BROND: USSR recommended evaluated neutron data library
International Nuclear Information System (INIS)
Manokhin, V.N.
1988-03-01
BROND is the recommended evaluated data library of the USSR for neutron induced nuclear reactions. It is a computer library recorded on magnetic tape presented in the internationally recommended format ENDF-5. It contains 65 files with recommended data for 65 elements or isotopes. For each file the present report gives a summary documentation on the contents, the evaluation methods and the originators of the files. (author). Refs
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
Measurements of neutron cross sections of radioactive waste nuclides
Energy Technology Data Exchange (ETDEWEB)
Katoh, Toshio [Gifu College of Medical Technology, Seki, Gifu (Japan); Harada, Hideo; Nakamura, Shoji; Tanase, Masakazu; Hatsukawa, Yuichi
1998-01-01
Accurate nuclear reaction cross sections of radioactive fission products and transuranic elements are required for research on nuclear transmutation methods in nuclear waste management. Important fission products in the nuclear waste management are {sup 137}Cs, {sup 135}Cs, {sup 90}Sr, {sup 99}Tc and {sup 129}I because of their large fission yields and long half-lives. The present authors have measured the neutron capture cross sections and resonance integrals of {sup 137}Cs, {sup 90}Sr and {sup 99}Tc. The purpose of this study is to measure the neutron capture cross sections and resonance integrals of nuclides, {sup 129}I and {sup 135}Cs accurately. Preliminary experiments were performed by using Rikkyo University Reactor and JRR-3 reactor at Japan Atomic Energy Research Institute (JAERI). Then, it was decided to measure the cross section and resonance integral of {sup 135}Cs by using the JRR-3 Reactor because this measurement required a high flux reactor. On the other hand, those of {sup 129}I were measured at the Rikkyo Reactor because the product nuclides, {sup 130}I and {sup 130m}I, have short half-lives and this reactor is suitable for the study of short lived nuclide. In this report, the measurements of the cross section and resonance integral of {sup 135}Cs are described. To obtain reliable values of the cross section and resonance integral of {sup 135}Cs(n, {gamma}){sup 136}Cs reaction, a quadrupole mass spectrometer was used for the mass analysis of nuclide in the sample. A progress report on the cross section of {sup 134}Cs, a neighbour of {sup 135}Cs, is included in this report. A report on {sup 129}I will be presented in the Report on the Joint-Use of Rikkyo University Reactor. (author)
View-CXS neutron and photon cross-sections viewer
International Nuclear Information System (INIS)
Subbaiah, K.V.; Sunil Sunny, C.
2004-01-01
A graphical user-friendly interface is developed in Visual Basic (VB)-6 to view the variation of neutron and photon interaction cross-sections of different isotopes as a function of energy. VB subroutines developed read the binary data files of cross-sections created in MCNP-ACE (Briesmeister, J.F., 1993. MCNP - a general purpose Monte Carlo N-Particle Transport code. Version 4A. LANL, USA), ANISN-DLC (Engle W.W. Jr., 1967, A User's Manual for ANISN, K-1693; ORNL, 1974. 100 group neutron cross section data based on ENDF/B-III. Oak Ridge National Laboratory, USA) and KENO-AMPX (Petrie, L.M., Landers, N.F., 1984 KENO-Va- An Improved Monte Carlo Criticality Program with Super Grouping. RSICC-CCC-548, USA) formats using LAHEY-77 Fortran Compiler. The information on isotopes present in each library will be displayed with the help of database files prepared using Micro-Soft ACESS. The cross-section data can be viewed in different presentation styles namely, line graphs, bar graphs, histograms etc., with different color and symbol options. The cross-section plots generated can be saved as Bit-Map file to embed in any other text files. This software enables inter comparison of cross-sections from different type of libraries for isotopes as well as mixtures. Provision is made to view the cross-sections for nuclear reactions such as (n,γ), (n,f), (n,α), etc. The software can be obtained from Radiation Safety Information and Computational Centre (RSICC), ORNL, USA with the code package identification number PSR-514. The software package needs a hard disk space of about 80 MB when installed and works in WINDOWS-95/98/2000 operating systems
International Nuclear Information System (INIS)
1986-01-01
For all stable and experimentally studied radionuclides evaluated data are presented on cross sections of thermal neutrons, on resonance integrals and medium neutron cross sections with energy of 30 KeV. Refs, figs and tabs
Joint evaluated file qualification for thermal neutron reactors
International Nuclear Information System (INIS)
Tellier, H.; Van der Gucht, C.; Vanuxeem, J.
1986-09-01
The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processed to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. We show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data. 8 refs
Joint evaluated file qualification for thermal neutron reactors
International Nuclear Information System (INIS)
Tellier, H.; van der Gucht, C.; Vanuxeem, J.
1986-01-01
The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processes to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. The authors show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the Joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data
Energy Technology Data Exchange (ETDEWEB)
Takao, Yoshiyuki; Kanda, Yukinori; Nagae, Koji; Fujimoto, Toshihiro [Kyushu Univ., Fukuoka (Japan); Ikeda, Yujiro
1997-03-01
Helium production cross sections of Iron were measured by helium accumulation method for neutron energies from 13.5 to 14.9 MeV. Iron samples were irradiated with FNS, an intense d-T neutron source of JAERI. As the neutron energy varies according to the emission angle at the neutron source, the samples were set around the neutron source and were irradiated by neutrons of different energy depending on each sample position. The amount of helium produced in a sample was measured by Helium Atoms Measurement System at Kyushu University. The results of this work are in good agreement with other experimental data in the literature and also compared with the evaluated values in JENDL-3. (author)
International intercomparison of neutron spectra evaluating methods using activation detectors
International Nuclear Information System (INIS)
Fischer, A.
1975-06-01
The international intercomparison of neutron spectrum evaluation methods using activation detectors was organized by the IAEA in 1971 - 1972. All of the contributions and the results of a critical evaluation are presented here. The spectra of different contributors are compared to a reference spectrum by means of different integrals and weighting functions. Different cross section sets, foil numbers, energy point systems, guess spectra used by the contributors cause differences in the resulting spectra. The possible ways of separating these effects are also investigated. Suggestions are made for the organization of a new intercomparison on the basis of more uniform input data. (orig.) [de
International Nuclear Information System (INIS)
Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.
1986-01-01
For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig
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
Fission-neutron displacement cross sections in metals
International Nuclear Information System (INIS)
Takamura, Saburo; Aruga, Takeo; Nakata, Kiyotomo
1985-01-01
The sensitivity damage rates for 22 metals were measured after fission-spectrum neutron irradiation at low temperature and the experimental damage rates were compared with the theoretical calculation. The relation between the theoretical displacement cross section and the atomic weight of metals can be written by two curves; one is for fcc and hcp metals, and another is for bcc metals. On the other hand, the experimental displacement cross section versus atomic weight is shown approximately by a curve for both fcc and bcc metals, and the cross section for hcp metals deviates from the curve. The defect production efficiency is 0.3-0.4 for fcc metals and 0.6-0.8 for bcc metals. (orig.)
Energy-averaged neutron cross sections of fast-reactor structural materials
International Nuclear Information System (INIS)
Smith, A.; McKnight, R.; Smith, D.
1978-02-01
The status of energy-averaged cross sections of fast-reactor structural materials is outlined with emphasis on U.S. data programs in the neutron-energy range 1-10 MeV. Areas of outstanding accomplishment and significant uncertainty are noted with recommendations for future efforts. Attention is primarily given to the main constituents of stainless steel (e.g., Fe, Ni, and Cr) and, secondarily, to alternate structural materials (e.g., V, Ti, Nb, Mo, Zr). Generally, the mass regions of interest are A approximately 50 to 60 and A approximately 90 to 100. Neutron total and elastic-scattering cross sections are discussed with the implication on the non-elastic-cross sections. Cross sections governing discrete-inelastic-neutron-energy transfers are examined in detail. Cross sections for the reactions (n;p), (n;n',p), (n;α), (n;n',α) and (n;2n') are reviewed in the context of fast-reactor performance and/or diagnostics. The primary orientation of the discussion is experimental with some additional attention to the applications of theory, the problems of evaluation and the data sensitivity of representative fast-reactor systems
Stellar Neutron Capture Cross Sections of the Lu and Hf Isotopes
International Nuclear Information System (INIS)
Wisshak, K.; Voss, F.; Kaeppeler, F.; Kazakov, L.; Krticka, M.
2005-01-01
The neutron capture cross sections of 175,176Lu and 176,177,178,179,180Hf have been measured in the energy range from 3 to 225 keV at the Karlsruhe 3.7 MV Van de Graaff accelerator relative to the gold standard. Neutrons were produced by the 7Li(p,n)7Be reaction and capture events were detected by the Karlsruhe 4πBaF2 detector. The cross section ratios could be determined with uncertainties between 0.9 and 1.8% about a factor of five more accurate than previous data. A strong population of isomeric states was found in neutron capture of the Hf isotopes, which are only partially explained by CASINO/GEANT simulations based on the known level schemes.Maxwellian averaged neutron capture cross sections were calculated for thermal energies between kT = 8 keV and 100 keV. Severe differences up to40% were found to the data of a recent evaluation based on existing experimental results. The new data allow for a much more reliable analysis of the important branching in the s-process synthesis path at 176Lu which can be interpreted as an s-process thermometer
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
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
Measurement of reaction cross sections of {sup 129}I induced by DT neutrons
Energy Technology Data Exchange (ETDEWEB)
Nakano, Daisuke; Murata, Isao; Takahashi, Akito [Osaka Univ., Suita (Japan). Faculty of Engineering
1997-03-01
The cross sections were measured for the {sup 129}I(n,2n){sup 128}I and {sup 129}I(n,{gamma}){sup 130}I reactions by DT neutrons, at OKTAVIAN facility of Osaka University, Japan. The foil activation method was used in the measurement. The sample was a sealed source of {sup 129}I, which was covered with a Cd foil. The irradiations were performed for 75 minutes to obtain the cross section of reaction producing {sup 128}I (T{sub 1/2}=24.99m) and 22 hours for the {sup 130}I (T{sub 1/2}=12.36h), respectively. The gamma-rays emitted from the irradiated sample were measured with a high purity Ge detector. The measured cross sections of {sup 129}I(n,2n){sup 128}I and {sup 129}I(n,{gamma}){sup 130}I reactions were 0.92{+-}0.11 barn and 0.013{+-}0.002 barn, respectively. For the {sup 129}I(n,2n){sup 128}I reaction, the evaluation of JENDL-3.2 overestimates cross section about 60% to the experimental result. However, especially for the {sup 129}I(n,{gamma}) reaction, the measured cross section may include the contribution from the neutrons in MeV region as well as epithermal ones. Also, the obtained cross section of the {sup 129}I(n,{gamma}){sup 130}I reaction was evaluated as an effective production cross section of {sup 130}I including {sup 129}I(n,{gamma}){sup 130m}I reaction. In order to remove the contribution from the epithermal and MeV region neutrons. A new method was proposed for the measurement of (n,{gamma}) reaction cross section. (author)
Neutron total cross section measurements of gold and tantalum at the nELBE photoneutron source
Hannaske, Roland; Beyer, Roland; Junghans, Arnd; Bemmerer, Daniel; Birgersson, Evert; Ferrari, Anna; Grosse, Eckart; Kempe, Mathias; Kögler, Toni; Marta, Michele; Massarczyk, Ralph; Matic, Andrija; Schramm, Georg; Schwengner, Ronald; Wagner, Andreas
2014-01-01
Neutron total cross sections of 197 Au and nat Ta have been measured at the nELBE photoneutron source in the energy range from 0.1 - 10 MeV with a statistical uncertainty of up to 2 % and a total systematic uncertainty of 1 %. This facility is optimized for the fast neutron energy range and combines an excellent t ime structure of the neutron pulses (electron bunch width 5 ps) with a short flight path of 7 m. Because of the low instantaneous neutron flux transmission measurements of neutron total cross sections are possible, that exhibit very different beam and back ground conditions than found at other neutron sources.
Measurements of neutron capture cross sections of wolfram and thulium
International Nuclear Information System (INIS)
Xia Yijun; Wang Chunhao; Yang Jingfu; Yang Zhihua; Luo Xiaobing
1992-01-01
The neutron capture cross sections of wolfram and thulium were measured in the energy range from 10 to 100 KeV using gold as the standard. Kinematically collimated neutrons were produced via the 7 Li(p, n) 7 Be reaction with a 2.5 MV pulsed Van de Graaff accelerator at Sichuan University. The capture events were detected by a pair of Moxon-Rae detectors. Time-of-flight technique was used to improve the signal-background ratio. The present results are compared with data by other authors. The capture cross section were calculated from 10 to 100 KeV for two nuclides by the Hauser-Feshbach statistical theory with width fluctuation correction. The nonstatistical effects such as potential capture and radiative capture in elastic and inelastic channels of a compound nucleus were included in the calculations. The calculated results show that the nonstatistical contribution to the capture cross sections is negligible compared with that of the statistical effects
Performing Neutron Cross-Section Measurements at RIA
International Nuclear Information System (INIS)
Ahle, L.E.
2003-01-01
The Rare Isotope Accelerator (RIA) is a proposed accelerator for the low energy nuclear physics community. Its goal is to understand the natural abundances of the elements heavier than iron, explore the nuclear force in systems far from stability, and study symmetry violation and fundamental physics in nuclei. To achieve these scientific goals, RIA promises to produce isotopes far from stability in sufficient quantities to allow experiments. It would also produce near stability isotopes at never before seen production rates, as much as 10 12 pps. Included in these isotopes are many that are important to stockpile stewardship, such as 87 Y, 146-50 Eu, and 231 Th. Given the expected production rates at RIA and a reasonably intense neutron source, one can expect to make ∼10 μg targets of nuclei with a half-life of ∼1 day. Thus, it will be possible at RIA to obtain experimental information on the neutron cross section for isotopes that have to date only been determined by theory. There are two methods to perform neutron cross-section measurements, prompt and delayed. The prompt method tries to measure each reaction as it happens. The exact technique employed will depend on the reaction of interest, (n,2n), (n,γ), (n,p), etc. The biggest challenge with this method is designing a detector system that can handle the gamma ray background from the target. The delayed method, which is the traditional radiochemistry method for determining the cross-section, irradiates the targets and then counts the reaction products after the fact. While this allows one to avoid the target background, the allowed fraction of target impurities is extremely low. This is especially true for the desired reaction product with the required impurity fraction on the order of 10 -9 . This is particularly problematic for (n,2n) and (n,γ) reactions, whose reaction production cannot be chemically separated from the target. In either case, the first step at RIA to doing these measurements is
An evaluated neutronic data file for bismuth
International Nuclear Information System (INIS)
Guenther, P.T.; Lawson, R.D.; Meadows, J.W.; Smith, A.B.; Smith, D.L.; Sugimoto, M.; Howerton, R.J.
1989-11-01
A comprehensive evaluated neutronic data file for bismuth, extending from 10 -5 eV to 20.0 MeV, is described. The experimental database, the application of the theoretical models, and the evaluation rationale are outlined. Attention is given to uncertainty specification, and comparisons are made with the prior ENDF/B-V evaluation. The corresponding numerical file, in ENDF/B-VI format, has been transmitted to the National Nuclear Data Center, Brookhaven National Laboratory. 106 refs., 10 figs., 6 tabs
Neutron nuclear data evaluation for actinide nucleic
International Nuclear Information System (INIS)
Chen Guochang; Yu Baosheng; Duan Junfeng; Ge Zhigang; Cao Wentian; Tang Guoyou; Shi Zhaomin; Zou Yubin
2010-01-01
The nuclear data with high accuracy for minor actinides are playing an important role in nuclear technology applications, including reactor design and operation, fuel cycle concepts, estimation of the amount of minor actinides in high burn-up reactors and the minor actinides transmutation. Through describe the class of nuclear data and nuclear date library, and introduce the procedure of neutron nuclear data evaluation. 234 U(n, f) and 237 Np(n, 2n) reaction experimental data evaluation was evaluated. The fission nuclear data are updated and improved. (authors)
Measurement of neutron captured cross-sections in 1-2 MeV
Energy Technology Data Exchange (ETDEWEB)
Kim, Gi Dong; Kim, Young Sek; Kim, Jun Kon; Yang, Tae Keun [Korea Institutes of Geoscience and Mineral Resources, Taejeon (Korea)
2001-04-01
The measurement of neutron captured reaction cross sections was performed to build the infra system for the production of nuclear data. MeV neutrons were produced with TiT target and {sup 3}T(p,n){sup 3}He reaction. The characteristics of TiT thin film was analyzed with ERD-TOF and RBS. The results was published at Journal of the Korea Physical Society (SCI registration). The energy, the energy spread and the flux of the produced neutron were measured. The neutron excitation functions of {sup 12}C and {sup 16}O were obtained to confirm the neutron energy and neutron energy spread. The neutron energy spread found to be 1.3 % at the neutron energy of 2.077 MeV. The {sup 197}Au(n,{gamma}) reaction was performed to obtain the nerutron flux. The maximum neutron flux found to be 1 x 10{sup 8} neutrons/sec at the neutron energy of 2 MeV. The absolute efficiency of liquid scintillation detector was obtained in the neutron energy of 1 - 2 MeV. The fast neutron total reaction cross sections of Cu, Fe, and Au were measured with sample in-out method. Also the neutron captured reaction cross sections of {sup 63}Cu were measured with fast neutron activation method. The measurement of neutron total reaction cross sections and the neutron captured reaction cross sections with fast neutrons were first tried in Korea. The beam pulsing system was investigated and the code of calculating the deposition spectrums for primary gamma rays was made to have little errors at nuclear data. 25 refs., 28 figs., 14 tabs. (Author)
Neutron-capture cross-section measurement for 163Dy In the neutron energy range from 15 to 75 keV
International Nuclear Information System (INIS)
Kim, Hyun Duk; Jung, Eui Jung; Ahn, Jung Keun; Lee, Dae Won; Kim, Guin Yun; Ro, Tae Ik; Min, Young Ki; Igashira, Masayuki; Ohsaki, Toshiro; Mizuno, Satoshi
2002-01-01
The neutron-capture cross-section of 163 Dy were measured in the neutron energy range from 15 to 75 keV at the 3-MV Pelletron accelerator of the Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology. Pulsed neutrons were produced from the 7 Li(p,n) 7 Be reaction by bombarding a metallic lithium target with the 1.903-MeV proton beam. The incident neutron spectra were measured by means of a neutron time-of-flight method with a 6 Li-glass detector. Capture γ-rays were detected with a large anti-Compton NaI(Tl) spectrometer. A pulse-height weighting technique was applied to the capture γ-ray pulse-height spectra to obtain capture yields. The neutron capture cross-section were determined relative to the standard capture cross-sections of 197 Au. The present results were compared with the previous measurements and the evaluated values of ENDF/B-VI
Neutron cross-section determination in geological samples (U)
International Nuclear Information System (INIS)
Harris, J.M.; McDaniel, P.J.
1982-01-01
The Prompt Gamma Neutron Activation Analysis (PGAA) technique yields elemental composition data which can be used to calculate the macroscopic cross section for any sample. The Small Sample Reactivity Measurements (SSRM) technique yields the macroscopic thermal absorption directly. Experimentally, PGAA is somewhat more difficult because of the calibration and data handling than is SSRM. However, SSRM requires a mathematical model of the reactor which means a rather complicated analysis. Once the model and calibration are completed, data analysis is routine. The SSRM technique is production oriented. 9 figures
Measurement of the uranium-235 fission cross section over the neutron energy range 1 to 6 MeV
International Nuclear Information System (INIS)
Barton, D.M.; Diven, B.C.; Hansen, G.E.; Jarvis, G.A.; Koontz, P.G.; Smith, R.K.
1976-01-01
The ratio of the fission cross section of 235 U to the scattering cross section of 1 H was measured in the 1- to 6-MeV range using monoenergetic neutrons from a pulsed 3 H(p,n) 3 He source. In this measurement, solid-state detectors determined fission fragment and recoil proton emissions from back-to-back U(99.7%) and polyethylene disks. Timing permitted discrimination against room-scattered neutron backgrounds. Absolute values for 235 U(n,f) are obtained using the Hopkins-Breit evaluation of the hydrogen-scattering cross section
Investigation of the 232Th neutron cross-sections in resonance energy range
International Nuclear Information System (INIS)
Grigoriev, Yu.V.; Kitaev, V.Ya.; Sinitsa, V.V.; Zhuravlev, B.V.; Borzakov, S.B.; Faikov-Stanchik, H.; Ilchev, G.L.; Panteleev, Ts.Ts.; Kim, G.N.
2001-01-01
The alternative path in the development of atomic energy is the uranium-thorium cycle. In connection with this, the measurements of the 232 Th neutron capture and total cross-sections and its resonance self-shielding coefficients in resonance energy range are necessary because of their low accuracy. In this work, the results of the investigations of the thorium-232 neutron cross-sections are presented. The measurements have been carried out on the gamma-ray multisection liquid detector and neutron detector as a battery of boron counters on the 120 m flight path of the pulsed fast reactor IBR-30. As the filter samples were used the metallic disks of various thickness and diameter of 45 mm. Two plates from metallic thorium with thickness of 0.2 mm and with the square of 4.5x4.5 cm 2 were used as the radiator samples. The group neutron total and capture cross-sections within the accuracy of 2-7% in the energy range of (10 eV-10 keV) were obtained from the transmissions and the sum spectra of g-rays from the fourth multiplicity to the seventh one. The neutron capture group cross-sections of 238 U were used as the standard for obtaining of thorium ones. Analogous values were calculated on the GRUCON code with the ENDF/B-6, JENDL-3 evaluated data libraries. Within the limits of experimental errors an agreement between the experiment and calculation is observed, but in some groups the experimental values are larger than the calculated ones. (author)
Stellar neutron capture cross sections of the Ba isotopes
International Nuclear Information System (INIS)
Voss, F.; Wisshak, K.; Guber, K.; Kaeppeler, F.; Reffo, G.
1994-03-01
The neutron capture cross sections of 134 Ba, 135 Ba, 136 Ba, and 137 Ba were measured in the energy range from 5 to 225 keV at the Karlsruhe 3.75 MV Van de Graaff accelerator. Neutrons were produced via the 7 Li(p,n) 7 Be reaction by bombarding metallic Li targets with a pulsed proton beam. Capture events were registered with the Karlsruhe 4π Barium Fluoride Detector. Several runs have been performed under different experimental conditions to study the systematic uncertainties, which resulted mainly from the large ratios of total to capture cross sections of up to 400. The cross section ratios were determined with an overall uncertainty of ∼3%, an improvement by factors of five to eight compared to existing data. Severe discrepancies were found with respect to previous results. Maxwellian averaged neutron capture cross sections were calculated for thermal energies between kT=10 keV and 100 keV. These stellar cross sections were used in an s-process analysis. For the s-only isotopes 134 Ba and 136 Ba the N s ratio was determined to 0.875±0.025. Hence, a significant branching of the s-process path at 134 Cs can be claimed for the first time, in contrast to predictions from the classical approach. This branching yields information on the s-process temperature, indicating values around T 8 =2. The new cross sections are also important for the interpretation of barium isotopic anomalies, which were recently discovered in SiC grains of carbonaceous chondrite meteorites. Together with the results from previous experiments on tellurium and samarium, a general improvement of the N s systematics in the mass range A=120 to 150 is achieved. This allows for a more reliable separation of s- and r-process yields, resulting in an improved assignment of the respective contributions to elemental barium that is required for comparison with stellar observations. (orig.) [de
A basic evaluated neutronic data file for elemental scandium
International Nuclear Information System (INIS)
Smith, A.B.; Meadows, J.W.; Howerton, R.J.
1992-01-01
This report documents an evaluated neutronic data file for elemental scandium, presented in the ENDF/B-VI format. This file should provide basic nuclear data essential for neutronic calculations involving elemental scandium. No equivalent file was previously available
International Nuclear Information System (INIS)
Mori, Takamasa
1985-05-01
In order to assess the cross section data for future reactor materials, such as molybdenum, niobium, titanium, lithium and fluorine, the angular neutron spectra in test piles of these materials or their chemical compounds have been measured in the energy range from a few keV to a few MeV by the linac time-of-flight method. The results have been compared with those theoretically calculated from the evaluated cross section data in such as JENDL-2 (or JENDL-1, JENDL-3PR1) and ENDF/B-IV. For both of molybdenum and niobium, it has been found that the energy distribution of inelastically scattered neutrons plays an important role in the analysis, and the JENDL library gives better predictions of spectrum shapes than ENDF/B-IV for both cases. In the case of niobium, however, it appears that the values of inelastic scattering cross section in JENDL-2 are too small around 2 MeV. It has been also found for niobium that the cross section data below 100 keV in ENDF/B-IV are inadequate. In a titanium pile, a discrepancy between the measured spectrum and the calculated one from ENDF/B-IV has been found in the energy range from about 60 keV to a few 100 keV. In order to investigate the cause of this discrepancy, the total cross sections for titanium have been measured by the transmission method. In the case of lithium, the discrepancy between the measured and calculated spectra is considerably reduced by adopting the angular distribution for 7 Li from ENDF/B-IV above about 500 keV. In the case of fluorine, spatial distributions of neutrons and X-rays have been also measured in both piles by the activation method to estimate the influence of photoneutrons generated in the sample material on the neutron distribution, and it has been found that their influence below 1 MeV is not so large as is necessary to be taken into account for the present assessment. (J.P.N)
Photo-neutron cross sections for unstable neutron-rich oxygen isotopes
International Nuclear Information System (INIS)
Leistenschneider, A.; Aumann, T.; Boretzky, K.
2001-05-01
The dipole response of stable and unstable neutron-rich oxygen nuclei of masses A = 17 to A = 22 has been investigated experimentally utilizing electromagnetic excitation in heavy-ion collisions at beam energies around 600 MeV/nucleon. A kinematically complete measurement of the neutron decay channel in inelastic scattering of the secondary beam projectiles from a Pb target was performed. Differential electromagnetic excitation cross sections dσ/dE were derived up to 30 MeV excitation energy. In contrast to stable nuclei, the deduced dipole strength distribution appears to be strongly fragmented and systematically exhibits a considerable fraction of low-lying strength, exhausting up to 12% of the energy-weighted dipole sum rule at excitation energies below 15 MeV. (orig.)
International Nuclear Information System (INIS)
Stewart, L.
1979-03-01
In response to an action by the Standards Subcommittee of the Cross Section Evaluation Working Group, a workshop was convened to determine the status of available information on prompt fission neutron spectra. The experimental data were reviewed and theoretical models were developed. The current ENDF/B fission neutron spectra files were summarized. Further work is currently under way, especially to provide a better theoretical tool to represent energy-dependent fission spectra. 5 references
Neutron cross-section library for SAND-2 and its service program
International Nuclear Information System (INIS)
Berzonis, M.A.; Bondars, Kh.Ya.; Lapenas, A.A.
1978-01-01
The logical structure of the neutron cross-section library used in the SAND-2 program complex is considered. The organization of the DSIG01 program creating and servicing the neutron cross section library is described. The DSIG 01 program is written on FORTRAN and permits to create the neutron cross section library on the ES computer magnetic discs operating under the control of the ES operating system and to perform certain manipulations therewith
Evaluation of cross sections of Th-232 and U-233
International Nuclear Information System (INIS)
Dias, A.M.
1978-01-01
The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt
Neutron cross sections for the interpretation of a spallation experiment
International Nuclear Information System (INIS)
Bortignon, P.F.; Mariani, F.; Perini, A.; Sangiust, V.
1994-01-01
An irradiation was carried out in a spallation neutron spectrum at the beam stop of LAMPF (Los Alamos); many activation detectors were used in order to obtain a fluence mapping inside the capsule volume. The cross sections were derived from ENDF B/V Dosimetry File up to 20 MeV and were based on calculations with the code ALICE up to 200 MeV. From 200 to 800 MeV an empirical extrapolation was employed since no data, calculated or measured, were available at the moment. An attempt has been made in this paper to revisit the interpretation of the experiment reconsidering the cross sections in the energy range from 200 to 800 MeV as given by the semiempirical models of Rudstam, Silberger and Tsao and Hufner
Measurement of secondary gamma-ray production cross sections of vanadium induced by D-T neutrons
International Nuclear Information System (INIS)
Kondo, Tetsuo; Murata, Isao; Takahashi, Akito
1999-01-01
The secondary gamma-ray production cross sections of vanadium induced by D-T neutrons have been measured. The experimental values were compared with the theoretical calculation results by SINCROS-II and the evaluation result based on experimental data compiled by Simakov. The calculation results supported our data, while Simakov's evaluation did not agree with the present result very well. (author)
International Nuclear Information System (INIS)
Chernukhin, Yu.G.; Kandiev, Ya.Z.; Lartsev, V.D.; Levakov, B.G.; Modestov, D.G.; Simonenko, V.A.; Streltsov, S.I.; Khmel'nitskij, D.V.
2006-01-01
The main stage of experiment for direct measurement of cross section of neutron-neutron scattering σ nn at low energies (E nn determination. It was shown, that for achieving the criterion ε ∼ 4% it will be necessary to have 40-50 pulses of a reactor [ru
Fast-neutron total and scattering cross sections of 103Rh
International Nuclear Information System (INIS)
Smith, A.B.; Guenther, P.T.; Whalen, J.F.
1982-07-01
Fast-neutron total cross sections of 103 Rh are measured with 30 to 50 keV resolutions from 0.7 to 4.5 MeV. Differential elastic- and inelastic-scattering cross sections are measured from 1.45 to 3.85 MeV. Scattered-neutron groups corresponding to excited levels at 334 +- 13, 536 +- 7, 648 +- 25, 796 +- 20, 864 +- 22, 1120 +- 22, 1279 +- 50, 1481 +- 27, 1683 +- 39, 1840 +- 79, 1991 +- 71 and 2050 (tentative) keV are observed. An optical-statistical model is derived from the elastic-scattering results. The experimental values are compared with comparable quantities given in the ENDF/B-V evaluation
Sensitivity Analysis of Nuclide Importance to One-Group Neutron Cross Sections
International Nuclear Information System (INIS)
Sekimoto, Hiroshi; Nemoto, Atsushi; Yoshimura, Yoshikane
2001-01-01
The importance of nuclides is useful when investigating nuclide characteristics in a given neutron spectrum. However, it is derived using one-group microscopic cross sections, which may contain large errors or uncertainties. The sensitivity coefficient shows the effect of these errors or uncertainties on the importance.The equations for calculating sensitivity coefficients of importance to one-group nuclear constants are derived using the perturbation method. Numerical values are also evaluated for some important cases for fast and thermal reactor systems.Many characteristics of the sensitivity coefficients are derived from the derived equations and numerical results. The matrix of sensitivity coefficients seems diagonally dominant. However, it is not always satisfied in a detailed structure. The detailed structure of the matrix and the characteristics of coefficients are given.By using the obtained sensitivity coefficients, some demonstration calculations have been performed. The effects of error and uncertainty of nuclear data and of the change of one-group cross-section input caused by fuel design changes through the neutron spectrum are investigated. These calculations show that the sensitivity coefficient is useful when evaluating error or uncertainty of nuclide importance caused by the cross-section data error or uncertainty and when checking effectiveness of fuel cell or core design change for improving neutron economy
Neutron data evaluation of 233U
International Nuclear Information System (INIS)
Maslov, V.M.; Tetereva, N.A.; Kagalenko, A.B.; Kornilov, N.V.; Baba, M.; Hasegawa, A.
2003-08-01
Consistent evaluation of 233 U measured data base is performed. Hauser-Feshbach- Moldauer theory, coupled channel model and double-humped fission barrier model are employed. Total, differential scattering, fission and (n,xn) data are calculated using fission cross section data description as a major constraint. The direct excitation of ground state is calculated within rigid rotator model. Average resonance parameters are provided, which reproduce evaluated cross sections in the range of 0.6-40.5 keV. (author)
Research on Fast-Doppler-Broadening of neutron cross sections
International Nuclear Information System (INIS)
Li, S.; Wang, K.; Yu, G.
2012-01-01
A Fast-Doppler-Broadening method is developed in this work to broaden Continuous Energy neutron cross-sections for Monte Carlo calculations. Gauss integration algorithm and parallel computing are implemented in this method, which is unprecedented in the history of cross section processing. Compared to the traditional code (NJOY, SIGMA1, etc.), the new Fast-Doppler-Broadening method shows a remarkable speedup with keeping accuracy. The purpose of using Gauss integration is to avoid complex derivation of traditional broadening formula and heavy load of computing complementary error function that slows down the Doppler broadening process. The OpenMP environment is utilized in parallel computing which can take full advantage of modern multi-processor computers. Combination of the two can reduce processing time of main actinides (such as 238 U, 235 U) to an order of magnitude of 1∼2 seconds. This new method is fast enough to be applied to Online Doppler broadening. It can be combined or coupled with Monte Carlo transport code to solve temperature dependent problems and neutronics-thermal hydraulics coupled scheme which is a big challenge for the conventional NJOY-MCNP system. Examples are shown to determine the efficiency and relative errors compared with the NJOY results. A Godiva Benchmark is also used in order to test the ACE libraries produced by the new method. (authors)
Nuclear theory for fast neutron nuclear data evaluation
International Nuclear Information System (INIS)
1988-11-01
The proceedings contain all invited and contributed papers presented at the Advisory Group Meeting on Nuclear Theory for Fast Neutron Data Evaluation held in Beijing 12-16 October 1987, as well as the conclusions and recommendations and the Chairman's summary of the meeting. The meeting presentations have been divided into six sessions devoted to the following topics: introductory speech (1 paper), optical potential (9 papers), compound nuclear theory (10 papers), pre-compound nuclear theory (13 papers), isomeric cross-section (1 paper) and intercomparison of nuclear model computer codes (1 paper). A separate abstract was prepared for each of these papers. Refs, figs and tabs
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)
International Nuclear Information System (INIS)
Georgiev, G.P.; Ermakov, V.A.; Grigor'ev, Yu.V.
1988-01-01
A multiplicity detector of the ''Romashka'' type has been used on the 500 m flight part of the IBR-30 pulsed reactor. The detector consists of 16 independent sections with NaJ(Tl) crystals with a total volume of 36 liters. The geometric efficiency of single-ray detection is ∼ 80%. The gamma-ray to neutron detection efficiency ratio is ≥600 for neutrons with energies below 200 keV. This detector allows one to perform neutron capture and fission cross section measurements and to study gamma-ray multiplicity and resonance selfabsorption effects in the 20 eV-200keV neutron energy range
Generation of neutron scattering cross sections for silicon dioxide
International Nuclear Information System (INIS)
Ramos, R; Marquez Damian, J.I; Granada, J.R.; Cantargi, F
2009-01-01
A set of neutron scattering cross sections for silicon and oxygen bound in silicon dioxide were generated and validated. The cross sections were generated in the ACE format for MCNP using the nuclear data processing system NJOY, and the validation was done with published experimental data. This cross section library was applied to the calculation of five critical configurations published in the benchmark Critical Experiments with Heterogeneous Compositions of Highly Enriched Uranium, Silicon Dioxide and Polyethylene. The original calculations did not use the thermal scattering libraries generated in this work and presented significant differences with the experimental results. For this reason, the newly generated library was added to the input and the multiplication factor for each configuration was recomputed. The utilization of the thermal scattering libraries did not result in an improvement of the computational results. Based on this we conclude that integral experiments to validate this type of thermal cross sections need to be designed with a higher influence of thermal scattering in the measured result, and the experiments have to be performed under more controlled conditions. [es
International Nuclear Information System (INIS)
Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.
1986-02-01
Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs
Abbondanno, U; Alvarez-Velarde, F; Alvarez-Pol, H; Andriamonje, Samuel A; Andrzejewski, J; Badurek, G; Baumann, P; Becvar, F; Benlliure, J; Berthoumieux, E; Calviño, F; Cano-Ott, D; Capote, R; Cennini, P; Chepel, V; Chiaveri, Enrico; Colonna, N; Cortés, G; Cortina-Gil, D; Couture, A; Cox, J; Dababneh, S; Dahlfors, M; David, S; Dolfini, R; Domingo-Pardo, C; Durán, I; Embid-Segura, M; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Frais-Kölbl, H; Furman, W; Gonçalves, I; Gallino, R; Gonzalez-Romero, E; Goverdovski, A; Gramegna, F; Griesmayer, E; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martínez, A; Isaev, S; Jericha, E; Kappeler, F; Kadi, Y; Karadimos, D; Kerveno, M; Ketlerov, V; Köhler, P; Konovalov, V; Krticka, M; Lamboudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marrone, S; Martinez-Val, J; Mastinu, P; Mengoni, A; Milazzo, P M; Molina-Coballes, A; Moreau, C; Mosconi, M; Neves, F; Oberhummer, Heinz; O'Brien, S; Pancin, J; Papaevangelou, T; Paradela, C; Pavlik, A; Pavlopoulos, P; Perlado, J M; Perrot, L; Pignatari, M; Plag, R; Plompen, A; Plukis, A; Poch, A; Policarpo, Armando; Pretel, C; Quesada, J; Raman, S; Rapp, W; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, Carlo; Rudolf, G; Rullhusen, P; Salgado, J; Soares, J C; Stéphan, C; Tagliente, G; Taín, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarín, D; Vincente, M C; Vlachoudis, V; Voss, F; Wendler, H; Wiescher, M; Wissha, K
2004-01-01
The measurement of **1**5**1Sm(n, gamma)**1**5**2Sm (samarium) cross section showed improved performance of the new spallation neutron facility. It covered a wide energy range with good resolution, high neutron flux, low backgrounds and a favourable duty factor. The samarium cross section was found to be of great importance for characterizing neutron capture nucleosynthesis in asymptotic giant stars. The combination of these features provided a promising basis for a broad experimental program directed towards application in astrophysics and advanced nuclear technologies. (Edited abstract)
Analytical methods for analysis of neutron cross sections of amino acids and proteins
International Nuclear Information System (INIS)
Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin; Carvalheira, Luciana; Rocha, Hélio F. da
2017-01-01
Two unpublished analytical processes were developed at IEN-CNEN-RJ for the analysis of neutron cross sections of chemical compounds and complex molecules, the method of data parceling and grouping (P and G) and the method of data equivalence and similarity (E and S) of cross-sections. The former allows the division of a complex compound or molecule so that the parts can be manipulated to construct a value of neutron cross section for the compound or the entire molecule. The second method allows by comparison obtain values of neutron cross-sections of specific parts of the compound or molecule, as the amino acid radicals or its parts. The processes were tested for the determination of neutron cross-sections of the 20 human amino acids and a small database was built for future use in the construction of neutron cross-sections of proteins and other components of the human being cells, also in other industrial applications. (author)
Analytical methods for analysis of neutron cross sections of amino acids and proteins
Energy Technology Data Exchange (ETDEWEB)
Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin; Carvalheira, Luciana, E-mail: dante@ien.gov.br, E-mail: fferreira@ien.gov.br, E-mail: Chaffin@ien.gov.br, E-mail: luciana@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Rocha, Hélio F. da, E-mail: helionutro@gmail.com.br [Universidade Federal do Rio de Janeiro (IPPMG/UFRJ), Rio de Janeiro, RJ (Brazil). Instituto de Pediatria
2017-07-01
Two unpublished analytical processes were developed at IEN-CNEN-RJ for the analysis of neutron cross sections of chemical compounds and complex molecules, the method of data parceling and grouping (P and G) and the method of data equivalence and similarity (E and S) of cross-sections. The former allows the division of a complex compound or molecule so that the parts can be manipulated to construct a value of neutron cross section for the compound or the entire molecule. The second method allows by comparison obtain values of neutron cross-sections of specific parts of the compound or molecule, as the amino acid radicals or its parts. The processes were tested for the determination of neutron cross-sections of the 20 human amino acids and a small database was built for future use in the construction of neutron cross-sections of proteins and other components of the human being cells, also in other industrial applications. (author)
Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi
1993-02-01
A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)
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.
Thermal neutron capture cross section for the K isomer 177Lum
International Nuclear Information System (INIS)
Belier, G.; Roig, O.; Daugas, J.-M.; Giarmana, O.; Meot, V.; Letourneau, A.; Marie, F.; Foucher, Y.; Aupiais, J.; Abt, D.; Jutier, Ch.; Le Petit, G.; Bettoni, C.; Gaudry, A.; Veyssiere, Ch.; Barat, E.; Dautremer, T.; Trama, J.-Ch.
2006-01-01
The thermal neutron radiative capture cross section for the K isomeric state in 177 Lu has been measured for the first time. Several 177 Lu m targets have been prepared and irradiated in various neutron fluxes at the Lauee Langevin Institute in Grenoble and at the CEA reactors OSIRIS and ORPHEE in Saclay. The method consists of measuring the 178 Lu activity by γ-ray spectroscopy. The values obtained in four different neutron spectra have been used to calculate the resonance integral of the radiative capture cross section for 177 Lu m . In addition, an indirect method leads to the determination of the 177 Lu g neutron radiative capture cross section
Energy Technology Data Exchange (ETDEWEB)
Kim, Je Hyun; Shim, Chang Ho [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of); Kim, Sung Hyun [Nuclear Fuel Cycle Waste Treatment Research Division, Research Reactor Institute, Kyoto University, Osaka (Japan); Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo [Ionizing Radiation Center, Nuclear Fuel Cycle Waste Treatment Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho [Ionizing Radiation Center, Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)
2016-12-15
For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.
International Nuclear Information System (INIS)
Kim, Je Hyun; Shim, Chang Ho; Kim, Sung Hyun; Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo; Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho
2016-01-01
For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers
Utilization of new 150-MeV neutron and proton evaluations in MCNP
International Nuclear Information System (INIS)
Little, R.C.; Frankle, S.C.; Hughes, H.G. III; Prael, R.E.
1997-01-01
MCNP trademark and LAHET trademark are two of the codes included in the LARAMIE (Los Alamos Radiation Modeling Interactive Environment) code system. Both MCNP and LAHET are three-dimensional continuous-energy Monte Carlo radiation transport codes. The capabilities of MCNP and LAHET are currently being merged into one code for the Accelerator Production of Tritium (APT) program at Los Alamos National Laboratory. Concurrently, a significant effort is underway to improve the accuracy of the physics in the merged code. In particular, full nuclear-data evaluations (in ENDF6 format) for many materials of importance to APT are being produced for incident neutrons and protons up to an energy of 150-MeV. After processing, cross-section tables based on these new evaluations will be available for use fin the merged code. In order to utilize these new cross-section tables, significant enhancements are required for the merged code. Neutron cross-section tables for MCNP currently specify emission data for neutrons and photons only; the new evaluations also include complete neutron-induced data for protons, deuterons, tritons, and alphas. In addition, no provision in either MCNP or LAHET currently exists for the use of incident charged-particle tables other than for electrons. To accommodate the new neutron-induced data, it was first necessary to expand the format definition of an MCNP neutron cross-section table. The authors have prepared a 150-MeV neutron cross-section library in this expanded format for 15 nuclides. Modifications to MCNP have been implemented so that this expanded neutron library can be utilized
Evaluation of neutron nuclear data for 241Am and 243Am
International Nuclear Information System (INIS)
Kikuchi, Yasuyuki
1982-08-01
Neutron nuclear data of 241 Am and 243 Am were evaluated for JENDL-2. Evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) and (n,4n) reaction cross sections, the resolved and unresolved resonance parameters, the angular or energy distribution of the emitted neutrons, and the average number of neutrons emitted per fission. The fission cross section was evaluated on the basis of newly measured data, and lower values than JENDL-1 were given in the subthreshold energy region. The reliability of the calculation parameters are also much improved, because experimental data became available for the total and capture cross sections of 241 Am in the high energy region. (author)
Evaluation of response matrix of a multisphere neutron spectrometer ...
Indian Academy of Sciences (India)
Abstract. Neutron energy responses of water sphere spectrometers (WSS) to 30 MeV have been calculated by means of Monte Carlo calculations, using the computer code MCNP4C with ENDF/. B-VI.0 neutron cross-section. The calculations have been performed for 3He detector (typical SP9) placed inside 2, 3, 5, 8, ...
On unambiguous parametrization of neutron cross-sections in the low-energy region
International Nuclear Information System (INIS)
Novoselov, G.M.; Kolomiets, V.M.
1982-08-01
One of the most important aims of analysis in the resonance region is the evaluation of neutron resonance parameters on the basis of a given formalism of the theory of nuclear reactions. However, the task of finding resonance parameters from experimental data on the energy dependence of cross-sections is subject to a number of difficulties. These difficulties are not only of a theoretical character associated with the selection of one version or another of the theory taking into account the effects necessary (interference between resonances, Doppler effect etc.), but also involve problems of principle. Whether the set of parameters found is the only possible one within the context of a single formalism used remains open. The specific features of processing the experimental data are such that even with good resolution a number of overlapping resonances (occurring as a result of the fluctuation in inter-level distances or the Doppler effect) may be classified as an isolated resonance. Moreover, even given a very weak inter-level interference and Doppler effect, unambiguous parametrization of the cross-sections is not always possible. In the present paper these questions (the choice of the approximation needed for describing experimentally observed cross-sections, allowance for inter-level interference and the Doppler effect and the possibility of ambiguous reproduction of the resonance structure of cross-sections) are examined with reference to the parametrization of the total cross-sections for non-fissionable nuclei in the low-neutron-energy region
International Nuclear Information System (INIS)
Gerstl, S.A.W.; LaBauve, R.J.; Young, P.G.
1980-05-01
On the example of General Atomic's well-documented Power Generating Fusion Reactor (PGFR) design, this report exercises a comprehensive neutron cross-section and secondary energy distribution (SED) uncertainty analysis. The LASL sensitivity and uncertainty analysis code SENSIT is used to calculate reaction cross-section sensitivity profiles and integral SED sensitivity coefficients. These are then folded with covariance matrices and integral SED uncertainties to obtain the resulting uncertainties of three calculated neutronics design parameters: two critical radiation damage rates and a nuclear heating rate. The report documents the first sensitivity-based data uncertainty analysis, which incorporates a quantitative treatment of the effects of SED uncertainties. The results demonstrate quantitatively that the ENDF/B-V cross-section data files for C, H, and O, including their SED data, are fully adequate for this design application, while the data for Fe and Ni are at best marginally adequate because they give rise to response uncertainties up to 25%. Much higher response uncertainties are caused by cross-section and SED data uncertainties in Cu (26 to 45%), tungsten (24 to 54%), and Cr (up to 98%). Specific recommendations are given for re-evaluations of certain reaction cross-sections, secondary energy distributions, and uncertainty estimates
Measurements of fission cross-sections and of neutron production rates
International Nuclear Information System (INIS)
Billaud, P.; Clair, C.; Gaudin, M.; Genin, R.; Joly, R.; Leroy, J.L.; Michaudon, A.; Ouvry, J.; Signarbieux, C.; Vendryes, G.
1958-01-01
a) Measurements of neutron induced fission cross-sections in the low energy region. The variation of the fission cross sections of several fissile isotopes has been measured and analysed, for neutron energies below 0,025 eV. The monochromator was a crystal spectrometer used in conjunction with a mechanical velocity selector removing higher order Bragg reflections. The fissile material was laid down on the plates of a fission chamber by painting technic. An ionization chamber, having its plates coated with thin 10 B layers, was used as the neutron flux monitor. b) Measurement of the fission cross section of 235 U. We intend to measure the variation of the neutron induced fission cross section of 235 U over the neutron energy range from 1 keV by the time of flight method. The neutron source is the uranium target of a pulsed 28 MeV electron linear accelerator. The detector is a large fission chamber, with parallel plates, containing about 10 g of 235 U (20 deposits of 25 cm diameter). The relative fission data were corrected for the neutron spectrum measured with a set of BF 3 proportional counters. c) Mean number ν of neutrons emitted in neutron induced fission. We measured the value of ν for several fissile isotopes in the case of fission induced by 14 MeV neutrons. The 14 MeV neutrons were produced by D (t, n) α reaction by means of a 300 kV Cockcroft Walton generator. (author) [fr
Thermal neutron cross section measurements for technetium-99
International Nuclear Information System (INIS)
Yates, M.A.; Schroeder, N.C.; Fowler, M.M.
1993-01-01
Technetium, because of its long half-like (213,000 years) and ability to migrate in the environment, is a primary contributor to the long-term radioactivity related risk associated with geologic nuclear waste disposal. One proposal for converting technetium to an environmentally benign element investigating transmutation with an accelerator-based system, (i.e., Accelerator Transmutation of Waste, ATW). Planning for efficient processing of technetium through the transmuter will require knowledge of the thermal neutron cross section for the 99 Tc (n,γ) 100 Tc reaction. The authors have recently remeasured this cross section. Weighed aliquots (19-205 μg) of a NIST traceable 99 Tc standard were irradiated for 30-150 sec using the pneumatic open-quotes rabbitclose quotes system of LANL's Omega West Reactor. The two gamma rays from the 15.7-sec half-life product were measured immediately after irradiation on a high-resolution Ge detector. Thermal fluxes were measured using gold foils and Cd wrapped gold foils. The observation cross section is 19 ± 1 b. This agrees well with the 1977 value but has half the uncertainty
Effective thermal neutron absorption cross section for heterogeneous mixture
International Nuclear Information System (INIS)
Gabanska, B.; Igielski, A.; Krynicka-Drozdowicz, E.; Woznicka, U.
1989-01-01
The first estimations (basing on Umiastowski's theory) of the influence of the sample heterogeneity of the effective thermal neutron absorption cross section were compared with the results obtained for the homogeneous mixture which components and concentration were the same as those of the heterogeneous sample. An experiment was prepared to determine how good this estimate is. Three artificial heterogeneous cylindrical samples (2R = H = 9 cm) were manufactured from pure silver cylinders embedded in plexiglass, keeping the Ag content and varying the size of cylinders (2R = H = 1.0 cm, 0.6 cm and 0.4 cm). Calculations performed show that the experimental effect of the sample heterogeneity can be significant. 5 figs., 5 tabs, 11 refs. (author)
Energy Technology Data Exchange (ETDEWEB)
Muhrer, G., E-mail: muhrer@lanl.gov [Los Alamos National Laboratory, Los Alamos, 87545 NM (United States); Hartl, M.; Mocko, M.; Tovesson, F.; Daemen, L. [Los Alamos National Laboratory, Los Alamos, 87545 NM (United States)
2012-07-21
In the search for moderator materials encapsulated materials have been discussed, but very little is known regarding the effect of encapsulation on neutron moderation properties. As a first step toward a better understanding, we present the measured total neutron cross-section of water confined in silica microspheres and compare the measured data to the predicted theoretical cross-section.
International Nuclear Information System (INIS)
Alonso V, G.; Viais J, J.
1990-10-01
There is developed a method to generate the library of neutron cross sections for the Thermos code by means of the database ENDF-B/IV and the NJOY code. The obtained results are compared with the version previous of the library of neutron cross sections which was processed using the version ENDF-B/III. (Author)
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.
R matrix analysis of 239Pu neutron cross sections in the energy range up to 1000 eV
International Nuclear Information System (INIS)
de Saussure, G.; Perez, R.B.
1990-01-01
This paper reports on the results of an R matrix analysis of the 239 Pu neutron cross sections up to 1000-eV neutron energy. The analysis was performed with the multilevel multichannel Reich-Moore code SAMMY. The method of analysis is describe, and the selection of experimental data is discussed. Some tabular and graphical comparisons between calculated and measured cross sections and transmissions are presented. The statistical properties of the resonance parameters are examined. The resonance parameters are proposed for the new evaluated data files ENDF/B-VI and JEF2
Neutron Cross section Covariances in the Resonance region: 50,53Cr, 54,57Fe and 60Ni
Energy Technology Data Exchange (ETDEWEB)
Oblozinsky, P.; Cho,Y.-S.; Mattoon,C.M.; Mughabghab,S.F.
2010-11-23
We evaluated covariances in the neutron resonance region for capture and elastic scattering cross sections on minor structural materials, {sup 50,53}Cr, {sup 54,57}Fe and {sup 60}Ni. Use was made of the recently developed covariance formalism based on kernel approximation along with data in the Atlas of Neutron Resonances. Our results of most interest for advanced fuel cycle applications, elastic scattering cross section uncertainties at energies around 100 keV, are on the level of about 7-10%.
The importance of fast neutron scattering cross sections for neutron dosimetry in soft tissues
International Nuclear Information System (INIS)
Jahr, R.; Brede, H.J.
1979-05-01
Tissue equivalent plastic materials are used for the construction of accurate neutron dosemeters. As compared to real tissue, in materials most of the oxygen content is replaced by carbon. In order to determine the dose to human tissue a kerma correction factor has to be used. It is shown that the uncertainty (corresponding to 1 delta) of the correction factor at E = 14.5 MeV amounts to at least 5.2%. An important contribution to the uncertainties results from the lack of experimental data of the 12 C(n, n' 3α), 16 O(n,n'p) and 16 O(n,n'α)-cross-sections. These data are to be calculated by subtracting all other cross sections from the total cross section of ( 16 O + n) and ( 12 C + n). It is shown that the uncertainties of the kerma correction factor can be considerably reduced by an accurate measurement of the scattering cross sections of carbon and oxygen. (orig.) [de
International Nuclear Information System (INIS)
Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.
1990-09-01
Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations
International Nuclear Information System (INIS)
Konshin, V.A.
1981-01-01
Neutron cross section calculational methods for actinides in the unresolved resonance energy range (1-150 kev) are discussed, with a special emphasis on calculation of width fluctuation factors for the generalized distribution, as well as for a sub-threshold fission. It is shown that the energy dependence of sub(J), the (n,n') -process competition and the structure in neutron cross section has to be taken into account in the energy range considered. Analysis of different approaches in the statistical theory for heavy nuclei neutron cross-section calculation is given, and it is shown to be important to allow for the (n,γf)-reaction in neutron cross section calculations for fissile nuclei. The use of the non-spherical potential, the Lorentzian spectral factor and the Fermi-gas model involving the collective modes enables to obtain the self-consistent data for all neutron cross sections, including σnγ. (author)
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.; Blokhin, A.I.; Kulagin, N.T.; Pronyaev, V.G.; Simakov, S.P.
1997-01-01
The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs
Energy Technology Data Exchange (ETDEWEB)
Androsenko, A A; Androsenko, P A; Blokhin, A I; Kulagin, N T; Pronyaev, V G; Simakov, S P [Institute of Physics and Power Engineering, Obninsk (Russian Federation)
1997-06-01
The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs.
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)
54Fe neutron elastic and inelastic scattering differential cross sections from 2-6 MeV
Vanhoy, J. R.; Liu, S. H.; Hicks, S. F.; Combs, B. M.; Crider, B. P.; French, A. J.; Garza, E. A.; Harrison, T.; Henderson, S. L.; Howard, T. J.; McEllistrem, M. T.; Nigam, S.; Pecha, R. L.; Peters, E. E.; Prados-Estévez, F. M.; Ramirez, A. P. D.; Rice, B. G.; Ross, T. J.; Santonil, Z. C.; Sidwell, L. C.; Steves, J. L.; Thompson, B. K.; Yates, S. W.
2018-04-01
Measurements of neutron elastic and inelastic scattering cross sections from 54Fe were performed for nine incident neutron energies between 2 and 6 MeV. Measured differential scattering cross sections are compared to those from previous measurements and the ENDF, JENDL, and JEFF data evaluations. TALYS calculations were performed and modifications of the default parameters are found to better describe the experimental cross sections. A spherical optical model treatment is generally adequate to describe the cross sections in this energy region; however, in 54Fe the direct coupling is found to increase suddenly above 4 MeV and requires an increase in the DWBA deformation parameter by approximately 25%. This has little effect on the elastic scattering differential cross sections but makes a significant improvement in both the strength and shape of the inelastic scattering angular distribution, which are found to be very sensitive to the size and extent of the surface absorption region.
Tornow, W.; Howell, C. R.; Crowell, A. S.
2013-12-01
In order to confirm or refute the present discrepancy between data and calculation for the neutron-neutron quasi-free scattering cross section in the neutron-deuteron breakup reaction, we describe a new experimental approach currently being pursued at TUNL.
Total and (n, 2n) neutron cross section measurements on 241Am
International Nuclear Information System (INIS)
Sage, C.
2009-01-01
Neutron induced reaction cross sections on 241 Am have been measured at the IRMM in Geel, Belgium, in the frame of a collaboration between the EC Joint Research Centres IRMM and ITU and French laboratories from CNRS and CEA. Raw material coming from the Atalante facility of CEA Marcoule has been transformed into suitable AmO 2 samples embedded in Al 2 O 3 and Y 2 O 3 matrices. The irradiations for the 241 Am(n, 2n) 240 Am reaction cross section measurement were carried out at the 7 MV Van de Graaff accelerator using the activation technique with quasi mono-energetic neutrons from 8 to 21 MeV produced via the D(d, n) 3 He and the T(d, n) 4 He reactions. The cross section was determined relative to the 27 Al(n, α) 24 Na standard cross section and was investigated for the first time above 15 MeV. The induced activity was measured off-line by standard γ-ray spectrometry using a high purity Ge detector. A special effort was made for the estimation of the uncertainties and the correlations between our experimental points. A different sample of the same isotope 241 Am has been measured in transmission and capture experiments in the resolved resonance region at the neutron ToF facility GELINA. The transmission measurement was performed in two campaigns, with an upgrade of the whole data acquisition system in between, followed by an investigation of its new performances. A preliminary analysis of the resonance parameters tends to confirm the recent evaluation to a higher value for the cross section at the bottom of the first resonances. A new design of C 6 D 6 detectors for capture measurements has been studied, but the data reduction and analysis of the measurement are not part of this work. (author) [fr
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)
Evaluation of neutron nuclear data of 241pu for JENDL-2
International Nuclear Information System (INIS)
Kikuchi, Yasuyuki; Sekine, Nobuo
1984-06-01
Neutron nuclear data of 241 Pu were newly evaluated for JENDL-2. Evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) and (n,4n) reaction cross sections, the resolved and unresolved resonance parameters, the angular and energy distributions of emitted neutrons, and the average number of neutrons emitted per fission. The simultaneous evaluation method was adopted for the fission cross section so as to keep the consistency among the main fissile and fertile material nuclides. The theoretical calculations based on the spherical optical model and the statistical model were also used, when the experimental data were not sufficient. Discussion is given on the evaluation method. (author)
Energy Technology Data Exchange (ETDEWEB)
Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H., E-mail: tiago.tardelli@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear
2013-08-15
Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)
International Nuclear Information System (INIS)
Vuong Huu Tan; Nguyen Canh Hai; Pham Ngoc Son; Tran Tuan Anh
2004-12-01
The neutron capture cross sections of the 109 Ag(n, γ) 110 mAg, 186 W(n, γ) 187 W and 158 Gd(n, γ) 159 Gd have been measured at 55 and 144 keV by the activation method with filtered neutron beams of the Dalat nuclear research reactor. The cross sections were determined relative to the standard capture cross sections of 197 Au using highly purity metallic foils of Ag, W, Gd and Au. The high efficient HPGe detector was used for the gamma rays measurement from the samples, and absolute efficiency calibration was performed by using a set of standard radioisotope sources and a multi-nuclides standard solution. The present results were compared with the previous measurements listed in EXFOR-CINDA, and the evaluated data of ENDF/B-VI. (author)
Neutron cross sections in the unresolved resonance region
International Nuclear Information System (INIS)
Janeva, N.; Lukyanov, A.; Koyumdjieva, N.; Volev, K.
2005-01-01
In this work a development of the characteristic function model, created to reveal the resonance cross section structure in the unresolved resonance region is presented. The main advantage of this model is the calculation of resonance averaged self-shielding factors analytically. To determine average values of the cross sections and their functionals the function of joint statistical distribution of the R-matrix real and imaginary parts should be used. The characteristic function of such distribution is determined and the resonance ladder for the unresolved region is optimized to calculate the group averaged functionals in the same way as it is in the resolved resonance region. The main advantage of this model is the calculation of resonance averaged self-shielding factors analytically. The neutron width energy dependence leads to some deformation in the shape of resonances. This deformation is most apparent near the inelastic scattering threshold. For the case when the inelastic channel momentum is zero we present the formula for level shape bellow and over the inelastic threshold and the calculated resonance deformation in dependence of the position of the resonance in respect to the threshold. (authors)
Benchmark test of evaluated nuclear data files for fast reactor neutronics application
International Nuclear Information System (INIS)
Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki
2007-07-01
A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)
RSAP - A Code for Display of Neutron Cross Section Data and SAMMY Fit Results
International Nuclear Information System (INIS)
Sayer, R.O.
2001-01-01
RSAP is a computer code for display of neutron cross section data and selected SAMMY output. SAMMY is a multilevel R-matrix code for fitting neutron time-of-flight cross-section data using Bayes' method. RSAP, which runs on the Digital Unix Alpha platform, reads ORELA Data Files (ODF) created by SAMMY and uses graphics routines from the PLPLOT package. In addition, RSAP can read data and/or computed values from ASCII files with a format specified by the user. Plot output may be displayed in an X window, sent to a postscript file (rsap.ps), or sent to a color postscript file (rsap.psc). Thirteen plot types are supported, allowing the user to display cross section data, transmission data, errors, theory, Bayes fits, and residuals in various combinations. In this document the designations theory and Bayes refer to the initial and final theoretical cross sections, respectively, as evaluated by SAMMY. Special plot types include Bayes/Data, Theory--Data, and Bayes--Data. Output from two SAMMY runs may be compared by plotting the ratios Theory2/Theory1 and Bayes2/Bayes1 or by plotting the differences (Theory2-Theory1) and (Bayes2-Bayes1)
International Nuclear Information System (INIS)
Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.
1976-01-01
This paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: the spontaneous californium-252 fission neutron spectrum standard field; the thermal-neutron induced uranium-235 fission neutron spectrum standard field; the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN-ΣΣ facilities; the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility; the reference neutron field at the center of the 10% enriched uranium metal, cylindrical, fast critical; the (primary) Intermediate-Energy Standard Neutron Field
International Nuclear Information System (INIS)
Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.
1976-10-01
The paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: (1) the spontaneous californium-252 fission neutron spectrum standard field; (2) the thermal-neutron induced uranium-235 fission neutron spectrum standard field; (3) the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN--ΣΣ facilities; (4) the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility (CFRMF); (5) the reference neutron field at the center of the 10 percent enriched uranium metal, cylindrical, fast critical; and (6) the (primary) Intermediate-Energy Standard Neutron Field
International Nuclear Information System (INIS)
Adib, M.; Abdel-Kawy, A.; Maayouf, R.M.A.; Fayek, M.; Mostafa, M.; Hamouda, I.
1981-09-01
Total neutron cross-section measurements have been performed for natural Nb at liquid nitrogen, room and 425 0 K temperatures in the energy range from 2 MeV - 1 eV. The measurements were performed using two time-of-flight spectrometers installed in front of two of the ET-RR-1 reactor horizontal channels. The neutron diffraction pattern of Nb, at room temperature, was obtained using a double axis crystal spectrometer installed also at the ET-RR-1 reactor. The obtained total neutron cross-sections were analyzed using the single level Breit-Wigner formula. The coherent scattering amplitude was determined from the Bragg reflections observed in the total neutron cross-section of Nb and the analysis of its neutron diffraction pattern. The incoherent and thermal inelastic scattering cross-sections of Nb were determined from the analysis of the total cross-section of Nb beyond the cut-off wavelength. The following results have been obtained: sigmasub(t) = (6.30+-0.20)b; sigmasub(coh) = (6.0+-0.3)b; sigmasub(incoh) = (2.0+-1.0)b; bsub(coh) = (6.91+-0.08)fm
Neutron cross sections of 28 fission product nuclides adopted in JENDL-1
International Nuclear Information System (INIS)
Kikuchi, Yasuyuki; Nakagawa, Tsuneo; Igarasi, Sin-iti; Matsunobu, Hiroyuki; Kawai, Masayoshi; Iijima, Shungo.
1981-02-01
This is the final report concerning the evaluated neutron cross sections of 28 fission product nuclides adopted in the first version of Japanese Evaluated Nuclear Data Library (JENDL-1). These 28 nuclides were selected as being most important for fast reactor calculations, and are 90 Sr, 93 Zr, 95 Mo, 97 Mo, 99 Tc, 101 Ru, 102 Ru, 103 Rh, 104 Ru, 105 Pd, 106 Ru, 107 Pd, 109 Ag, 129 I, 131 Xe, 133 Cs, 135 Cs, 137 Cs, 143 Nd, 144 Ce, 144 Nd, 145 Nd, 147 Pm, 147 Sm, 149 Sm, 151 Sm, 153 Eu and 155 Eu. The status of the experimental data was reviewed over the whole energy range. The present evaluation was performed on the basis of the measured data with the aid of theoretical calculations. The optical and statical models were used for evaluation of the smooth cross sections. An improved method was developed in treating the multilevel Breit-Wigner formula for the resonance region. Various physical parameters and the level schemes, adopted in the present work are discussed by comparing with those used in the other evaluations such as ENDF/B-IV, CEA, CNEN-2 and RCN-2. Furthermore, the evaluation method and results are described in detail for each nuclide. The evaluated total, capture and inelastic scattering cross sections are compared with the other evaluated data and some recent measured data. Some problems of the present work are pointed out and ways of their improvement are suggested. (author)
Evaluation of neutron nuclear data of 9Be for JENDL-3
International Nuclear Information System (INIS)
Shibata, Keiichi
1984-12-01
Neutron nuclear data of 9 Be have been evaluated for JENDL-3 in the energy range from 10 -5 eV to 20 MeV. Evaluated quantities are the total, elastic and inelastic scattering, photon-production, (n,γ), (n,p), (n,d), (n,t) and (n,α) reaction cross sections and the angular and energy distributions of neutrons. The total cross section below 830 keV was calculated with the R-matrix theory. The statistical model was applied to the prediction of the inelastic scattering and charged-particle emission cross sections. The (n,2n) reaction cross section is given by the sum of the inelastic scattering and (n,α 1 ) reaction cross sections. (author)
Additional evaluation of alpha induced neutron production nuclear data. 9Be, 27Al, 28,29,30Si
International Nuclear Information System (INIS)
Murata, Toru; Shibata, Keiichi
2005-01-01
Alpha particle induced neutron production cross sections, emitted neutron energy spectrum and angular distributions were evaluated for the target nucleus 9 Be, 27 Al and Si isotopes; 28 Si, 29 Si and 30 Si in the incident energy region below 15 MeV. (author)
International Nuclear Information System (INIS)
Belgya, T.
2006-01-01
A complete elemental gamma-ray library was measured with our guided thermal beam at the Budapest PGAA facility in the period of 1995-2000. Using this data library in an IAEA CRP on PGAA it was managed to re-normalize the ENSDF intensity data with the Budapest intensities. Based on this renormalization thermal neutron cross sections were deduced for several isotopes. Most of these calculations were done by Richard B. Firestone. The Budapest PGAA-NIPS facilities have been used for routine prompt gamma activation analysis with cold neutrons since the year of 2000. The advantage of the cold neutron beam is that the neutron guide has much higher neutron transmission. This resulted in a gain factor about 20 relative to our thermal guide. For the analytical works a precise comparator technique was developed that is routinely used to determine partial gamma-ray production cross sections. An additional development of our methodology was necessary to be worked out to determine thermal neutron capture cross sections based on the partial gamma-ray production cross sections. In this talk our methodology of radiative capture cross section determination will be presented, including our latest results on 129 I, 204,206,207 Pb and 209 Bi. Most of these works were done in cooperation with people from EU-JRC-IRMM, Geel, Belgium and CEA Cadarache, France. Many partial cross sections of short lived nuclei have been re-measured with our new chopper technique. The uncertainty calculations of the radiative capture cross section determination procedures will be also shown. (authors)
Evaluation method for uncertainty of effective delayed neutron fraction βeff
International Nuclear Information System (INIS)
Zukeran, Atsushi
1999-01-01
Uncertainty of effective delayed neutron fraction β eff is evaluated in terms of three quantities; uncertainties of the basic delayed neutron constants, energy dependence of delayed neutron yield ν d m , and the uncertainties of the fission cross sections of fuel elements. The uncertainty of β eff due to the delayed neutron yield is expressed by a linearized formula assuming that the delayed neutron yield does not depend on the incident energy, and the energy dependence is supplemented by using the detailed energy dependence proposed by D'Angelo and Filip. The third quantity, uncertainties of fission cross section, is evaluated on the basis of the generalized perturbation theory in relation to reaction rate rations such as central spectral indexes or average reaction rate ratios. Resultant uncertainty of β eff is about 4 to 5%s, in which primary factor is the delayed neutron yield, and the secondary one is the fission cross section uncertainty, especially for 238 U. The energy dependence of ν d m systematically reduces the magnitude of β eff about 1.4% to 1.7%, depending on the model of the energy vs. ν d m correlation curve. (author)
SASSI, Total and Differential Elastic and Inelastic Neutron Cross-Sections by Hauser-Feshbach
International Nuclear Information System (INIS)
Benzi, V.; Fabbri, F.; Zuffi, L.
2001-01-01
1 - Nature of physical problem solved: Neutron total and differential elastic and inelastic cross-section evaluation by means of the statistical model of Hauser-Feshbach (1) as modified by D. Goldman (2) (3). The Goldman modification includes the effect of spin-orbit coupling on transmission coefficients. 2 - Method of solution: For numerical integration the Fox-Goodwins method is used. 3 - Restrictions on the complexity of the problem: Angular momentum I less than or equal to 50. Number of excited levels less than or equal to 30
Neutron-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)
International Nuclear Information System (INIS)
Zolotarev, K.I.
2010-11-01
Evaluations of cross sections and their associated covariance matrices have been carried out for five dosimetry reactions: excitation functions were re-evaluated for the 89 Y(n,2n) 88 Y, 93 Nb(n,2n) 92 mNb and 169 Tm(n,2n) 168 Tm reactions over the neutron energy range from threshold up to 40 MeV; excitation functions were re-evaluated for the 59 Co(n,3n) 57 Co and 209 Bi(n,3n) 207 Bi reactions over the neutron energy range from threshold to 85 and 45 MeV, respectively. Uncertainties in the cross sections for all of those reactions were also derived in the form of relative covariance matrices. Benchmark calculations performed for 235 U thermal fission and 252 Cf spontaneous fission neutron spectra show that the integral cross sections calculated from the newly evaluated excitation functions exhibit improved agreement with related experimental data when compared with the equivalent data from the IRDF-2002 library. (author)
Hue, B. M.; Isataev, T.; Erdemchimeg, B.; Artukh, A. G.; Aznabaev, D.; Davaa, S.; Klygin, S. A.; Kononenko, G. A.; Khuukhenkhuu, G.; Kuterbekov, K.; Lukyanov, S. M.; Mikhailova, T. I.; Maslov, V. A.; Mendibaev, K.; Sereda, Yu M.; Penionzhkevich, Yu E.; Vorontsov, A. N.
2017-12-01
Preliminary results of measurements of the total reaction cross sections σR and neutron removal cross section σ-xn for weakly bound 6He, 8Li, 9Be and 10Be nuclei at energy range (20-35) A MeV with 28Si target is presented. The secondary beams of light nuclei were produced by bombardment of the 22Ne (35 A MeV) primary beam on Be target and separated by COMBAS fragment-separator. In dispersive focal plane a horizontal slit defined the momentum acceptance as 1% and a wedge degrader of 200 μm Al was installed. The Bρ of the second section of the fragment-separator was adjusted for measurements in energy range (20-35) A MeV. Two-neutron removal cross sections for 6He and 10Be and one -neutron removal cross sections 8Li and 9Be were measured.
Database for 238U inelastic scattering cross section evaluation
International Nuclear Information System (INIS)
Kanda, Yukinori; Fujikawa, Noboru; Kawano, Toshihiko
1993-10-01
There are discrepancies among evaluated neutron inelastic scattering cross sections for 238 U in the evaluated nuclear data files, JENDL-3, ENDF/B-VI, JEF-2, BROND-2 and CENDL-2. Re-evaluating them is internationally being discussed to obtain the best outcome which can be accepted in common at the present by experts in the world. This report has been compiled to review the discrepancies among the evaluations in the present data files and to provide a common database for the re-evaluation work (author)
Energy Technology Data Exchange (ETDEWEB)
de Saussure, G.; Perez, R.B. (Oak Ridge National Lab., TN (USA))
1982-01-01
Several problems related to the measurement, analysis and evaluation of the neutron cross sections of the main fertile and fissile nuclides in the resonance region are reviewed. In particular the ENDF/B-V representation of these cross sections is discussed. In recent years little progress has been made in improving our knowledge of the resolved resonance parameters of the fertile nuclei. It is suggested that this absence of progress is due to a lack of adequate methodologies to deal with the systematic errors arising from uncertainties in the analysis of the measurements. The ENDF/B treatment of the unresolved resonance region is commented on and the authors recommend the validation of the unresolved resonance range evaluations with appropriate transmission and self-indication measurements.
Theoretical and experimental cross sections for neutron reactions on 64Zinc
International Nuclear Information System (INIS)
Rutherford, D.A.
1987-01-01
Accurate measurements of the 64 Zn (n,2n) 64 Cu and 64 Zn (n,p) 63 Zn cross sections at 14.8 MeV have been made using a Texas Nuclear Neutron Generator and the activation technique. A NaI(T1) spectrometer (using two 6'' x 6'' NaI detectors/crystals) was used to measure the gamma radiation emitted in coincidence from the positron-emitting decay products. The measurements were made relative to 65 Cu (n,2n) /64/Cu and 63 Cu (n,2n) 62 Cu cross sections, which have similar half-lives, radiation emission, and were previously measured to high accuracy (2 percent). The value obtained for the (n,2n) measurement was 199 /+-/ 6 millibarns, and a value of 176 /+-/ 4.5 millibarns was obtained for the (n,p) measurement. In concert, a theoretical analysis of neutron induced reactions on /64/Zn was performed at Los Alamos National Laboratory using the Hauser-Feshbach statistical theory in the GNASH code over an energy range of 100 keV to 20 MeV. Calculations included width fluctuation corrections, direct reaction contributions, and preequilibrium corrections above 6 MeV. Neutron optical model potentials were determined for zinc. The theoretical values agree with the new 14.8 MeV measurements approximately within experimental error, with calculations of 201 millibarns for the (n,2n) cross section and 170 millibarns for the (n,p) cross section. Results from the analysis will be made available in National Evaluated Nuclear Data Format (ENDF/B) for fusion energy applications. 50 refs., 34 figs., 10 tabs
Simple, empirical approach to predict neutron capture cross sections from nuclear masses
Couture, A.; Casten, R. F.; Cakirli, R. B.
2017-12-01
Background: Neutron capture cross sections are essential to understanding the astrophysical s and r processes, the modeling of nuclear reactor design and performance, and for a wide variety of nuclear forensics applications. Often, cross sections are needed for nuclei where experimental measurements are difficult. Enormous effort, over many decades, has gone into attempting to develop sophisticated statistical reaction models to predict these cross sections. Such work has met with some success but is often unable to reproduce measured cross sections to better than 40 % , and has limited predictive power, with predictions from different models rapidly differing by an order of magnitude a few nucleons from the last measurement. Purpose: To develop a new approach to predicting neutron capture cross sections over broad ranges of nuclei that accounts for their values where known and which has reliable predictive power with small uncertainties for many nuclei where they are unknown. Methods: Experimental neutron capture cross sections were compared to empirical mass observables in regions of similar structure. Results: We present an extremely simple method, based solely on empirical mass observables, that correlates neutron capture cross sections in the critical energy range from a few keV to a couple hundred keV. We show that regional cross sections are compactly correlated in medium and heavy mass nuclei with the two-neutron separation energy. These correlations are easily amenable to predict unknown cross sections, often converting the usual extrapolations to more reliable interpolations. It almost always reproduces existing data to within 25 % and estimated uncertainties are below about 40 % up to 10 nucleons beyond known data. Conclusions: Neutron capture cross sections display a surprisingly strong connection to the two-neutron separation energy, a nuclear structure property. The simple, empirical correlations uncovered provide model-independent predictions of
IAEA CIELO Evaluation of Neutron-induced Reactions on 235U and 238U Targets
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
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
International Nuclear Information System (INIS)
Roig, O.; Belier, G.; Meot, V.; Daugas, J.-M.; Romain, P.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Letourneau, A.; Marie, F.; Veyssiere, Ch.
2006-01-01
Thermal neutron radiative capture and burn-up measurements of the K isomeric state in 177Lu form part of an original method to indirectly obtain the neutron super-elastic scattering cross section at thermal energy. Neutron super-elastic scattering, also called neutron inelastic acceleration, occurs during the neutron collisions with an excited nuclear level. In this reaction, the nucleus could partly transfer its excitation energy to the scattered neutron
Measurement of keV-neutron capture cross sections and capture gamma-ray spectra of Er isotopes
International Nuclear Information System (INIS)
Harun-Ar-Rashid, A.K.M.; Igashira, Masayuki; Ohsaki, Toshiro
2000-01-01
Neutron capture cross sections and capture γ-ray spectra of 166,167, 168 Er were measured in the energy region of 10 to 550 keV. The measurements were performed with a pulsed 7 Li(p,n) 7 Be neutron source and a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique and the standard capture cross sections of gold were used to derive the capture cross sections. The errors of the derived cross sections were about 5%. The present results were compared with other measurements and evaluations. The observed capture γ-ray pulse-height spectra were unfolded to obtain the corresponding γ-ray spectra. An anomalous shoulder was observed around 3 MeV in each of the capture γ-ray spectra. (author)
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)
Energy Technology Data Exchange (ETDEWEB)
Furuta, T.; Masuda, T.; Tsurita, Y.; Hashimoto, A.; Miyajima, N. [Department of Nuclear Engineering, Nagoya University, Nagoya, Aichi (Japan); Sakane, H.; Shibata, M.; Yamamoto, H.; Kawade, K.
1999-03-01
A d-D gas target producing monoenergetic neutrons has been constructed for measurement of activation cross sections in the energy region of 3 to 7 MeV at Van de Graaff accelerator of Nagoya University. Neutron spectra and neutron fluxes were measured as a function of the incident deuteron energy. Preliminary results of activation cross sections were obtained for reactions {sup 27}Al(n, p){sup 27}Mg, {sup 27}Al(n, {alpha}){sup 24}Na, {sup 47}Ti(n, p){sup 47}Sc, {sup 56}Fe(n, p){sup 56}Mn, {sup 58}Ni(n, p){sup 58}Co and {sup 64}Zn(n, p){sup 64}Cu. The results are compared with the evaluated values of JENDL-3.2. A well-type HPGe detector was used for highly efficient detection. (author)
Helium production cross section Measurement of Pb and Sn for 14.9 MeV neutrons
Energy Technology Data Exchange (ETDEWEB)
Takao, Yoshiyuki; Fujimoto, Toshihiro; Ozaki, Shuji; Muramasu, Masatomo; Nakashima, Hideki [Kyushu Univ., Fukuoka (Japan); Kanda, Yukinori; Ikeda, Yujiro
1998-03-01
Helium production cross sections of lead and tin for 14.9 MeV neutrons were measured by helium accumulation method. Lead and tin samples were irradiated with FNS, an intense d-T neutron source of JAERI. The amount of helium produced in the samples by the neutron irradiation was measured with the Helium Atoms Measurement System (HAMS) at Kyushu University. As the samples contained a small amount of helium because of their small helium production cross sections at 14.9 MeV, the samples were evaporated by radiation from a tungsten filament to decrease background gases at helium measurement. Uncertainties of the present results were less than {+-}4.4%. The results were compared with other experimental data in the literature and also compared with the evaluated values in JENDL-3.2. (author)
Evaluation of temperature dependent neutron resonance integrals
International Nuclear Information System (INIS)
Menon, S.V.G.; Sahni, D.C.
1975-01-01
The Fourier transform method is extended for evaluating temperature dependent resonance integrals and Doppler coefficients. With the temperature dependent cross-sections, the slowing-down equation is transformed into a Fredholm integral equation of second kind. A method of solution is presented using the familiar Gauss-Hermite quadrature formulae. As a byproduct of the above technique, a fast and accurate method for computing the resonance integral J-function is given. (orig.) [de
Neutron cross section standards for the energy region above 20 MeV
International Nuclear Information System (INIS)
1991-01-01
These proceedings of a specialists' meeting on Neutron cross section standards for the energy region above 20 MeV are divided into 6 sessions bearing on: - session 1: status of the date base for (n-p) scattering (2 conferences) - session 2: status of nucleon-nucleon phase shift calculations (1 conference) - session 3: recent and planned experimental work on n-p cross section measurements and facilities (7 conferences) - session 4: Instruments for utilizing the H (n.n) standard for neutron fluence measurement (4 conferences) - session 5: proposal for other neutron cross-section standards (4 conferences) - session 6: monitor reactions for radiation dosimetry (3 conferences)
Remarks on the comparison of cross section libraries for neutron metrology
International Nuclear Information System (INIS)
Zijp, W.L.; Nolthenius, H.J.; Appelman, K.H.
1977-01-01
Cross section libraries in a 620 group structure were available from different origin: CCC-112B, DETAN-74 and ENDF/B-IV. For a few well known neutron spectra (CFRMF spectrum, ΣΣ spectrum, fission neutron spectrum, HFR neutron spectrum) a comparison was made of the available experimental reaction rates in foil detectors and the reaction rates as calculated with the different cross section libraries. This investigation is dealing with the consistency of cross section data within a library, and the consistency of activity data in actual reaction rate determinations. Some preliminary conclusions are given
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
International Nuclear Information System (INIS)
Iwamoto, Yosuke; Shigyo, Nobuhiro; Satoh, Daiki
2002-01-01
Neutron-production double-differential cross sections for 870-MeV π + and π - and 2.1-GeV π + mesons incident on iron and lead targets were measured with NE213 liquid scintillators by time-of-flight technique. NE213 liquid scintillators 12.7 cm in diameter and 12.7 cm thick were placed in directions of 15, 30, 60, 90, 120 and 150deg. The typical flight path length was 15 m. Neutron detection efficiencies were derived from the calculation results of SCINFUL and CECIL codes. The experimental results were compared with the JAM code. The double differential cross sections calculated by the JAM code disagree with experimental data at neutron energies below about 30 MeV. JAM overestimates π + -incident neutron-production cross sections in forward angles at neutron energies of 100 to 500 MeV. (author)
EJ2-MCNPlib. Contents of the JEF-2.2 based neutron cross-section library for MCNP4A
International Nuclear Information System (INIS)
Hogenbirk, A.; Oppe, J.
1995-05-01
In this report a description is given of the EJ2-MCNPlib library. The EJ2-MCNPlib library is to be used for reactivity/critically calculations and general neutron/photon transport calculations with the Monte Carlo code MCNP4A. The library is based on the European JEF-2.2 nuclear data evaluation and contains data for all (i.e. 313) nuclides available on this evaluation.The cross-section data were generated using the NJOY cross-section processing code system, version 91.118. For easy reference cross-section plots are given in this report for the total, elastic and absorption cross sections for all nuclides on the EJ2-MCNPlib library. Furthermore, for verification purposes a graphical intercomparison is given of the results of standard benchmark calculations performed with JEF-2.2 cross-section data and with ENDF/B-V cross-section data (whenever available). 6 refs
New evaluation of 238U neutron resonance parameters
International Nuclear Information System (INIS)
Derrien, Herve; Leal, Luiz C.; Larson, Nancy M.
2003-01-01
The neutron resonance parameters of 238 U were obtained in the energy range 1 keV to 20 keV from a SAMMY Reich-Moore analysis of high resolution transmission measurements performed at ORELA. In the energy range 1 keV to 10 keV, the analysis used as prior values the ENDF/B-VI resonance parameters. The analysis in the energy range 10 keV to 20 keV resulted in the creation of a set of resonance parameters for the representation of the cross section in this energy range. The results are compared to the ENDF/B-VI evaluation. Some statistical properties of the new resonance parameters are examined. (author)
Measurements of the {sup 235}U(n,f) cross section in the 3 to 30 MeV neutron energy region
Energy Technology Data Exchange (ETDEWEB)
Carlson, A.D.; Wasson, O.A. [National Institute of Standards and Technology, Gaithersburg, MD (United States); Lisowski, P.W. [Los Alamos National Lab., NM (United States)] [and others
1991-12-31
To improve the accuracy of the {sup 235}U(n,f) cross section, measurements have been made of this standard cross section at the target 4 facility at Los Alamos National Laboratory (LANL). The data were obtained at the 20-meter flight path of that facility. The fission reaction rate was determined with a fast parallel plate ionization chamber and the neutron fluence was measured with an annular proton recoil telescope. The measurements provide the shape of the {sup 235}U(n,f) cross section relative to the hydrogen scattering cross section for neutron energies from about 3 to 30 MeV neutron energy. The data have been normalized to the very accurately known value near 14 MeV. The results are in good agreement with the ENDF/B-VI evaluation up to about 15 MeV neutron energy. Above this energy differences as large as 5% are observed.
International Nuclear Information System (INIS)
Tang Hongqing; Qi Bujia; Zhou Zuying; Sa Jun; Ke Zunjian; Sui Qingchang; Xia Haihong; Shen Guanren
1992-01-01
The status and technique of double differential cross section measurement of secondary neutrons in the incident neutron energy range 9 to 13 MeV is reviewed with emphasis on the work done at CIAE. There are scarce measurements of secondary neutron double differential cross sections in this energy region up to now. A main difficulty for this is lack of an applicable monoenergetic neutron source. When monoenergetic neutron energy reaches 8 Me/v, the break-up neutrons from the d + D or p + T reaction starts to become significant. It is difficult to get a pure secondary neutron spectrum induced only by monoenergetic neutrons. To solve this problem an abnormal fast neutron TOF facility was designed and tested. Double differential neutron emission cross sections of 238 U and 209 Bi at 10 MeV were obtained by combining the data measured by both normal and abnormal TOF spectrometers and a good agreement between measurement and calculation was achieved
Status of measured neutron cross sections of transactinium isotopes in the fast region
International Nuclear Information System (INIS)
Igarasi, S.
1976-01-01
This paper reviews present status of measured neutron cross sections of transactinium isotopes from a viewpoint of requested data in application field of the nuclear data. The measured cross sections from 1 keV to 15 MeV are examined. Comparison between different data sets is mainly performed on the fission cross sections
Estimated 55Mn and 90Zr cross section covariances in the fast neutron energy region
International Nuclear Information System (INIS)
Pigni, M.T.; Herman, M.; Oblozinsky, P.
2008-01-01
We completed estimates of neutron cross section covariances for 55 Mn and 90 Zr, from keV range to 25 MeV, considering the most important reaction channels, total, elastic, inelastic, capture, and (n,2n). The nuclear reaction model code EMPIRE was used to calculate sensitivity to model parameters by perturbation of parameters that define the optical model potential, nuclear level densities and strength of the pre-equilibrium emission. The sensitivity analysis was performed with the set of parameters which reproduces the ENDF/B-VII.0 cross sections. The experimental data were analyzed and both statistical and systematic uncertainties were extracted from almost 30 selected experiments. Then, the Bayesian code KALMAN was used to combine the sensitivity analysis and the experiments to obtain the evaluated covariance matrices
Estimated 55Mn and 90Zr cross section covariances in the fast neutron energy region
Energy Technology Data Exchange (ETDEWEB)
Pigni,M.T.; Herman, M.; Oblozinsky, P.
2008-06-24
We completed estimates of neutron cross section covariances for {sup 55}Mn and {sup 90}Zr, from keV range to 25 MeV, considering the most important reaction channels, total, elastic, inelastic, capture, and (n,2n). The nuclear reaction model code EMPIRE was used to calculate sensitivity to model parameters by perturbation of parameters that define the optical model potential, nuclear level densities and strength of the pre-equilibrium emission. The sensitivity analysis was performed with the set of parameters which reproduces the ENDF/B-VII.0 cross sections. The experimental data were analyzed and both statistical and systematic uncertainties were extracted from almost 30 selected experiments. Then, the Bayesian code KALMAN was used to combine the sensitivity analysis and the experiments to obtain the evaluated covariance matrices.
Determination of Unknown Neutron Cross Sections for the Production of Medical Isotopes
Energy Technology Data Exchange (ETDEWEB)
Stephen E. Binney
2004-04-09
Calculational assessment and experimental verification of certain neutron cross sections that are related to widely needed new medical isotopes. Experiments were performed at the Oregon State University TRIGA Reactor and the High Flux Irradiation Reactor at Oak Ridge National Laboratory.
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.
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.
Measurements of Integral Cross Section Ratios in Two Dosimetry Benchmark Neutron Fields
International Nuclear Information System (INIS)
Fabry, A.; Czock, K.H.
1974-12-01
In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the 103 Rh(n,n') 103m Rh and 58 Ni(n,p) 58 Co integral cross sections have been accurately measured relatively to the 115 In(n,n') 115m In cross section in the 235 U thermal dission neutron spectrum and in the MOLΣΣ Intermediate-Energy Standard Neutron field. In this last neutron field, the data are related also to the 235 U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific 103 Rh(n,n') 103m Rh differential-energy cross section among the existing, conflicting data. (author)
Measurements of integral cross section ratios in two dosimetry benchmark neutron fields
International Nuclear Information System (INIS)
Fabry, A.; Czock, K.H.
1974-12-01
In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the 103 Rh(n,n') 103m Rh and 58 Ni(n,p) 58 Co integral cross sections have been accurately measured relatively to the 115 In(n,n') 115m In cross section in the 235 U thermal fission neutron spectrum and in the MOL-ΣΣ intermediate-energy standard neutron field. In this last neutron field, the data are related also to the 235 U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific 103 Rh(n,n') 103m Rh differential-energy cross section among the existing, conflicting data. (author)
Neutron capture cross section of /sup 197/Au: A standard for stellar nucleosynthesis
International Nuclear Information System (INIS)
Ratynski, W.; Kaeppeler, F.
1988-01-01
We have measured the neutron capture cross section of gold using the 7 Li(p,n) 7 Be reaction for neutron production. This reaction not only provides the integrated neutron flux via the 7 Be activity of the target, but also allows for the simulation of a Maxwellian neutron energy spectrum at kT = 25 keV. As this spectrum is emitted in a forward cone of 120 0 opening angle, the cross section can be measured in good geometry and independent of any other standard. Systematic uncertainties were studied experimentally in a series of activations. The final stellar cross section at kT = 25 keV was found to be 648 +- 10 mb, and extrapolation to the common s-process temperature kT = 30 keV yields 582 +- 9 mb. This result is used for renormalization of a number of cross sections which had been measured relative to gold
Measurements of integral cross section ratios in two dosimetry benchmark neutron fields
Energy Technology Data Exchange (ETDEWEB)
Fabry, A [CEN-SCK, Mol (Belgium); Czock, K H [International Atomic Energy Agency, Laboratory Seibersdorf, Vienna (Austria)
1974-12-01
In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the {sup 103}Rh(n,n'){sup 103m}Rh and {sup 58}Ni(n,p){sup 58}Co integral cross sections have been accurately measured relatively to the {sup 115}In(n,n'){sup 115m} In cross section in the {sup 235}U thermal fission neutron spectrum and in the MOL-{sigma}{sigma} intermediate-energy standard neutron field. In this last neutron field, the data are related also to the {sup 235}U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific {sup 103}Rh(n,n'){sup 103m}Rh differential-energy cross section among the existing, conflicting data. (author)
Measurements of Integral Cross Section Ratios in Two Dosimetry Benchmark Neutron Fields
Energy Technology Data Exchange (ETDEWEB)
Fabry, A. [CEN-SCK, Mol (Belgium); Czock, K. H. [International Atomic Energy Agency, Vienna (Austria)
1974-12-15
In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the {sup 103}Rh(n,n'){sup 103m}Rh and {sup 58}Ni(n,p){sup 58}Co integral cross sections have been accurately measured relatively to the {sup 115}In(n,n'){sup 115m}In cross section in the {sup 235}U thermal dission neutron spectrum and in the MOL{Sigma}{Sigma} Intermediate-Energy Standard Neutron field. In this last neutron field, the data are related also to the {sup 235}U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific {sup 103}Rh(n,n'){sup 103m}Rh differential-energy cross section among the existing, conflicting data. (author)
Energy Technology Data Exchange (ETDEWEB)
Smith, D.L.
1987-11-01
Ratio measurements are routinely employed in studies of neutron interaction processes in order to generate new differential cross-section data or to test existing differential cross-section information through examination of the corresponding response in integral neutron spectra. Interpretation of such data requires that careful attention be given to details of the neutron spectra involved in these measurements. Two specific tasks are undertaken in the present investigation: (1) Using perturbation theory, a formula is derived which permits one to relate the ratio measured in a realistic quasimonoenergetic spectrum to the desired pure monoenergetic ratio. This expression involves only the lowest-order moments of the neutron energy distribution and corresponding parameters which serve to characterize the energy dependence of the differential cross sections, quantities which can generally be estimated with reasonable precision from the uncorrected data or from auxiliary information. (2) Using covariance methods, a general formalism is developed for calculating the uncertainty of a measured integral cross-section ratio which involves an arbitrary neutron spectrum. This formalism is employed to further examine the conditions which influence the sensitivity of such measured ratios to details of the neutron spectra and to their uncertainties. Several numerical examples are presented in this report in order to illustrate these principles, and some general conclusion are drawn concerning the development and testing of neutron cross-section data by means of ratio experiments. 16 refs., 1 fig., 4 tabs.
International Nuclear Information System (INIS)
Smith, D.L.
1987-11-01
Ratio measurements are routinely employed in studies of neutron interaction processes in order to generate new differential cross-section data or to test existing differential cross-section information through examination of the corresponding response in integral neutron spectra. Interpretation of such data requires that careful attention be given to details of the neutron spectra involved in these measurements. Two specific tasks are undertaken in the present investigation: (1) Using perturbation theory, a formula is derived which permits one to relate the ratio measured in a realistic quasimonoenergetic spectrum to the desired pure monoenergetic ratio. This expression involves only the lowest-order moments of the neutron energy distribution and corresponding parameters which serve to characterize the energy dependence of the differential cross sections, quantities which can generally be estimated with reasonable precision from the uncorrected data or from auxiliary information. (2) Using covariance methods, a general formalism is developed for calculating the uncertainty of a measured integral cross-section ratio which involves an arbitrary neutron spectrum. This formalism is employed to further examine the conditions which influence the sensitivity of such measured ratios to details of the neutron spectra and to their uncertainties. Several numerical examples are presented in this report in order to illustrate these principles, and some general conclusion are drawn concerning the development and testing of neutron cross-section data by means of ratio experiments. 16 refs., 1 fig., 4 tabs
Measurement of thermal neutron cross section for {sup 241}Am(n,f) reaction
Energy Technology Data Exchange (ETDEWEB)
Kobayashi, Katsuhei; Yamamoto, Shuji; Fujita, Yoshiaki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.; Miyoshi, Mitsuharu; Kimura, Itsuro; Kanno, Ikuo; Shinohara, Nobuo
1997-03-01
Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the {sup 241}Am(n,f) reaction has been measured relative to the reference value of 586.2b for the {sup 235U}(n,f) reaction. For the present measurement, electrodeposited layers of {sup 241}Am and {sup 235}U have been employed as back-to-back type double fission chambers. The present result at neutron energy of 0.0253 eV is 3.15 {+-} 0.097b. The ENDF/B-VI data is in good agreement with the present value, while the JENDL-3.2 data is lower by 4.2%. The evaluated data in JEF-2.2 and by Mughabghab are higher by 0.9% and 1.6%, respectively than the present result. The ratios of the earlier experimental data to the present value are distributed between 0.89 and 1.02. (author)
Revision of the inelastic scattering cross section evaluation of 238U for CENDL-2.1
International Nuclear Information System (INIS)
Tang Guoyou; Zhang Guohui; Shi Zhaomin; Chen Jinxiang
1995-11-01
Revised evaluated data for the inelastic neutron scattering cross-section and the secondary neutron spectrum are presented for U-238 in graphical form, compared with the earlier data that exist in the evaluated nuclear data libraries ENDF/B-6 and JENDL-3. The new data will be included in the Chinese evaluated nuclear data library CENDL-2.1. (author). 14 refs, 9 figs
International Nuclear Information System (INIS)
Fu, C.Y.; Hetrick, D.M.
1986-07-01
An update of the ENDF/B-V Mod-3 evaluation for natural iron is described. The cross sections of (n,n') and (n,2n) reactions are revised. Energy-angle correlations in the secondary (n,n') neutrons are introduced in the ENDF/B-V formats. Anisotropic angular distributions are provided for the secondary neutrons in (n,2n), (n,np), and (n,nα) reactions. Revelant integral results, microscopic data, and nuclear model calculations that influence the revised results are summarized. 54 refs., 9 figs., 2 tabs
Mei, B.; Tu, X. L.; Wang, M.
2018-04-01
An evident odd-even staggering (OES) in fragment cross sections has been experimentally observed in many fragmentation and spallation reactions. However, quantitative comparisons of this OES effect in different reaction systems are still scarce for neutron-rich nuclei near the neutron drip line. By employing a third-order difference formula, the magnitudes of this OES in extensive experimental cross sections are systematically investigated for many neutron-rich nuclei with (N -Z ) from 1 to 23 over a broad range of atomic numbers (Z ≈3 -50 ). A comparison of these magnitude values extracted from fragment cross sections measured in different fragmentation and spallation reactions with a large variety of projectile-target combinations over a wide energy range reveals that the OES magnitude is almost independent of the projectile-target combinations and the projectile energy. The weighted average of these OES magnitudes derived from cross sections accurately measured in different reaction systems is adopted as the evaluation value of the OES magnitude. These evaluated OES magnitudes are recommended to be used in fragmentation and spallation models to improve their predictions for fragment cross sections.
Energy Technology Data Exchange (ETDEWEB)
Auclair, J M; Hubert, P; Joly, R; Vendryes, G; Jacrot, B; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Galula, M [Centre National de la Recherche Scientifique (CNRS), 91 - Gif-sur-Yvette (France)
1955-07-01
It presents the experimental measurements of cross section of fissile nuclei for slow neutrons to improve the understanding of some heavy nuclei of great importance in the study of nuclear reactors. The different experiments are divided in three categories. In the first part, it studied the variation with energy of the cross sections of natural uranium, {sup 233}U, {sup 235}U and {sup 239}Pu. Two measurement techniques are used: the time-of-flight spectrometer and the crystal spectrometer. In a second part, the fission cross sections of {sup 233}U and {sup 239}Pu for thermal neutrons are compared using a neutron flux from EL-2 going through a double fission chamber. The matter quantity contained in each source is measured by counting the {alpha} activity with a solid angle counter. Finally, the average cross section of {sup 236}U for a spectra of neutrons from the reactor is measured by studying the {beta} activity of {sup 237}U formed by the reaction {sup 236}U (n, {gamma}) {sup 237}U in a sample of {sup 236}U irradiated in the Saclay reactor (EL-2). (M.P.)
Calibration and evaluation of neutron moisturemeter
International Nuclear Information System (INIS)
Tang Zhangxiong; Hu Jiangchao; Sun Laiyan; Wang Huaihui; Wu Weixue
1992-02-01
Factors influencing the calibration curve of neutron moisture meter, such as soil type, texture, volume weight and depth, were studied. When the soil bulk density water content is between 15% to 45%, the calibration curve is approximately a straight line, and the intercept and slope are only influenced by the above factors. The growing plants also influence the calibration curve slightly. The measuring error for top soil (< 20 cm) is larger. The relative error between neutron method and weighing method is about 8%. The neutron method has many advantages such as non-interfering, simple, fast and non-time-delay
Analysis of the 239Pu neutron cross sections from 300 to 2000 eV
International Nuclear Information System (INIS)
Derrien, H.; de Saussure, G.
1990-01-01
A recent high-resolution measurement of the neutron fission cross section of 239 Pu has allowed the extension from 1 to 2 keV of a previously reported resonance analysis of the neutron cross sections, and an improvement of the previous analysis in the range 0.3 to 1 keV. This report analyzes this region. 8 refs., 1 fig., 2 tabs
Measurement of total reaction cross sections of exotic neutron rich nuclei
International Nuclear Information System (INIS)
Mittig, W.; Chouvel, J.M.; Wen Long, Z.
1987-01-01
Total reaction cross-sections of neutron rich nuclei from C to Mg in a thick Si-target have been measured using the detection of the associated γ-rays in a 4Π-geometry. This cross-section strongly increases with neutron excess, indicating an increase of as much as 15% of the reduced strong absorption radius with respect to stable nuclei
The CERN n_TOF Facility: Neutron Beams Performances for Cross Section Measurements
Chiaveri, E; Andrzejewski, J; Audouin, L; Barbagallo, M; Bécares, V; Bečvář, F; Belloni, F; Berthoumieux, E; Billowes, J; Boccone, V; Bosnar, D; Brugger, M; Calviani, M; Calviño, F; Cano-Ott, D; Carrapiço, C; Cerutti, F; Chin, M; Colonna, N; Cortés, G; Cortés-Giraldo, M A; Diakaki, M; Domingo-Pardo, C; Duran, I; Dressler, R; Dzysiuk, N; Eleftheriadis, C; Ferrari, A; Fraval, K; Ganesan, S; García, A R; Giubrone, G; Gómez-Hornillos, M B; Gonçalves, I F; González-Romero, E; Griesmayer, E; Guerrero, C; Gunsing, F; Gurusamy, P; Hernández-Prieto, A; Jenkins, D G; Jericha, E; Kadi, Y; Käppeler, F; Karadimos, D; Kivel, N; Koehler, P; Kokkoris, M; Krtička, M; Kroll, J; Lampoudis, C; Langer, C; Leal-Cidoncha, E; Lederer, C; Leeb, H; Leong, L S; Losito, R; Mallick, A; Manousos, A; Marganiec, J; Martínez, T; Massimi, C; Mastinu, P F; Mastromarco, M; Meaze, M; Mendoza, E; Mengoni, A; Milazzo, P M; Mingrone, F; Mirea, M; Mondalaers, W; Paradela, C; Pavlik, A; Perkowski, J; Plompen, A; Praena, J; Quesada, J M; Rauscher, T; Reifarth, R; Riego, A; Robles, M S; Roman, F; Rubbia, C; Sabaté-Gilarte, M; Sarmento, R; Saxena, A; Schillebeeckx, P; Schmidt, S; Schumann, D; Tagliente, G; Tain, J L; Tarrío, D; Tassan-Got, L; Tsinganis, A; Valenta, S; Vannini, G; Variale, V; Vaz, P; Ventura, A; Versaci, R; Vermeulen, M J; Vlachoudis, V; Vlastou, R; Wallner, A; Ware, T; Weigand, M; Weiss, C; Wright, T; Žugec, P
2014-01-01
This paper presents the characteristics of the existing CERN n\\_TOF neutron beam facility (n\\_TOF-EAR1 with a flight path of 185 meters) and the future one (n\\_TOF EAR-2 with a flight path of 19 meters), which will operate in parallel from Summer 2014. The new neutron beam will provide a 25 times higher neutron flux delivered in 10 times shorter neutron pulses, thus offering more powerful capabilities for measuring small mass, low cross section and/or high activity samples.
Cross-section of single-crystal materials used as thermal neutron filters
International Nuclear Information System (INIS)
Adib, M.
2005-01-01
Transmission properties of several single crystal materials important for neutron scattering instrumentation are presented. A computer codes are developed which permit the calculation of thermal diffuse and Bragg-scattering cross-sections of silicon., and sapphire as a function of material's constants, temperature and neutron energy, E, in the range 0.1 MeV .A discussion of the use of their single-crystal as a thermal neutron filter in terms of the optimum crystal thickness, mosaic spread, temperature, cutting plane and tuning for efficient transmission of thermal-reactor neutrons is given
Testing of the IRDF-90 cross-section library in benchmark neutron spectra
International Nuclear Information System (INIS)
Nolthenius, H.J.; Zsolnay, E.M.; Szondi, E.J.
1993-09-01
The new version of the International Reactor Dosimetry File IRDF-90 (called ''Version April 1993'') has been tested by calculation of average cross-sections and their uncertainties in a coarse three energy group structure and by neutron spectrum adjustments in reference neutron spectra. This paper presents the results obtained and compares them with the corresponding ones of the old IRDF-85 and with the data of the Nuclear Data Guide for Reactor Neutron Metrology. The applicability of the new library in the field of neutron metrology is discussed. (orig.)
Neutron cross-sections database for amino acids and proteins analysis
Energy Technology Data Exchange (ETDEWEB)
Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin, E-mail: dante@ien.gov.br, E-mail: fferreira@ien.gov.br, E-mail: Chaffin@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Rocha, Helio F. da, E-mail: hrocha@gbl.com.br [Universidade Federal do Rio de Janeiro (IPPMG/UFRJ), Rio de Janeiro, RJ (Brazil). Instituto de Pediatria
2015-07-01
Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)
Neutron cross-sections database for amino acids and proteins analysis
International Nuclear Information System (INIS)
Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin; Rocha, Helio F. da
2015-01-01
Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)
Evaluation of room-scattered neutrons at the JNC Tokai neutron reference field
Energy Technology Data Exchange (ETDEWEB)
Yoshida, Tadayoshi; Tsujimura, Norio [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan). Tokai Works; Oyanagi, Katsumi [Japan Radiation Engineering Co., Ltd., Hitachi, Ibaraki (Japan)
2002-09-01
Neutron reference fields for calibrating neutron-measuring devices in JNC Tokai Works are produced by using radionuclide neutron sources, {sup 241}Am-Be and {sup 252}Cf sources. The reference field for calibration includes scattered neutrons from the material surrounding sources, wall, floor and ceiling of the irradiation room. It is, therefore, necessary to evaluate the scattered neutrons contribution and their energy spectra at reference points. Spectral measurements were performed with a set of Bonner multi-sphere spectrometers and the reference fields were characterized in terms of spectral composition and the fractions of room-scattered neutrons. In addition, two techniques stated in ISO 10647, the shadow-cone method and the polynomial fit method, for correcting the contributions from the room-scattered neutrons to the readings of neutron survey instruments were compared. It was found that the two methods gave an equivalent result within a deviation of 3.3% at a source-to-detector distance from 50cm to 500cm. (author)
Evaluation of room-scattered neutrons at the JNC Tokai neutron reference field
International Nuclear Information System (INIS)
Yoshida, Tadayoshi; Tsujimura, Norio
2002-01-01
Neutron reference fields for calibrating neutron-measuring devices in JNC Tokai Works are produced by using radionuclide neutron sources, 241 Am-Be and 252 Cf sources. The reference field for calibration includes scattered neutrons from the material surrounding sources, wall, floor and ceiling of the irradiation room. It is, therefore, necessary to evaluate the scattered neutrons contribution and their energy spectra at reference points. Spectral measurements were performed with a set of Bonner multi-sphere spectrometers and the reference fields were characterized in terms of spectral composition and the fractions of room-scattered neutrons. In addition, two techniques stated in ISO 10647, the shadow-cone method and the polynomial fit method, for correcting the contributions from the room-scattered neutrons to the readings of neutron survey instruments were compared. It was found that the two methods gave an equivalent result within a deviation of 3.3% at a source-to-detector distance from 50cm to 500cm. (author)
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
Thermal neutron radiative capture cross-section of 186W(n, γ)187W reaction
International Nuclear Information System (INIS)
Tan, V H; Son, P N
2016-01-01
The thermal neutron radiative capture cross section for 186 W(n, γ) 187 W reaction was measured by the activation method using the filtered neutron beam at the Dalat research reactor. An optimal composition of Si and Bi, in single crystal form, has been used as neutron filters to create the high-purity filtered neutron beam with Cadmium ratio of R cd = 420 and peak energy E n = 0.025 eV. The induced activities in the irradiated samples were measured by a high resolution HPGe digital gamma-ray spectrometer. The present result of cross section has been determined relatively to the reference value of the standard reaction 197 Au(n, γ) 198 Au. The necessary correction factors for gamma-ray true coincidence summing, and thermal neutron self-shielding effects were taken into account in this experiment by Monte Carlo simulations. (paper)
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)
Amino acids analysis using grouping and parceling of neutrons cross sections techniques
International Nuclear Information System (INIS)
Voi, Dante Luiz Voi; Rocha, Helio Fenandes da
2002-01-01
Amino acids used in parenteral administration in hospital patients with special importance in nutritional applications were analyzed to compare with the manufactory data. Individual amino acid samples of phenylalanine, cysteine, methionine, tyrosine and threonine were measured with the neutron crystal spectrometer installed at the J-9 irradiation channel of the 1 kW Argonaut Reactor of the Instituto de Engenharia Nuclear (IEN). Gold and D 2 O high purity samples were used for the experimental system calibration. Neutron cross section values were calculated from chemical composition, conformation and molecular structure analysis of the materials. Literature data were manipulated by parceling and grouping neutron cross sections. (author)
International Nuclear Information System (INIS)
Poenitz, Erik
2010-01-01
In future nuclear facilities, the materials lead and bismuth can play a more important role than in today's nuclear reactors. Reliable cross section data are required for the design of those facilities. In particular the neutron transport in the lead spallation target of an Accelerator-Driven Subcritical Reactor strongly depends on the inelastic neutron scattering cross sections in the energy region from 0.5 MeV to 6 MeV. In the recent 20 years, elastic and inelastic neutron scattering cross sections were measured with high precision for a variety of elements at the PTB time-of-flight spectrometer. The D(d,n) reaction was primarily used for the production of neutrons. Because of the Q value of the reaction and the available deuteron energies, neutrons in the energy range from 6 MeV to 16 MeV can be produced. For the cross section measurement at lower energies, however, another neutron producing reaction is required. The 15 N(p,n) 15 O reaction was chosen, as it allows the production of monoenergetic neutrons with up to 5.7MeV energy. In this work, the 15 N(p,n) reaction was studied with focus on the suitability as a source for monoenergetic neutrons in scattering experiments. This includes the measurement of differential cross sections for the neutron producing reaction and the choice of optimum target conditions. Differential elastic and inelastic neutron scattering cross sections were measured for lead at four energies in the region from 2 MeV to 4 MeV incident neutron energy using the time-of-flight technique. A lead sample with natural isotopic composition was used. NE213 liquid scintillation detectors with well-known detection efficiencies were used for the detection of the scattered neutrons. Angle-integrated cross sections were determined by a Legendre polynomial expansion using least-squares methods. Additionally, measurements were carried out for isotopically pure 209 Bi and 181 Ta samples at 4 MeV incident neutron energy. Results are compared with other
Quantitative neutron radiography using neutron absorbing honeycomb
International Nuclear Information System (INIS)
Tamaki, Masayoshi; Oda, Masahiro; Takahashi, Kenji; Ohkubo, Kohei; Tasaka, Kanji; Tsuruno, Akira; Matsubayashi, Masahito.
1993-01-01
This investigation concerns quantitative neutron radiography and computed tomography by using a neutron absorbing honeycomb collimator. By setting the neutron absorbing honeycomb collimator between object and imaging system, neutrons scattered in the object were absorbed by the honeycomb material and eliminated before coming to the imaging system, but the neutrons which were transmitted the object without interaction could reach the imaging system. The image by purely transmitted neutrons gives the quantitative information. Two honeycombs were prepared with coating of boron nitride and gadolinium oxide and evaluated for the quantitative application. The relation between the neutron total cross section and the attenuation coefficient confirmed that they were in a fairly good agreement. Application to quantitative computed tomography was also successfully conducted. The new neutron radiography method using the neutron-absorbing honeycomb collimator for the elimination of the scattered neutrons improved remarkably the quantitativeness of the neutron radiography and computed tomography. (author)
International Nuclear Information System (INIS)
Padron, I.; Dominguez, O.; Sarria, P. Sandin, C.
1996-01-01
The time -of-Flight neutron spectrometry technique by associated particle method was improved using a D-T neutron generator at Laboratory of Nuclear Analysis. This technique was implemented for double differential cross section measurements and supported by the IAEA Project CUB/01/005. An stilbene scintillation detector (dia=100 mm, length=50 mm) was used as principal neutron detector detector and was situated outside a hole in the concrete wall. This way the fligth path was extended and the scattered neutron cone accurate collimated throught the 2 m concrete wall. For the associated particle α detection a thin plastic NE-102 scint illator was used, as well as, two scintilation detectors and a long counter for the neutron flux monitoring. In this TOF neutron spectrometer (3.40 m flight path) a 1.7 nseg. temporal resolution was obtained
International Nuclear Information System (INIS)
Smith, D.L.; Meadows, J.W.; Greenwood, L.R.
1990-01-01
Integral neutron-reaction cross sections have been measured, relative to the U-238 neutron fission cross-section standard, for 27 reactions which are of contemporary interest in various nuclear applications (e.g., fast-neutron dosimetry, neutron radiation damage and the production of long-lived activities which affect nuclear waste disposal). The neutron radiation field employed in this study was produced by bombarding a thick Be-metal target with 7-MeV deuterons from an accelerator. The experimental results are reported along with detailed information on the associated measurement uncertainties and their correlations. These data are also compared with corresponding calculated values, based on contemporary knowledge of the differential cross sections and of the Be(d,n) neutron spectrum. Some conclusions are reached on the utility of this procedure for neutron-reaction data testing
Energy Technology Data Exchange (ETDEWEB)
An, So Hyun; Lee, Young Ouk; Lee, Cheol Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Young Seok [National Fusion Research Institute, Daejeon (Korea, Republic of)
2007-10-15
It is essential that neutron cross sections are measured precisely for many areas of research and technique. In Korea, these experiments have been performed in the Pohang Neutron Facility (PNF) with the pulsed neutron facility based on the 100 MeV electron linear accelerator. In PNF, the neutron energy spectra have been measured for different water levels inside the moderator and compared with the results of the MCNPX calculation. The optimum size of the water moderator has been determined on the base of these results. In this study, Monte Carlo simulations for the TOF technique were performed and neutron spectra of neutrons were calculated to predict the measurements.
Evaluation of secondary and prompt fission neutron spectra
Energy Technology Data Exchange (ETDEWEB)
Porodzinskij, Yu.V.; Sukhovitskij, E.Sh. [Radiation Physics and Chemistry Problems Inst., Minsk-Sosny (Belarus)
1997-03-01
A simple model allowing to split neutron emission spectra into reaction partials is suggested. Predicted spectra of (n,n`{gamma}), (n,n`f), etc appear to be much harder than usually evaluated. (author)
High-energy Neutron-induced Fission Cross Sections of Natural Lead and Bismuth-209
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.
A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry Applications
Energy Technology Data Exchange (ETDEWEB)
Alpan, F.A. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)
2011-07-01
A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and {sup 58}Ni(n,{gamma}) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contribution and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within {+-} 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190. (authors)
International Nuclear Information System (INIS)
Kim, Eun Ae; Shvetsov, Valery; Cho, Moo Hyun; Won, Nam Kung; Kim, Kwang Soo; Yang, Sung Chul; Lee, Man Woo; Kim, Guin Yun; Yi, Kyoung Rak; Choi, Hong Yub; Ro, Tae Ik; Mizumoto, Motoharu; Katabuchi, Tatsuya; Igashira, Masayuki
2012-01-01
The measurement of neutron cross section using Time-Of-Flight (TOF) method gives significant information for the nuclear data research. In the present work, the neutron total cross section of 56 Fe and 57 Fe has been measured in the energy range between 10 eV and 100 keV by using the neutron beam produced from 3-GeV proton synchrotron accelerator. The 3-GeV proton synchrotron accelerator is located at Japan Proton Accelerator Research Complex (J-PARC) facility in Tokai village. In this study, the neutron total cross section data measured by 6 Li glass scintillator detector was compared with the evaluated values of ENDF/B-VII.0
Energy Technology Data Exchange (ETDEWEB)
Kim, Eun Ae; Shvetsov, Valery; Cho, Moo Hyun [Pohang University of Science and Technology, Pohang (Korea, Republic of); Won, Nam Kung [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Kim, Kwang Soo; Yang, Sung Chul; Lee, Man Woo; Kim, Guin Yun [Kyungpook National University, Daegu (Korea, Republic of); Yi, Kyoung Rak; Choi, Hong Yub; Ro, Tae Ik [Dong-A University, Pusan (Korea, Republic of); Mizumoto, Motoharu; Katabuchi, Tatsuya; Igashira, Masayuki [Tokyo Institute of Technology, Tokyo (Japan)
2012-05-15
The measurement of neutron cross section using Time-Of-Flight (TOF) method gives significant information for the nuclear data research. In the present work, the neutron total cross section of {sup 56}Fe and {sup 57}Fe has been measured in the energy range between 10 eV and 100 keV by using the neutron beam produced from 3-GeV proton synchrotron accelerator. The 3-GeV proton synchrotron accelerator is located at Japan Proton Accelerator Research Complex (J-PARC) facility in Tokai village. In this study, the neutron total cross section data measured by {sup 6}Li glass scintillator detector was compared with the evaluated values of ENDF/B-VII.0
Leal, L C; Kitis, G; Guber, K H; Yuasa nakagawa, K; Koehler, P E; Quaranta, A
2002-01-01
The experimental determination of neutron cross section data has always been of primary importance in Nuclear Physics. Many of the salient features of nuclear levels and densities can be determined from the resonant structure of such cross sections and of their decay scheme. An associated importance of precise neutron induced reaction cross sections has resulted from the worldwide interest in Accelerator Driven Systems (ADS) that has emerged at CERN and elsewhere. Many applications, such as accelerator-based transmutation of nuclear waste, energy amplification medical research, astrophysical applications and also fusion research require nuclear data that quantitatively and qualitatively go beyond the presently available traditional evaluation.\\\\ \\\\We consider a spallation driven TOF facility at the CERN-PS with an unprecedented neutron flux (1000 times the existing ones) in the broad energy range between 1 eV and 250 MeV and with very high energy resolution. The present concept for an intense neutron source m...
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
Neutron radiation capture; Radiatsionnyj zakhvat nejtronov
Energy Technology Data Exchange (ETDEWEB)
NONE
1987-12-31
For all stable and experimentally studied radionuclides evaluated data are presented on cross sections of thermal neutrons, on resonance integrals and medium neutron cross sections with energy of 30 KeV. Refs, figs and tabs.
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.
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.)
Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE
International Nuclear Information System (INIS)
Leconte, P.; Di-Salvo, J.; Antony, M.; Pepino, A.; Hentati, A.
2012-01-01
Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of 157 Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of nat Gd which is (49360 ± 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1σ, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 ± 500) b. (authors)
Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE
Energy Technology Data Exchange (ETDEWEB)
Leconte, P.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Hentati, A. [International School in Nuclear Engineering, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)
2012-07-01
Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of {sup 157}Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of {sup nat}Gd which is (49360 {+-} 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1{sigma}, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 {+-} 500) b. (authors)
Directory of Open Access Journals (Sweden)
Leconte Pierre
2017-01-01
Full Text Available EXCALIBUR is an integral transmission experiment based on the fast neutron source produced by the bare highly enriched fast burst reactor CALIBAN, located in CEA/DAM Valduc (France. Two experimental campaigns have been performed, one using a sphere of diameter 17 cm and one using two cylinders of 17 cm diameter 9 cm height, both made of metallic Uranium 238. A set of 15 different dosimeters with specific threshold energies have been employed to provide information on the neutron flux attenuation as a function of incident energy. Measurements uncertainties are typically in the range of 0.5–3% (1σ. The analysis of these experiments is performed with the TRIPOLI4 continuous energy Monte Carlo code. A calculation benchmark with validated simplifications is defined in order to improve the statistical convergence under 2%. Various 238U evaluations have been tested: JEFF-3.1.1, ENDF/B-VII.1 and the IB36 evaluation from IAEA. A sensitivity analysis is presented to identify the contribution of each reaction cross section to the integral transmission rate. This feedback may be of interest for the international effort on 238U, through the CIELO project.
Andrianova, Olga; Lomakov, Gleb; Manturov, Gennady
2017-09-01
The neutron transmission experiments are one of the main sources of information about the neutron cross section resonance structure and effect in the self-shielding. Such kind of data for niobium and silicon nuclides in energy range 7 keV to 3 MeV can be obtained from low-resolution transmission measurements performed earlier in Russia (with samples of 0.027 to 0.871 atom/barn for niobium and 0.076 to 1.803 atom/barn for silicon). A significant calculation-to-experiment discrepancy in energy range 100 to 600 keV and 300 to 800 keV for niobium and silicon, respectively, obtained using the evaluated nuclear data library ROSFOND, were found. The EVPAR code was used for estimation the average resonance parameters in energy range 7 to 600 keV for niobium. For silicon a stochastic optimization method was used to modify the resolved resonance parameters in energy range 300 to 800 keV. The improved ROSFOND evaluated nuclear data files were tested in calculation of ICSBEP integral benchmark experiments.
Review and Assessment of Neutron Cross Section and Nubar Covariances for Advanced Reactor Systems
Energy Technology Data Exchange (ETDEWEB)
Maslov,V.M.; Oblozinsky, P.; Herman, M.
2008-12-01
In January 2007, the National Nuclear Data Center (NNDC) produced a set of preliminary neutron covariance data for the international project 'Nuclear Data Needs for Advanced Reactor Systems'. The project was sponsored by the OECD Nuclear Energy Agency (NEA), Paris, under the Subgroup 26 of the International Working Party on Evaluation Cooperation (WPEC). These preliminary covariances are described in two recent BNL reports. The NNDC used a simplified version of the method developed by BNL and LANL that combines the recent Atlas of Neutron Resonances, the nuclear reaction model code EMPIRE and the Bayesian code KALMAN with the experimental data used as guidance. There are numerous issues involved in these estimates of covariances and it was decided to perform an independent review and assessment of these results so that better covariances can be produced for the revised version in future. Reviewed and assessed are uncertainties for fission, capture, elastic scattering, inelastic scattering and (n,2n) cross sections as well as prompt nubars for 15 minor actinides ({sup 233,234,236}U, {sup 237}Np, {sup 238,240,241,242}Pu, {sup 241,242m,243}Am and {sup 242,243,244,245}Cm) and 4 major actinides ({sup 232}Th, {sup 235,238}U and {sup 239}Pu). We examined available evaluations, performed comparison with experimental data, taken into account uncertainties in model parameterization and made use state-of-the-art nuclear reaction theory to produce the uncertainty assessment.
Neutron slowing down and transport in a medium of constant cross section. I. Spatial moments
International Nuclear Information System (INIS)
Cacuci, D.G.; Goldstein, H.
1977-01-01
Some aspects of the problem of neutron slowing down and transport have been investigated in an infinite medium consisting of a single nuclide scattering elastically and isotropically without absorption and with energy-independent cross sections. The method of singular eigenfunctions has been applied to the Boltzmann equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. Formulas have been obtained for the lethargy dependent spatial moments of the scalar flux applicable in the limit of large lethargy. In deriving these formulas, use has been 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 have been greatly aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use has also been made of the methods of combinatorial analysis and of computer evaluation, via FORMAC, of complicated sequences of manipulations. It has been 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, valid through terms of order of u -5 . Higher order correction terms could be obtained at the expense of additional computer time. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent the end product of this investigation
Leconte, Pierre; Bernard, David
2017-09-01
EXCALIBUR is an integral transmission experiment based on the fast neutron source produced by the bare highly enriched fast burst reactor CALIBAN, located in CEA/DAM Valduc (France). Two experimental campaigns have been performed, one using a sphere of diameter 17 cm and one using two cylinders of 17 cm diameter 9 cm height, both made of metallic Uranium 238. A set of 15 different dosimeters with specific threshold energies have been employed to provide information on the neutron flux attenuation as a function of incident energy. Measurements uncertainties are typically in the range of 0.5-3% (1σ). The analysis of these experiments is performed with the TRIPOLI4 continuous energy Monte Carlo code. A calculation benchmark with validated simplifications is defined in order to improve the statistical convergence under 2%. Various 238U evaluations have been tested: JEFF-3.1.1, ENDF/B-VII.1 and the IB36 evaluation from IAEA. A sensitivity analysis is presented to identify the contribution of each reaction cross section to the integral transmission rate. This feedback may be of interest for the international effort on 238U, through the CIELO project.
What can be learnt from the channel analysis of the 232Th neutron fission cross section
International Nuclear Information System (INIS)
Abou Yehia, H.; Jary, J.; Trochon, J.; Boldeman, J.W.; Musgrove, A.R. de L.
1979-10-01
Channel analyses of the neutron fission cross section of 232 Th have been made in two laboratories. The calculated fission cross sections and fission fragment anisotropies are compared with the experimental data. Despite some differences in the methods used, the conclusions on the physical aspects of the fission process are very similar
International Nuclear Information System (INIS)
Fukahori, Tokio; Iwamoto, Osamu; Chiba, Satoshi
2003-01-01
For an accelerator-driven nuclear waste transmutation system, it is very important to estimate sub-criticality of core system for feasibility and design study of the system. The fission cross section in the intermediate energy range has an important role. A program FISCAL has been developed to calculate neutron-, proton- and photon-induced fission cross sections in the energy region from several tens of MeV to 3 GeV. FISCAL adopts the systematics considering experimental data for Ag- 243 Am. It is found that unified description of neutron-, proton- and photon-induced fission cross sections is available. (author)
International Nuclear Information System (INIS)
Adib, M.; Abdel Kawy, A.; Habib, N.; El Mesiry, M.
2010-01-01
An investigation of pyrolytic graphite (PG) crystal as an efficient second order neutron filter at tuned boundary crossings has been carried out. The neutron transmission through PG crystal at these tuned crossing points as a function of first- and second-order wavelengths were calculated in terms of PG mosaic spread and thickness. The filtering features of PG crystals at these tuned boundary crossings were deduced. It was shown that, there are a large number of tuned positions at double and triple boundary crossings of the curves (hkl) are very promising as tuned filter positions. However, only fourteen of them are found to be most promising ones. These tuned positions are found to be within the neutron wavelengths from 0.133 up to 0.4050 nm. A computer package GRAPHITE has been used in order to provide the required calculations in the whole neutron wavelength range in terms of PG mosaic spread and its orientation with respect to incident neutron beam direction. It was shown that 0.5 cm thick PG crystal with angular mosaic spread of 2 0 is sufficient to remove 2nd-order neutrons at the wavelengths corresponding to the positions of the intersection boundaries curves (hkl).
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
Molecular dynamical and structural studies for the bakelite by neutron cross section measurements
International Nuclear Information System (INIS)
Voi, D.L.
1992-05-01
Neutron reaction cross sections were determined by transmission and scattering measurements, to study the dynamics and molecular structure of calcined bakelites. Total cross sections were determined, with a deviation smaller than 5%, from the literature values, by neutron transmission method and a specially devised approximation. These cross sections were then correlated with data obtained with infra-red spectroscopy, elemental analysis and other techniques to get the probable molecular formulae of bakelite. Double differential scattering cross sections, scattering law values and frequency distributions were determined with 15% error using the neutron inelastic scattering method. The frequency distributions as well as the overall results from all experimental techniques used in this work allowed to suggest a structural model like polycyclic hydrocarbons, for calcined bakelite at 800 0 C. (author)
Neutron based evaluation in support of NEAMS
Energy Technology Data Exchange (ETDEWEB)
Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bourke, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian Simon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-06-07
The primary objective of the Advanced Non-Destructive fuel Examination (ANDE) work package is to develop capability that has the potential to accelerate insight and development of ceramic and metallic fuels. Establishing unique validation opportunities for new models is a key component of this effort. To explore opportunities a series of interactions were held with NEAMS modelers at LANL. The focus was to identify experiments that draw on the unique capabilities of neutron scattering and imaging for studies of nuclear fuel particularly in areas where experimental data can be valuable for of models validation. The neutron characterization techniques applied in the ANDE program span length scales from millimeter to micrometer to angstroms. Spatial heterogeneities of interest include cracks, pores and inclusions, crystal structure, phase composition, stoichiometry texture, chemistry and atomic thermal motion. Neutrons offer characterization opportunities that are distinct from other probes such as X-rays, electrons or protons. This report describes a variety of opportunities whereby neutron data can be related to models and lists some opportunities.
International Nuclear Information System (INIS)
Kehayias, J.J.
2000-01-01
Nutritional status of patients can be evaluated by monitoring changes in body composition, including depletion of protein and muscle, adipose tissue distribution and changes in hydration status, bone or cell mass. Fast neutron activation (for N and P) and neutron inelastic scattering (for C and O) are used to assess in vivo elements characteristic of specific body compartments. The fast neutrons are produced with a sealed deuterium-tritium (D-T) neutron generator. This method provides the most direct assessment of body composition. Non-bone phosphorus for muscle is measured by the 31 P(n,α) 28 Al reaction, and nitrogen for protein via the (n,2n) fast neutron reaction. Inelastic neutron scattering is used for the measurement of total body carbon and oxygen. Carbon is used to derive body fat, after subtracting carbon contributions due to protein, bone and glycogen. Carbon-to-oxygen (C/O) ratio is used to measure distribution of fat and lean tissue in the body and to monitor small changes of lean mass and its quality. In addition to evaluating the efficacy of new treatments, the system is used to study the mechanisms of lean tissue depletion with aging and to investigate methods for preserving function and quality of life in the elderly. (author)
International Nuclear Information System (INIS)
Kosako, Toshiso; Nakamura, Takashi; Iwai, Satoshi; Katsuki, Shinji; Kamata, Masashi.
1983-08-01
The energy-dependent response function of a multi-cylinder moderating-type BF 3 counter, so-called Bonner counter, was calculated by the time-dependent multi-group Monte Carlo code, TMMCR. The calculated response function was evaluated experimentally for neutron energy below about 50 keV down to epithermal energy by the time-of-flight method combining with a large lead pile at the Nuclear Engineering Research Laboratory, University of Tokyo and also above 50 keV by using the monoenergetic neutron standard field a t the Electrotechnical Laboratory. The time delay in the polyethylene moderator of the Bonner counter due to multiple collisions with hydrogen was analyzed by the TMMCR code and used for the time-spectrum analysis of the time-of-flight measurement. The response function obtained by these two experiments showed good agreement with the calculated results. This Bonner counter having a response function evaluated from thermal to MeV energy range was used for spectrometry and dosimetry of environmental neutrons around some nuclear facilities. The neutron spectra and dose measured in the environment around a 252 Cf fission source, fast neutron source reactor and electron synchrotron were all in good agreement with the calculated results and the measured results with other neutron detectors. (author)
Recommended evaluation procedure for photonuclear cross section
Energy Technology Data Exchange (ETDEWEB)
Lee, Young-Ouk; Chang, Jonghwa; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In order to generate photonuclear cross section library for the necessary applications, data evaluation is combined with theoretical evaluation, since photonuclear cross sections measured cannot provide all necessary data. This report recommends a procedure consisting of four steps: (1) analysis of experimental data, (2) data evaluation, (3) theoretical evaluation and, if necessary, (4) modification of results. In the stage of analysis, data obtained by different measurements are reprocessed through the analysis of their discrepancies to a representative data set. In the data evaluation, photonuclear absorption cross sections are evaluated via giant dipole resonance and quasi-deutron mechanism. With photoabsorption cross sections from the data evaluation, theoretical evaluation is applied to determine various decay channel cross sections and emission spectra using equilibrium and preequilibrium mechanism. After this, the calculated results are compared with measured data, and in some cases the results are modified to better describe measurements. (author)
Evaluation of light nuclides for fusion neutronics application
International Nuclear Information System (INIS)
Shibata, Keiichi
1984-01-01
Neutron nuclear data of 6 Li, 7 Li, 12 C and 16 O have been evaluated for JENDL-3. The methods of evaluation and the results are described. The present evaluated data are employed for the analysis of the Japan-US co-operative experiment using FNS. (author)
International Nuclear Information System (INIS)
Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.
2007-01-01
Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm [fr
MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1
International Nuclear Information System (INIS)
Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis
2015-01-01
A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35 Cl and 233 U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.
Improvement of evaluated neutron nuclear data for {sup 237}Np and {sup 241}Am
Energy Technology Data Exchange (ETDEWEB)
Nakagawa, Tsuneo; Iwamoto, Osamu; Hasegawa, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-01-01
The nuclear data of {sup 237}Np and {sup 241}Am that are particularly important among the minor actinides were investigated by comparing JENDL-3.2 with the recent evaluated data and available experimental data. As a result of the study, several defects of JENDL-3.2 data were revealed. They were improved on the basis of experimental data or recent evaluated data. For the both nuclides, main quantities revised in the present work were the resonance parameters, cross sections, angular and energy distributions of secondary neutrons, number of neutrons per fission. The data were given in the neutron energy range from 10{sup -5} eV to 20 MeV, and compiled in the ENDF-6 format. (author)
Neutron cross-sections for advanced nuclear systems: the n_TOF project at CERN
Directory of Open Access Journals (Sweden)
Barbagallo M.
2014-01-01
Full Text Available The study of neutron-induced reactions is of high relevance in a wide variety of fields, ranging from stellar nucleosynthesis and fundamental nuclear physics to applications of nuclear technology. In nuclear energy, high accuracy neutron data are needed for the development of Generation IV fast reactors and accelerator driven systems, these last aimed specifically at nuclear waste incineration, as well as for research on innovative fuel cycles. In this context, a high luminosity Neutron Time Of Flight facility, n_TOF, is operating at CERN since more than a decade, with the aim of providing new, high accuracy and high resolution neutron cross-sections. Thanks to the features of the neutron beam, a rich experimental program relevant to nuclear technology has been carried out so far. The program will be further expanded in the near future, thanks in particular to a new high-flux experimental area, now under construction.
International Nuclear Information System (INIS)
Peelle, R.W.; de Sassure, G.
1977-01-01
Refined knowledge of the thermal neutron cross sections of the fissile nuclides and of the (n,α) reaction standards, together with the reasonably well known energy dependence of the latter, have permitted resonance-region and low-keV fissile nuclide cross sections to be based on these standards together with count-rate ratios observed as a function of energy using a pulsed ''white'' source. As one evaluates cross sections for energies above 20 keV, optimum results require combination of cross section shape measurements with all available absolute measurements. The assumptions of the ''thermal normalization method'' are reviewed, and an opinion is given of the status of some of the standards required for its use. The complications which may limit the accuracy of results using the method are listed and examples are given. For the 235 U(n,f) cross section, the option is discussed of defining resonance-region fission integrals as standards. The area of the approximately 9 eV resonances in this nuclide may be known to one percent accuracy, but at present the fission integral from 0.1 to 1.0 keV is known to no better than about two percent. This uncertainty is based on the scatter among independent results, and has not been reduced by the most recent measurements. This uncertainty now limits the accuracy attainable for the 235 U(n,f) cross section below about 50 keV. Suggestions are given to indicate how future detailed work might overcome past sources of error
Neutron cross section measurements at n-TOF for ADS related studies
Mastinu, P. F.; Abbondanno, U.; Aerts, G.; Álvarez, H.; Alvarez-Velarde, F.; Andriamonje, S.; Andrzejewski, J.; Assimakopoulos, P.; Audouin, L.; Badurek, G.; Bustreo, N.; aumann, P.; vá, F. Be; Berthoumieux, E.; Calviño, F.; Cano-Ott, D.; Capote, R.; Carrillo de Albornoz, A.; Cennini, P.; Chepel, V.; Chiaveri, E.; Colonna, N.; Cortes, G.; Couture, A.; Cox, J.; Dahlfors, M.; David, S.; Dillmann, I.; Dolfini, R.; Domingo-Pardo, C.; Dridi, W.; Duran, I.; Eleftheriadis, C.; Embid-Segura, M.; Ferrant, L.; Ferrari, A.; Ferreira-Marques, R.; itzpatrick, L.; Frais-Kölbl, H.; Fujii, K.; Furman, W.; Guerrero, C.; Goncalves, I.; Gallino, R.; Gonzalez-Romero, E.; Goverdovski, A.; Gramegna, F.; Griesmayer, E.; Gunsing, F.; Haas, B.; Haight, R.; Heil, M.; Herrera-Martinez, A.; Igashira, M.; Isaev, S.; Jericha, E.; Kadi, Y.; Käppeler, F.; Karamanis, D.; Karadimos, D.; Kerveno, M.; Ketlerov, V.; Koehler, P.; Konovalov, V.; Kossionides, E.; Krti ka, M.; Lamboudis, C.; Leeb, H.; Lindote, A.; Lopes, I.; Lozano, M.; Lukic, S.; Marganiec, J.; Marques, L.; Marrone, S.; Massimi, C.; Mengoni, A.; Milazzo, P. M.; Moreau, C.; Mosconi, M.; Neves, F.; Oberhummer, H.; O'Brien, S.; Oshima, M.; Pancin, J.; Papachristodoulou, C.; Papadopoulos, C.; Paradela, C.; Patronis, N.; Pavlik, A.; Pavlopoulos, P.; Perrot, L.; Plag, R.; Plompen, A.; Plukis, A.; Poch, A.; Pretel, C.; Quesada, J.; Rauscher, T.; Reifarth, R.; Rosetti, M.; Rubbia, C.; Rudolf, G.; Rullhusen, P.; Salgado, J.; Sarchiapone, L.; Savvidis, I.; Stephan, C.; Tagliente, G.; Tain, J. L.; Tassan-Got, L.; Tavora, L.; Terlizzi, R.; Vannini, G.; Vaz, P.; Ventura, A.; Villamarin, D.; Vincente, M. C.; Vlachoudis, V.; Vlastou, R.; Voss, F.; Walter, S.; Wendler, H.; Wiescherand, M.; Wisshak, K.
2006-05-01
A neutron Time-of-Flight facility (n_TOF) is available at CERN since 2001. The innovative features of the neutron beam, in particular the high instantaneous flux, the wide energy range, the high resolution and the low background, make this facility unique for measurements of neutron induced reactions relevant to the field of Emerging Nuclear Technologies, as well as to Nuclear Astrophysics and Fundamental Nuclear Physics. The scientific motivations that have led to the construction of this new facility are here presented. The main characteristics of the n_TOF neutron beam are described, together with the features of the experimental apparata used for cross-section measurements. The main results of the first measurement campaigns are presented. Preliminary results of capture cross-section measurements of minor actinides, important to ADS project for nuclear waste transmutation, are finally discussed.
Measurement of dijet cross sections for events with a leading neutron in photoproduction at HERA
International Nuclear Information System (INIS)
Breitweg, J.; Chekanov, S.; Derrick, M.; Krakauer, D.; Magill, S.; Musgrave, B.; Pellegrino, A.; Repond, J.; Stanek, R.; Yoshida, R.; Mattingly, M.C.K.; Antonioli, P.; Bari, G.; Basile, M.; Bellagamba, L.; Boscherini, D.; Bruni, A.; Bruni, G.; Cara Romeo, G.; Cifarelli, L.; Cindolo, F.; Contin, A.; Corradi, M.; De Pasquale, S.; Giusti, P.; Iacobucci, G.; Levi, G.; Margotti, A.; Massam, T.; Nania, R.; Palmonari, F.; Pesci, A.; Sartorelli, G.; Zichichi, A.; Amelung, C.; Bornheim, A.; Brock, I.; Coboeken, K.; Crittenden, J.; Deffner, R.; Hartmann, H.; Heinloth, K.; Hilger, E.; Irrgang, P.; Jakob, H.-P.; Kappes, A.; Katz, U.F.; Kerger, R.; Paul, E.; Rautenberg, J.; Schnurbusch, H.; Stifutkin, A.; Tandler, J.; Voss, K.C.; Weber, A.; Wieber, H.; Bailey, D.S.; Barret, O.; Brook, N.H.; Foster, B.; Heath, G.P.; Heath, H.F.; Rodrigues, E.; Scott, J.; Tapper, R.J.; Capua, M.; Schioppa, M.; Susinno, G.; Jeoung, H.Y.; Kim, J.Y.; Lee, J.H.; Lim, I.T.; Ma, K.J.; Pac, M.Y.; Caldwell, A.; Liu, W.; Liu, X.; Mellado, B.; Paganis, S.; Sampson, S.; Schmidke, W.B.; Sciulli, F.; Chwastowski, J.; Eskreys, A.; Figiel, J.; Klimek, K.; Olkiewicz, K.; Piotrzkowski, K.; Przybycien, M.B.; Stopa, P.; Zawiejski, L.; Bednarek, B.; Jelen, K.; Kisielewska, D.; Kowal, A.M.; Kowalski, T.; Przybycien, M.; Rulikowska-Zarebska, E.; Suszycki, L.; Szuba, D.; Kotanski, A.; Bauerdick, L.A.T.; Behrens, U.; Bienlein, J.K.; Borras, K.; Chiochia, V.; Dannheim, D.; Desler, K.; Drews, G.; Fox-Murphy, A.; Fricke, U.; Goebel, F.; Goers, S.; Goettlicher, P.; Graciani, R.; Haas, T.; Hain, W.; Hartner, G.F.; Hebbel, K.; Hillert, S.; Koch, W.; Koetz, U.; Kowalski, H.; Labes, H.; Loehr, B.; Mankel, R.; Martens, J.; Martinez, M.; Milite, M.; Moritz, M.; Notz, D.; Petrucci, M.C.; Polini, A.; Rohde, M.; Savin, A.A.; Schneekloth, U.; Selonke, F.; Sievers, M.; Stonjek, S.; Wolf, G.; Wollmer, U.; Youngman, C.; Zeuner, W.; Coldewey, C.; Lopez-Duran Viani, A.; Meyer, A.; Schlenstedt, S.; Straub, P.B.; Barbagli, G.; Gallo, E.; Parenti, A.; Pelfer, P.G.; Bamberger, A.; Benen, A.; Coppola, N.; Eisenhardt, S.; Markun, P.; Raach, H.; Woelfle, S.; Bussey, P.J.; Bell, M.; Doyle, A.T.; Glasman, C.; Lee, S.W.; Lupi, A.; Macdonald, N.; McCance, G.J.; Saxon, D.H.; Sinclair, L.E.; Skillicorn, I.O.; Waugh, R.; Bohnet, I.; Gendner, N.; Holm, U.; Meyer-Larsen, A.; Salehi, H.; Wick, K.; Carli, T.; Garfagnini, A.; Gialas, I.; Gladilin, L.K.; Kcira, D.; Klanner, R.; Lohrmann, E.; Goncalo, R.; Long, K.R.; Miller, D.B.; Tapper, A.D.; Walker, R.; Cloth, P.; Filges, D.; Ishii, T.; Kuze, M.; Nagano, K.; Tokushuku, K.; Yamada, S.; Yamazaki, Y.; Ahn, S.H.; Lee, S.B.; Park, S.K.; Lim, H.; Son, D.; Barreiro, F.; Garcia, G.; Gonzalez, O.; Labarga, L.; del Peso, J.; Redondo, I.; Terron, J.; Vazquez, M.; Barbi, M.; Corriveau, F.; Hanna, D.S.; Ochs, A.; Padhi, S.; Stairs, D.G.; Wing, M.; Tsurugai, T.; Antonov, A.; Bashkirov, V.; Danilov, M.; Dolgoshein, B.A.; Gladkov, D.; Sosnovtsev, V.; Suchkov, S.; Dementiev, R.K.; Ermolov, P.F.; Golubkov, Yu.A.; Katkov, I.I.; Khein, L.A.; Korotkova, N.A.; Korzhavina, I.A.; Kuzmin, V.A.; Lukina, O.Yu.; Proskuryakov, A.S.; Shcheglova, L.M.; Solomin, A.N.; Vlasov, N.N.; Zotkin, S.A.; Bokel, C.; Botje, M.; Bruemmer, N.; Engelen, J.; Grijpink, S.; Koffeman, E.; Kooijman, P.; Schagen, S.; van Sighem, A.; Tassi, E.; Tiecke, H.; Tuning, N.; Velthuis, J.J.; Vossebeld, J.; Wiggers, L.; de Wolf, E.; Bylsma, B.; Durkin, L.S.; Gilmore, J.; Ginsburg, C.M.; Kim, C.L.; Ling, T.Y.; Boogert, S.; Cooper-Sarkar, A.M.; Devenish, R.C.E.; Grosse-Knetter, J.; Matsushita, T.; Ruske, O.; Sutton, M.R.; Walczak, R.; Bertolin, A.; Brugnera, R.; Carlin, R.; Dal Corso, F.; Dusini, S.; Limentani, S.; Longhin, A.; Posocco, M.; Stanco, L.; Turcato, M.; Adamczyk, L.; Iannotti, L.; Oh, B.Y.; Okrasinski, J.R.; Saull, P.R.B.; Toothacker, W.S.; Whitmore, J.J.; Iga, Y.; D'Agostini, G.; Marini, G.; Nigro, A.; Cormack, C.; Hart, J.C.; McCubbin, N.A.; Shah, T.P.; Epperson, D.; Heusch, C.; Sadrozinski, H.F.-W.; Seiden, A.; Wichmann, R.; Williams, D.C.; Park, I.H.; Pavel, N.; Abramowicz , H.; Dagan, S.; Kananov, S.; Kreisel, A.; Levy, A.; Abe, T.; Fusayasu, T.; Kohno, T.; Umemori, K.; Yamashita, T.; Hamatsu, R.; Hirose, T.; Inuzuka, M.; Kitamura, S.; Matsuzawa, K.; Nishimura, T.; Arneodo, M.; Cartiglia, N.; Cirio, R.; Costa, M.; Ferrero, M.I.; Maselli, S.; Monaco, V.; Peroni, C.; Ruspa, M.; Sacchi, R.; Solano, A.; Staiano, A.; Bailey, D.C.; Fagerstroem, C.-P.; Galea, R.; Koop, T.; Levman, G.M.; Martin, J.F.; Mirea, A.; Sabetfakhri, A.; Butterworth, J.M.; Hayes, M.E.; Heaphy, E.A.; Jones, T.W.; Lane, J.B.; West, B.J.; Ciborowski, J.; Ciesielski, R.; Grzelak, G.; Nowak, R.J.; Pawlak, J.M.; Pawlak, R.; Smalska, B.; Tymieniecka, T.; Wroblewski, A.K.; Zakrzewski, J.A.; Zarnecki, A.F.; Adamus, M.; Gadaj, T.; Deppe, O.; Eisenberg, Y.; Hochman, D.; Karshon, U.; Badgett, W.F.; Chapin, D.; Cross, R.; Foudas, C.; Mattingly, S.; Reeder, D.D.; Smith, W.H.; Vaiciulis, A.; Wildschek, T.; Wodarczyk, M.; Deshpande, A.; Dhawan, S.; Hughes, V.W.; Bhadra, S.; Catterall, C.; Cole, J.E.; Frisken, W.R.; Hall-Wilton, R.; Khakzad, M.; Menary, S.
2001-01-01
Differential cross sections for dijet photoproduction in association with a leading neutron using the reaction e + +p→e + +n+jet+jet+X r have been measured with the ZEUS detector at HERA using an integrated luminosity of 6.4 pb -1 . The fraction of dijet events with a leading neutron in the final state was studied as a function of the jet kinematic variables. The cross sections were measured for jet transverse energies E T jet >6 GeV, neutron energy E n >400 GeV, and neutron production angle θ n <0.8 mrad. The data are broadly consistent with factorization of the lepton and hadron vertices and with a simple one-pion-exchange model
Neutron cross section measurements at n-TOF for ADS related studies
International Nuclear Information System (INIS)
Mastinu, P F; Abbondanno, U; Aerts, G
2006-01-01
A neutron Time-of-Flight facility (n T OF) is available at CERN since 2001. The innovative features of the neutron beam, in particular the high instantaneous flux, the wide energy range, the high resolution and the low background, make this facility unique for measurements of neutron induced reactions relevant to the field of Emerging Nuclear Technologies, as well as to Nuclear Astrophysics and Fundamental Nuclear Physics. The scientific motivations that have led to the construction of this new facility are here presented. The main characteristics of the n T OF neutron beam are described, together with the features of the experimental apparata used for cross-section measurements. The main results of the first measurement campaigns are presented. Preliminary results of capture cross-section measurements of minor actinides, important to ADS project for nuclear waste transmutation, are finally discussed
Measurement of dijet cross sections for events with a leading neutron in photoproduction at HERA
Energy Technology Data Exchange (ETDEWEB)
Breitweg, J.; Chekanov, S.; Derrick, M.; Krakauer, D.; Magill, S.; Musgrave, B.; Pellegrino, A.; Repond, J.; Stanek, R.; Yoshida, R.; Mattingly, M.C.K.; Antonioli, P.; Bari, G.; Basile, M.; Bellagamba, L.; Boscherini, D.; Bruni, A.; Bruni, G.; Cara Romeo, G.; Cifarelli, L.; Cindolo, F.; Contin, A.; Corradi, M.; De Pasquale, S.; Giusti, P.; Iacobucci, G.; Levi, G.; Margotti, A.; Massam, T.; Nania, R.; Palmonari, F.; Pesci, A.; Sartorelli, G.; Zichichi, A.; Amelung, C.; Bornheim, A.; Brock, I.; Coboeken, K.; Crittenden, J.; Deffner, R.; Hartmann, H.; Heinloth, K.; Hilger, E.; Irrgang, P.; Jakob, H.-P.; Kappes, A.; Katz, U.F.; Kerger, R.; Paul, E.; Rautenberg, J.; Schnurbusch, H.; Stifutkin, A.; Tandler, J.; Voss, K.C.; Weber, A.; Wieber, H.; Bailey, D.S.; Barret, O.; Brook, N.H.; Foster, B. E-mail: b.foster@bristol.ac.uk; Heath, G.P.; Heath, H.F.; Rodrigues, E.; Scott, J.; Tapper, R.J.; Capua, M.; Schioppa, M.; Susinno, G.; Jeoung, H.Y.; Kim, J.Y.; Lee, J.H.; Lim, I.T.; Ma, K.J.; Pac, M.Y.; Caldwell, A.; Liu, W.; Liu, X.; Mellado, B.; Paganis, S.; Sampson, S.; Schmidke, W.B.; Sciulli, F.; Chwastowski, J.; Eskreys, A.; Figiel, J.; Klimek, K.; Olkiewicz, K.; Piotrzkowski, K.; Przybycien, M.B.; Stopa, P.; Zawiejski, L.; Bednarek, B.; Jelen, K.; Kisielewska, D.; Kowal, A.M.; Kowalski, T.; Przybycien, M.; Rulikowska-Zarebska, E.; Suszycki, L.; Szuba, D.; Kotanski, A.; Bauerdick, L.A.T.; Behrens, U.; Bienlein, J.K.; Borras, K.; Chiochia, V.; Dannheim, D.; Desler, K.; Drews, G.; Fox-Murphy, A.; Fricke, U.; Goebel, F.; Goers, S.; Goettlicher, P.; Graciani, R.; Haas, T.; Hain, W.; Hartner, G.F.; Hebbel, K.; Hillert, S.; Koch, W.; Koetz, U.; Kowalski, H.; Labes, H.; Loehr, B.; Mankel, R.; Martens, J.; Martinez, M.; Milite, M.; Moritz, M.; Notz, D.; Petrucci, M.C.; Polini, A.; Rohde, M.; Savin, A.A.; Schneekloth, U.; Selonke, F.; Sievers, M.; Stonjek, S.; Wolf, G.; Wollmer, U.; Youngman, C.; Zeuner, W.; Coldewey, C.; Lopez-Duran Viani, A.; Meyer, A.; Schlenstedt, S.[and others
2001-02-26
Differential cross sections for dijet photoproduction in association with a leading neutron using the reaction e{sup +}+p{yields}e{sup +}+n+jet+jet+X{sub r} have been measured with the ZEUS detector at HERA using an integrated luminosity of 6.4 pb{sup -1}. The fraction of dijet events with a leading neutron in the final state was studied as a function of the jet kinematic variables. The cross sections were measured for jet transverse energies E{sub T}{sup jet}>6 GeV, neutron energy E{sub n}>400 GeV, and neutron production angle {theta}{sub n}<0.8 mrad. The data are broadly consistent with factorization of the lepton and hadron vertices and with a simple one-pion-exchange model.
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
Neutron cross section measurements at n-TOF for ADS related studies
Mastinu, P F; Aerts, G; Alvarez, H; Alvarez-Velarde, F; Andriamonje, Samuel A; Andrzejewski, J; Assimakopoulos, P A; Audouin, L; Badurek, G; Bustreo, N; Aumann, P; Beva, F; Berthoumieux, E; Calviño, F; Cano-Ott, D; Capote, R; Carillo de Albornoz, A; Cennini, P; Chepel, V; Chiaveri, Enrico; Colonna, N; Cortés, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillmann, I; Dolfini, R; Domingo-Pardo, C; Dridi, W; Durán, I; Eleftheriadis, C; Segura, M E; Ferrant, L; Ferrari, A; Ferreira-Marques, R; itzpatrick, L; Frais-Kölbl, H; Fujii, K; Furman, W; Guerrero, C; Gonçalves, I; Gallino, R; González-Romero, E M; Goverdovski, A; Gramegna, F; Griesmayer, E; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martínez, A; Igashira, M; Isaev, S; Jericha, E; Kadi, Y; Käppeler, F K; Karamanis, D; Karadimos, D; Kerveno, M; Ketlerov, V; Köhler, P; Konovalov, V; Kossionides, E; Krticka, M; Lamboudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marques, L; Marrone, S; Massimi, C; Mengoni, A; Milazzo, P M; Moreau, C; Mosconi, M; Neves, F; Oberhummer, Heinz; O'Brien, S; Oshima, M; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Plag, R; Plompen, A; Plukis, A; Poch, A; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, Carlo; Rudolf, G; Rullhusen, P; Salgado, J; Sarchiapone, L; Savvidis, I; Stéphan, C; Tagliente, G; Taín, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarín, D; Vincente, M C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescherand, M; Wisshak, K
2006-01-01
A neutron Time-of-Flight facility (n_TOF) is available at CERN since 2001. The innovative features of the neutron beam, in particular the high instantaneous flux, the wide energy range, the high resolution and the low background, make this facility unique for measurements of neutron induced reactions relevant to the field of Emerging Nuclear Technologies, as well as to Nuclear Astrophysics and Fundamental Nuclear Physics. The scientific motivations that have led to the construction of this new facility are here presented. The main characteristics of the n_TOF neutron beam are described, together with the features of the experimental apparata used for cross-section measurements. The main results of the first measurement campaigns are presented. Preliminary results of capture cross-section measurements of minor actinides, important to ADS project for nuclear waste transmutation, are finally discussed.
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
International Nuclear Information System (INIS)
Brenner, D.J.; Prael, R.E.; Little, R.C.
1987-01-01
Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs
EPICS V4 Evaluation for SNS Neutron Data
Energy Technology Data Exchange (ETDEWEB)
Kasemir, Kay [ORNL; Pearson, Matthew R [ORNL; Guyotte, Greg S [ORNL
2015-01-01
Version 4 of the Experimental Physics and Industrial Control System (EPICS) toolkit allows defining application-specific structured data types (pvData) and offers a network protocol for their efficient exchange (pvAccess). We evaluated V4 for the transport of neutron events from the detectors of the Spallation Neutron Source (SNS) to data acquisition and experiment monitoring systems. This includes the comparison of possible data structures, performance tests, and experience using V4 in production on a beam line.
Semiclassical model of cross section for fast neutrons
International Nuclear Information System (INIS)
Rosato, A.; D'Oliveira, A.A.
1977-01-01
A study for main aspects of fast neutron scattering is presented and, a semiclassical approximation applying to several pratic cases is described. The obtained results are compared with experimental data for deformed nuclei, and, with theoretical data based on optical model without treatment of deformations. (M.C.K.) [pt
Neutron total cross section measurements in the energy region from 47 keV to 20 MeV
International Nuclear Information System (INIS)
Poenitz, W.P.; Whalen, J.F.
1983-05-01
Neutron total cross sections were measured for 26 elements. Data were obtained in the energy range from 47 keV to 20 MeV for 11 elements in the range of light-mass fission products. Previously reported measurements for eight heavy and actinide isotopes were extended to 20 MeV. Data were also obtained for Cu (47 keV to 1.4 MeV) and for Sc, Zn, Nd, Hf, and Pt (1.8 to 20 MeV). The present work is part of a continuing effort to provide accurate neutron total cross sections for evaluations and for optical-model parameteriztions. The latter are required for the derivation of other nuclear-data information of importance to applied programs. 37 references
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.)
International Nuclear Information System (INIS)
Takaki, S.; Kondo, K.; Shido, S.; Miyamaru, H.; Murata, I.; Ochiai, Kentaro; Nishitani, Takeo
2006-01-01
Angle-correlated differential cross-section for 9 Be(n,2n) reaction has been measured with the coincidence detection technique and a pencil-beam DT neutron source at FNS, JAEA. Energy spectra of two emitted neutrons were obtained for azimuthal and polar direction independently. It was made clear from the experiment that there are noise signals caused by inter-detector scattering. The ratio of the inter-detector scattering components in the detected signals was estimated by MCNP calculation to correct the measured result. By considering the inter-detector scattering components, the total 9 Be(n,2n) reaction cross-section agreed with the evaluated nuclear data within the experimental error. (author)
Removal cross section for 14 MeV neutrons in constructional materials
International Nuclear Information System (INIS)
Vasvary, L.; Divos, F.; Peto, G.; Csikai, J.; Mumba, N.K.
1986-01-01
Using flight time difference the direct and scattered neutrons and gammas produced in the target head and samples were separated. With this method the attenuation of primary neutrons and gammas originating from the target head has been studied. Thickness dependence of the secondary gamma yield from extended samples of Al, Fe, Pb, paraffin and reinforced concrete was also measured. Results indicate a geometry dependence of the removal cross sections. (author)
TEMPEST-2, Thermalization Program for Neutron Spectra and Multigroup Cross-Sections
International Nuclear Information System (INIS)
Gowins, G.
1984-01-01
Description of problem or function: TEMPEST2 is a neutron thermalization program based upon the Wigner-Wilkins approximation for light moderators and the Wilkins approximation for heavy moderators. A Maxwellian distribution may also be used. The model used may be selected as a function of energy. The second-order differential equations are integrated directly rather than transformed to the Riccati equation. The program provides microscopic and macroscopic cross-section averages over the thermal neutron spectrum
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
Secondary neutron double differential cross sections from 209Bi at 14.2 MeV
International Nuclear Information System (INIS)
Shen Guanren; Xia Haihong; Tang Hongqing
1992-01-01
The secondary neutron double differential cross sections from 209 Bi at 14.2 MeV have been measured at 6 angles between 25 degree-150 degree using an associated particle TOF spectrometer. Flight path was 2.7 m. The neutron detector was biased at 1.3 MeV. The time resolution was about 1.2 ns. The data were compared with existing data and theoretical calculated results. Good agreement is achieved
Measurement of neutron and gamma-ray production double differential cross section at KEK
International Nuclear Information System (INIS)
Ishibashi, Kenji
1995-01-01
High energy nuclear radiations were measured for 0.8-3.0 GeV proton induced reactions at KEK. The measurement was carried out to overcome the problems arising from the use of secondary beam line of a quite low incident beam intensity. Digital pulse shape discrimination method was applicable to separation between high energy neutrons and gamma-rays. By the use of a number of scintillators, cross sections were obtained for production of neutrons and gamma-rays. (author)
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
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
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.
A Neutron Sensitive Microchannel Plate Detector with Cross Delay Line Readout
International Nuclear Information System (INIS)
Berry, Kevin D.; Bilheux, Hassina Z.; Crow, Lowell; Diawara, Yacouba; Feller, W. Bruce; Iverson, Erik B.; Martin, Adrian; Robertson, J. Lee
2012-01-01
Microchannel plates containing neutron absorbing elements such as boron and gadolinium in the bulk glass are used as the sensing element in high spatial resolution, high rate neutron imaging systems. In this paper we describe one such device, using both 10 B and natural Gd, which employs cross delay line signal readout, with time-of-flight capability. This detector has a measured spatial resolution under 40 m FWHM, thermal neutron efficiency of 19%, and has recorded rates in excess of 500 kHz. A physical and functional description is presented, followed by a discussion of measurements of detector performance and a brief survey of some practical applications.
Measurements of integral cross sections in the californium-252 fission neutron spectrum
International Nuclear Information System (INIS)
Alberts, W.G.; Guenther, E.; Matzke, M.; Rassl, G.
1977-01-01
In a low-scattering arrangement cross sections averaged over the californium-252 spontaneous fission neutron spectrum were measured. The reactions 27 Al(n,α) 46 Ti, 47 Ti, 48 Ti(n,p), 54 Fe, 56 Fe(n,p), 58 Ni(n,p), 64 Zn(n,p), 115 In(n,n') were studied in order to obtain a consistent set of threshold detectors used in fast neutron flux density measurements. Overall uncertainties between 2 and 2.5% could be achieved; corrections due to neutron scattering in source and samples are discussed
Neutron scattering differential cross sections of carbon and bismuth at 37 MeV
International Nuclear Information System (INIS)
Zhou Zuying; Tang Hongqing; Qi Bujia; Zhou Chenwei; Du Yanfeng; Xia Haihong; Walter, R.L.; Tornow, W.; Howell, C.; Braun, R.; Roper, C.; Chen Zemin; Chen Zhengpeng; Chen Yingtang
1997-01-01
Elastic differential cross sections of 37 MeV neutrons scattered from carbon and bismuth were measured in the angular range 11 to 160 degrees by means of the multi-detector TOF facility. The 37 MeV neutrons were produced via the T(d,n) 4 He reaction in a tritium gas target. The pulsed 20 MeV deuteron beam was provided by the HI-13 tandem accelerator. The angular distribution of scattered neutrons from carbon and bismuth were measured in the angular range 11 degree to 145 degree and 11 degree to 160 degree respectively in steps of about 3 degree
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)
Surrogate Measurements of Actinide (n,2n) Cross Sections with NeutronSTARS
Energy Technology Data Exchange (ETDEWEB)
Casperson, R. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hughes, R. O. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Akindele, O. A. [Univ. of California, Berkeley, CA (United States); Koglin, J. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wang, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Tamashiro, A. [Oregon State Univ., Corvallis, OR (United States)
2016-09-27
Directly measuring (n,2n) cross sections on short-lived actinides presents a number of experimental challenges. The surrogate reaction technique is an experimental method for measuring cross sections on short-lived isotopes, and it provides a unique solution for measuring (n,2n) cross sections. This technique involves measuring a charged-particle reaction cross section, where the reaction populates the same compound nucleus as the reaction of interest. To perform these surrogate (n,2n) cross section measurements, a silicon telescope array has been placed along a beam line at the Texas A&M University Cyclotron Institute, which is surrounded by a large tank of gadolinium-doped liquid scintillator, which acts as a neutron detector. The combination of the charge-particle and neutron-detector arrays is referred to as NeutronSTARS. In the analysis procedure for calculating the (n,2n) cross section, the neutron detection efficiency and time structure plays an important role. Due to the lack of availability of isotropic, mono-energetic neutron sources, modeling is an important component in establishing this efficiency and time structure. This report describes the NeutronSTARS array, which was designed and commissioned during this project. It also describes the surrogate reaction technique, specifically referencing a ^{235}U(n,2n) commissioning measurement that was fielded during the past year. Advanced multiplicity analysis techniques have been developed for this work, which should allow for efficient analysis of ^{241}Pu(n,2n) and ^{239}Pu(n,2n) cross section measurements
Energy Technology Data Exchange (ETDEWEB)
Lan, Changlin [School of Nuclear Science & Technology, Lanzhou University, Lanzhou 730000 (China); Ruan, Xichao; Chen, Guochang; Nie, Yangbo; Huang, Hanxiong; Bao, Jie; Zhou, Zuying; Tang, Hongqing [Department of Nuclear Physics, China Institute of Atomic Energy, Beijing 102413 (China); Kong, Xiangzhong; Peng, Meng [School of Nuclear Science & Technology, Lanzhou University, Lanzhou 730000 (China)
2016-05-15
The neutron emission double-differential cross sections (DDX) of {sup 9}Be was measured at an incident neutron energy of 21.65 MeV, using the multi-detector fast neutron time-of-flight (TOF) spectrometer on HI-13 Tandem Accelerator at the China Institute of Atomic Energy (CIAE). The data were deduced by comparing the measured TOF spectra with the calculated ones using a realistic Monte-Carlo simulation. The DDX were normalized to n–p scattering cross sections which are a neutron scattering standard. The results of the elastic scattering angular distributions (DX) and the secondary neutron emission DDX at 25 different angles from 15 deg to 145 deg were presented. Meanwhile, a theoretical model based on the unified Hauser-Feshbach and exciton model for light nuclei was used to describe the double-differential cross sections of n+{sup 9}Be, and the theoretical calculation results were compared with the measured cross sections.
Holmes, Jesse Curtis
established that depends on uncertainties in the physics models and methodology employed to produce the DOS. Through Monte Carlo sampling of perturbations from the reference phonon spectrum, an S(alpha, beta) covariance matrix may be generated. In this work, density functional theory and lattice dynamics in the harmonic approximation are used to calculate the phonon DOS for hexagonal crystalline graphite. This form of graphite is used as an example material for the purpose of demonstrating procedures for analyzing, calculating and processing thermal neutron inelastic scattering uncertainty information. Several sources of uncertainty in thermal neutron inelastic scattering calculations are examined, including sources which cannot be directly characterized through a description of the phonon DOS uncertainty, and their impacts are evaluated. Covariances for hexagonal crystalline graphite S(alpha, beta) data are quantified by coupling the standard methodology of LEAPR with a Monte Carlo sampling process. The mechanics of efficiently representing and processing this covariance information is also examined. Finally, with appropriate sensitivity information, it is shown that an S(alpha, beta) covariance matrix can be propagated to generate covariance data for integrated cross sections, secondary energy distributions, and coupled energy-angle distributions. This approach enables a complete description of thermal neutron inelastic scattering cross section uncertainties which may be employed to improve the simulation of nuclear systems.
International Nuclear Information System (INIS)
Yanagisawa, Hiroshi; Sono, Hiroki
2003-06-01
For the examination of neutronic safety design of forthcoming experimental core configurations in the Static Experiment Critical Facility (STACY), neutronic characteristics of 80-cm-diameter cylindrical cores fueled with 6% enriched uranyl nitrate solution have been evaluated by computational analyses. In the analyses, the latest nuclear data library, JENDL-3.3, was used as neutron cross section data. The neutron diffusion and transport calculations were performed using a diffusion code, CITATION, in the SRAC code system and a continuous-energy Monte Carlo code, MVP. Critical level heights of the cores were obtained using such parameters as uranium concentration (up to 500 gU/l), free nitric acid concentration (up to 8 mol/l), and concentration of soluble neutron poisons, gadolinium and boron. It has been confirmed from the evaluation that all critical cores comply with safety criteria required in the STACY operation concerning excess reactivity, reactivity addition rates and shutdown margins by safety rods. (author)
Some remarks on the neutron elastic- and enelastic-scattering cross sections of palladium
International Nuclear Information System (INIS)
Chiba, S.; Guenther, P.T.; Smith, A.B.
1989-05-01
The cross sections for the elastic-scattering of 5.9, 7.1 and 8.0 MeV neutrons from elemental palladium were measured at forty scattering angles distributed between ∼15/degree/ and 160/degree/. The inelastic-scattering cross sections for the excitation of palladium levels at energies of 260 keV to 560 keV were measured with high resolution at the same energies, and at a scattering angle of 80/degree/. The experimental results were combined with lower-energy values previously obtained by this group to provide a comprehensive data base extending from near the inelastic-scattering threshold to 8 MeV. That data base was interpreted in terms of a coupled-channel model, including the inelastic excitation of one- and two-phonon vibrational levels of the even isotopes of palladium. It was concluded that the palladium inelastic-scattering cross section, at the low energies of interest in assessment of fast-fission-reactor performance, are large (∼50% greater than given in widely used evaluated fission-product data files). They primarily involve compound-nucleus processes, with only a small direct-reaction component attributable to the excitation of the one-phonon, 2 + , vibrational levels of the even isotopes of palladium. 24 refs., 6 figs
Evaluation of neutron nuclear data of natural calcium for CENDL-1
International Nuclear Information System (INIS)
Tang Gouyou; Bao Shanglian; Shi Zhaomin; Cao Wentian
1989-03-01
Neutron nuclear data of natural calcium for CENDL-1 has been evaluated in the energy region from 10 -5 ev to 20 MeV. Evaluated quantities are the total, non-elastic scattering, elastic and inelastic scattering, radiation capture, (n,p), (n,t), (n,2n), (n,a) reaction cross sections and the angular distributions of elastic and inelastic cross sections. Some of the data were calculated with the program AUJP based on the optical model and the program MUP2 based on the Hauser-Feshbach model and pre-equilibrium evaporation model. (author). 34 refs, 8 figs, 3 tabs
Fast-neutron total and scattering cross sections of sup 58 Ni and nuclear models
Energy Technology Data Exchange (ETDEWEB)
Smith, A.B.; Guenther, P.T.; Whalen, J.F. (Argonne National Lab., IL (United States)); Chiba, S. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)
1991-07-01
The neutron total cross sections of {sup 58}Ni were measured from {approx} 1 to > 10 MeV using white-source techniques. Differential neutron elastic-scattering cross sections were measured from {approx} 4.5 to 10 MeV at {approx} 0.5 MeV intervals with {ge} 75 differential values per distribution. Differential neutron inelastic-scattering cross sections were measured, corresponding to fourteen levels with excitations up to 4.8 MeV. The measured results, combined with relevant values available in the literature, were interpreted in terms of optical-statistical and coupled-channels model using both vibrational and rotational coupling schemes. The physical implications of the experimental results nd their interpretation are discussed in the contexts of optical-statistical, dispersive-optical, and coupled-channels models. 61 refs.
International Nuclear Information System (INIS)
Czubek, J.A.; Burda, J.; Drozdowicz, K.; Igielski, A.; Kowalik, W.; Krynicka-Drozdowicz, E.; Woznicka, U.
1986-03-01
Preparation of rock samples for the measurement of the thermal neutron macroscopic absorption cross-section in small cylindrical two-region systems by a pulsed technique is presented. Requirements which should be fulfilled during the preparation of the samples due to physical assumptions of the method are given. A cylindrical vessel is filled with crushed rock and saturated with a medium strongly absorbing thermal neutrons. Water solutions of boric acid of well-known macroscopic absorption cross-section are used. Mass contributions of the components in the sample are specified. This is necessary for the calculation of the thermal neutron macroscopic absorption cross-section of the rock matrix. The conditions necessary for assuring the required accuracy of the measurement are given and the detailed procedure of preparation of the rock sample is described. (author)
Amino acids analysis by total neutron cross-sections determinations: part V
International Nuclear Information System (INIS)
Voi, Dante L.; Ferreira, Francisco de O.; Rocha, Helio F. da
2013-01-01
Total neutron cross-sections of twenty essential and non-essential amino acids to human were determined using crystal spectrometer installed on the Argonauta reactor of IEN (Instituto de Engenharia Nuclear (CNEN-RJ) and compared with data generated by parceling and grouping methodologies developed at this institution. For each amino acid was calculated the respective neutron cross-section by molecular structure, conformation and chemistry analysis. The results obtained for eighteen of twenty amino acids confirm the specifications and product formulations indicated by manufactures. These initial results allow to build a neutron cross-sections database as part of quality control of the amino supplied to hospitals for production of nutriments for parenteral or enteral formulations used in critical patients dependent on artificial feed, and for application in future studies of structure and dynamics for more complex molecules, including proteins, enzymes, fatty acids, membranes, organelles and other cell components. (author)
Neutron capture cross section of $^{90}$Zr Bottleneck in the s-process reaction flow
Tagliente, G; Milazzo, P M; Moreau, C; Aerts, G; Abbondanno, U; Alvarez, H; Alvarez-Velarde, F; Andriamonje, Samuel A; Andrzejewski, J; Assimakopoulos, Panayiotis; Audouin, L; Badurek, G; Baumann, P; Bečvář, F; Berthoumieux, E; Bisterzo, S; Calviño, F; Calviani, M; Cano-Ott, D; Capote, R; Carrapiço, C; Cennini, P; Chepel, V; Chiaveri, Enrico; Colonna, N; Cortés, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillman, I; Domingo-Pardo, C; Dridi, W; Durán, I; Eleftheriadis, C; Embid-Segura, M; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Furman, W; Gallino, R; Gonçalves, I; Gonzalez-Romero, E; Gramegna, F; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martínez, A; Igashira, M; Jericha, E; Käppeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Kerveno, M; Köhler, P; Kossionides, E; Krtička, M; Lamboudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marrone, S; Martínez, T; Massimi, C; Mastinu, P; Mengoni, A; Mosconi, M; Neves, F; Oberhummer, Heinz; O'Brien, S; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M T; Plag, R; Plompen, A; Plukis, A; Poch, A; Praena, J; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rubbia, Carlo; Rudolf, G; Rullhusen, P; Salgado, J; Santos, J; Sarchiapone, L; Savvidis, I; Stéphan, C; Taín, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarín, D; Vincente, M, C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescher, M; Wisshak, K
2008-01-01
The neutron capture cross sections of the Zr isotopes have important implications in nuclear astrophysics and for reactor design. The small cross section of the neutron magic nucleus 90Zr, which accounts for more than 50% of natural zirconium represents one of the key isotopes for the stellar s-process, because it acts as a bottleneck in the neutron capture chain between the Fe seed and the heavier isotopes. The same element, Zr, also is an important component of the structural materials used in traditional and advanced nuclear reactors. The (n,γ) cross section has been measured at CERN, using the n_TOF spallation neutron source. In total, 45 resonances could be resolved in the neutron energy range below 70 keV, 10 being observed for the first time thanks to the high resolution and low backgrounds at n_TOF. On average, the Γγ widths obtained in resonance analyses with the R-matrix code SAMMY were 15% smaller than reported previously. By these results, the accuracy of the Maxwellian averaged cross section f...
Testing of ENDF/B cross section data in the Californium-252 neutron benchmark field
International Nuclear Information System (INIS)
Mannhart, W.
1979-01-01
The fission neutron field of 252 Cf presently represents one of the most well-known neutron benchmark fields. For 13 neutron reactions which are of importance in reactor metrology, measurements of spectrum-averaged cross sections, [sigma], performed in this neutron field were compared with calculated average cross sections. This comparison allows one to draw conclusions as to the quality of different sigma(E) data taken from ENDF/B-IV, from ENDF/B-V, and from recent experiments and used in the calculation of average cross sections. The comparison includes an uncertainty analysis regarding the different uncertainty contributions of [sigma], of sigma(E), and of the spectral distribution of 252 Cf fission neutrons. Additionally, in a few examples, sensitivity studies were carried out. The sensitivity of the spectrum-averaged cross sections to individual characteristics of the sigma(E) data, such as normalization factors or shifts in the energy scale, was investigated. Similarly, the sensitivity of [sigma] to the spectral distribution of 252 Cf was determined. 4 figures, 2 tables