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

International Nuclear Information System (INIS)

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

2014-01-01

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

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.

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)

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

Directory of Open Access Journals (Sweden)

Ducasse Q.

2013-12-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

Hirakawa, Naohiro

2000-06-01

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

Measurements of fusion neutron yields by neutron activation technique: Uncertainty due to the uncertainty on activation cross-sections

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.

Unified description of neutron-, proton- and photon-induced fission cross sections in intermediate energy region

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)

Measurement method of activation cross-sections of reactions producing short-lived nuclei with 14 MeV neutrons

CERN Document Server

Kawade, K; Kasugai, Y; Shibata, M; Iida, T; Takahashi, A; Fukahori, T

2003-01-01

We describe a method for obtaining reliable activation cross-sections in the neutron energy range between 13.4 and 14.9 MeV for the reactions producing short-lived nuclei with half-lives between 0.5 and 30 min. We noted neutron irradiation fields and measured induced activities, including (1) the contribution of scattered low-energy neutrons, (2) the fluctuation of the neutron fluence rate during the irradiation, (3) the true coincidence sum effect, (4) the random coincidence sum effect, (5) the deviation in the measuring position due to finite sample thickness, (6) the self-absorption of the gamma-ray in the sample material and (7) the interference reactions producing the same radionuclides or the ones emitting the gamma-ray with the same energy of interest. The cross-sections can be obtained within a total error of 3.6%, when good counting statistics are achieved, including an error of 3.0% for the standard cross-section of sup 2 sup 7 Al (n, alpha) sup 2 sup 4 Na. We propose here simple methods for measuri...

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)

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

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

Science.gov (United States)

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

2018-03-01

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

Partial neutron capture cross sections of actinides using cold neutron prompt gamma activation analysis

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

14.2 MeV neutron induced U-235 fission cross section measurement

International Nuclear Information System (INIS)

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

1986-01-01

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

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.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Active-interrogation measurements of fast neutrons from induced fission in low-enriched uranium

International Nuclear Information System (INIS)

Dolan, J.L.; Marcath, M.J.; Flaska, M.; Pozzi, S.A.; Chichester, D.L.; Tomanin, A.; Peerani, P.

2014-01-01

A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutrons to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials. -- Highlights: • We studied low-enriched uranium using active-interrogation experiments including a deuterium–tritium neutron generator and an americium–lithium isotopic neutron source. • Liquid scintillators measured induced-fission neutrons from the active-interrogation methods. • Fast-neutron (DT) and thermal-neutron (Am–Li) interrogation resulted in the measurement of trends in uranium mass and 235 U enrichment respectively. • MCNPX-PoliMi, the Monte Carlo transport code, simulated the measured induced-fission neutron trends in the liquid scintillators

Active-interrogation measurements of fast neutrons from induced fission in low-enriched uranium

Energy Technology Data Exchange (ETDEWEB)

Dolan, J.L., E-mail: jldolan@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Marcath, M.J.; Flaska, M.; Pozzi, S.A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Tomanin, A.; Peerani, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Ispra (Italy)

2014-02-21

A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutrons to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials. -- Highlights: • We studied low-enriched uranium using active-interrogation experiments including a deuterium–tritium neutron generator and an americium–lithium isotopic neutron source. • Liquid scintillators measured induced-fission neutrons from the active-interrogation methods. • Fast-neutron (DT) and thermal-neutron (Am–Li) interrogation resulted in the measurement of trends in uranium mass and {sup 235}U enrichment respectively. • MCNPX-PoliMi, the Monte Carlo transport code, simulated the measured induced-fission neutron trends in the liquid scintillators.

The European activation file EAF-3 with neutron activation and transmutation cross-sections

International Nuclear Information System (INIS)

Kopecky, J.; Kamp, H.A.J. van der; Gruppelaar, H.; Nierop, D.

1992-09-01

The work performed to obtain the 3rd version of European Activation File (EAF-3) is described, containing cross-sections for neutron induced reactions (0-20 MeV energy range), mainly for use in fusion reactor technology. The starter file was version EAF-2. The present version contains cross-section data for all target nuclides which have half-lives longer than 0.5 days including up to curium (60 targets). Cross-sections to isomeric states are listed separately and if the isomers have a half-life longer than 0.5 days they are also includes as targets. The EAF-3 contains 729 target nuclides with 12,899 reactions with non-zero cross-sections (>10 -7 b) below 20 MeV. A provisional uncertainty file has been generated for all reactions in a one-energy group structure for threshold reactions and in a two-groups structure for (n, γ) reactions. The error estimates for this file were adopted either from experimental information or from systematics. (author). 42 refs., 1 fig., 8 tabs

Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Calculation of the intermediate energy activation cross section

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-01

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

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

International Nuclear Information System (INIS)

Hirakawa, Naohiro

1997-03-01

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

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.)

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

International Nuclear Information System (INIS)

Wang Dahai.

1990-10-01

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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)

Cross-sections for formation of 99mTc through natRu(n,x) 99mTc reaction induced by neutrons at 13.5 and 14.8 MeV

International Nuclear Information System (INIS)

Luo Junhua; Han Jiuning; Tuo Fei; Kong Xiangzhong; Liu Rong; Jiang Li

2012-01-01

The cross-sections for formation of metastable state of 99 Tc ( 99m Tc, 140.511 keV, 6.01 h) through nat Ru(n,x) 99m Tc reaction induced by 13.5 MeV and 14.8 MeV neutrons were measured. Fast neutrons were produced via the 3 H(d,n) 4 He reaction on the K-400 neutron generator. Induced gamma activities were measured by a high-resolution gamma-ray spectrometer with a high-purity germanium (HpGe) detector. Measurements were corrected for gamma-ray attenuations, dead time and fluctuation of neutron flux. Data for nat Ru(n,x) 99m Tc reaction cross sections are reported to be 9.6±1.5 and 9.2±1.1 mb at 13.5±0.2 and 14.8±0.2 MeV incident neutron energies, respectively. Results were compared with the data by other authors. - Highlights: ► D–T neutron source was used to measure cross sections using activation method. ► 27 Al(n,α) 24 Na was used as the monitor for the measurement. ► Cross-sections for formation of 99m Tc through nat Ru(n,x) 99m Tc were measured. ► Data for nat Ru(n,x) 99m Tc reaction cross sections are reported. ► The results were discussed and compared with experimental data in the literature.

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Directory of Open Access Journals (Sweden)

Tveten G. M.

2013-03-01

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

Measurements of activation cross-sections for the 101Ru(n,p)101Tc reaction for neutrons with energies between 13 and 15 MeV

International Nuclear Information System (INIS)

Junhua Luo; Peking University, Beijing; Jiuning Han; Suhong Ge; Zhenlai Liu; Guihua Sun; Rong Liu; Li Jiang

2013-01-01

In this study, activation cross-sections were measured for the 101 Ru(n,p) 101 Tc reaction at three different neutron energies from 13.5 to 14.8 MeV. The fast neutrons were produced via the 3 H(d,n) 4 He reaction on K-400 neutron generator. Induced gamma activities were measured by a high-resolution gamma-ray spectrometer with high-purity germanium detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. The data for 101 Ru(n,p) 101 Tc reaction cross-sections are reported to be 15.7 ± 2.0, 18.4 ± 2.7 and 22.0 ± 2.4 mb at 13.5 ± 0.2, 14.1 ± 0.2, and 14.8 ± 0.2 MeV incident neutron energies, respectively. Results were compared with the previous works. (author)

Measurement of fission cross-section for the {sup 232}Th(n,f){sup 141}Ba reaction induced by neutrons around 14 MeV

Energy Technology Data Exchange (ETDEWEB)

Lan, Chang-Lin; Fang, Kai-Hong [Lanzhou University, School of Nuclear Science and Technology, Lanzhou, Gansu Province (China); Lanzhou University, Engineering Research Center for Neutron Application, Ministry of Education, Lanzhou, Gansu Province (China); Liu, Shuang-Tong; Lv, Tao; Wang, Qiang; Zhang, Zheng-Wei [Lanzhou University, School of Nuclear Science and Technology, Lanzhou, Gansu Province (China); Lai, Cai-Feng [Chinese Academy of Engineering Physics, Institute of Nuclear Physics and Chemistry, Mianyang, Sichuan Province (China)

2016-11-15

The fission cross-section of the {sup 232}Th(n,f){sup 141}Ba reaction induced by neutrons around 14 MeV was measured precisely with the neutron activation and off-line gamma-ray spectrometric technique. Neutron fluence was monitored on-line using the accompanying α-particles from the {sup 3}H({sup 2}H,n){sup 4}He reaction, whereas the neutron energies were measured by the method of cross-section ratios of {sup 90}Zr(n,2n){sup 89}Zr to {sup 93}Nb(n,2n){sup 92m}Nb reactions. The experimentally determined {sup 232}Th(n,f){sup 141}Ba reaction cross-sections were 12.2 ± 0.4 mb at E{sub n} = 14.1 ± 0.3 MeV, 13.0 ± 0.5 mb at E{sub n} = 14.5 ± 0.3 MeV and 13.3 ± 0.5 mb at E{sub n} = 14.7 ± 0.3 MeV, respectively. (orig.)

Calculated neutron-activation cross sections for E/sub n/ /le/ 100 MeV for a range of accelerator materials

International Nuclear Information System (INIS)

Bozoian, M.; Arthur, E.D.; Perry, R.T.; Wilson, W.B.; Young, P.G.

1988-01-01

Activation problems associated with particle accelerators are commonly dominated by reactions of secondary neutrons produced in reactions of beam particles with accelerator or beam stop materials. Measured values of neutron-activation cross sections above a few MeV are sparse. Calculations with the GNASH code have been made for neutrons incident on all stable nuclides of a range of elements common to accelerator materials. These elements include B, C, N, O, Ne, Mg, Al, Si, P, S, Ar, K, Ca, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Mo, Nd, and Sm. Calculations were made for a grid of incident neutron energies extending to 100 MeV. Cross sections leading to the direct production of as many as 87 activation products for each of 84 target nuclide were tabulated on this grid of neutron energies, each beginning with the threshold for the product nuclide's formation. Multigrouped values of these cross sections have been calculated and are being integrated into the cross-section library of the REAC-2 neutron activation code. Illustrative cross sections are presented. 20 refs., 6 figs., 1 tab

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

Energy Technology Data Exchange (ETDEWEB)

Koegler, Toni Joerg

2016-04-26

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

The cross-section data from neutron activation experiments on niobium in the NPI p-7Li quasi-monoenergetic neutron field

Directory of Open Access Journals (Sweden)

Simakov S.P.

2010-10-01

Full Text Available The reaction of protons on 7Li target produces the high-energy quasi- monoenergetic neutron spectrum with the tail to lower energies. Proton energies of 19.8, 25.1, 27.6, 30.1, 32.6, 35.0 and 37.4 MeV were used to obtain quasi-monoenergetic neutrons with energies of 18, 21.6, 24.8, 27.6, 30.3, 32.9 and 35.6 MeV, respectively. Nb cross-section data for neutron energies higher than 22.5 MeV do not exist in the literature. Nb is the important material for fusion applications (IFMIF as well. The variable-energy proton beam of NPI cyclotron is utilized for the production of neutron field using thin lithium target. The carbon backing serves as the beam stopper. The system permits to produce neutron flux density about 109  n/cm2/s in peak at 30 MeV neutron energy. The niobium foils of 15 mm in diameter and approx. 0.75 g weight were activated. The nuclear spectroscopy methods with HPGe detector technique were used to obtain the activities of produced isotopes. The large set of neutron energies used in the experiment allows us to make the complex study of the cross-section values. The reactions (n,2n, (n,3n, (n,4n, (n,He3, (n,α and (n,2nα are studied. The cross-sections data of the (n,4n and (n,2nα are obtained for the first time. The cross-sections of (n,2n and (n,α reactions for higher neutron energies are strongly influenced by low energy tail of neutron spectra. This effect is discussed. The results are compared with the EAF-2007 library.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-06-15

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

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.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

Absorption and activation techniques in measurements of fast-neutron capture cross sections

International Nuclear Information System (INIS)

Bergqvist, I.

1982-01-01

The absorption and activation methods have been applied for a long time to systematic studies of fast neutron capture cross sections. Both methods are simple in principle but difficult in practice. The simplicity should ensure a wider use of the methods in particular for problems which may be complicated to approach with other methods. The difficulties encountered in absorption measurements are related to multiple scattering and resonance shielding effects. In activation experiments the influence of secondary low-energy neutrons causes the main problems

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)

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

International Nuclear Information System (INIS)

Behrens, J.W.

1977-01-01

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

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

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

Aborption and activation techniques in measurements of fast-neutron capture cross sections

International Nuclear Information System (INIS)

Bergqvist, I.

1982-01-01

The absorption and activation methods have been applied for a long time to systematic studies of fast neutron capture cross sections. Both methods are simple in principle but difficult in practice. The simplicity should ensure a wider use of the methods in particular for problems which may be complicated to approach with other methods The difficulties encountered in absorption measurements are related to multiple scattering and resonance shielding effects. In activation experiments the influence of secondary low-energy neutrons c causes the main problems. (Author)

Actinide neutron-induced fission up to 20 MeV

International Nuclear Information System (INIS)

Maslov, V.M.

2001-01-01

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

Actinide neutron-induced fission up to 20 MeV

Energy Technology Data Exchange (ETDEWEB)

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

2001-12-15

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

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

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

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

The Cross-Section for the Radiative Capture of Thermal Neutrons by Uranium

Energy Technology Data Exchange (ETDEWEB)

Broda, E.

1942-07-01

This report is based on an experiment performed at the Cavendish Laboratory (Cambridge) by E. Broda, J. Guéron and L. Kowarski in July 1942 where the intensity of the beta-activity induced in uranium by thermal neutrons has been compared with that induced in manganese or iodine. Care was taken to avoid losses due to a Szilard-Chalmers effect. The capture cross section of uranium for thermal neutrons is found to amount to (2.78 ±0.1)*10{sup -24} cm{sup 2}, assuming the value 581*10{sup -24} cm{sup 2} for σ{sub B}. (nowak)

Thermal neutron capture and resonance integral cross sections of {sup 45}Sc

Energy Technology Data Exchange (ETDEWEB)

Van Do, Nguyen; Duc Khue, Pham; Tien Thanh, Kim [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Thi Hien, Nguyen [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Guinyun, E-mail: gnkim@knu.ac.kr [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Kwangsoo [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Shin, Sung-Gyun; Cho, Moo-Hyun [Department of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang 790-784 (Korea, Republic of); Lee, Manwoo [Research Center, Dongnam Institute of Radiological and Medical Science, Busan 619-953 (Korea, Republic of)

2015-11-01

The thermal neutron cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been measured relative to that of the {sup 197}Au(n,γ){sup 198}Au reaction by means of the activation method. High-purity natural scandium and gold foils without and with a cadmium cover of 0.5 mm thickness were irradiated with moderated pulsed neutrons produced from the Pohang Neutron Facility (PNF). The induced activities in the activated foils were measured with a high purity germanium (HPGe) detector. In order to improve the accuracy of the experimental results the counting losses caused by the thermal (G{sub th}) and resonance (G{sub epi}) neutron self-shielding, the γ-ray attenuation (F{sub g}) and the true γ-ray coincidence summing effects were made. In addition, the effect of non-ideal epithermal spectrum was also taken into account by determining the neutron spectrum shape factor (α). The thermal neutron cross-section and resonance integral of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been determined relative to the reference values of the {sup 197}Au(n,γ){sup 198}Au reaction, with σ{sub o,Au} = 98.65 ± 0.09 barn and I{sub o,Au} = 1550 ± 28 barn. The present thermal neutron cross section has been determined to be σ{sub o,Sc} = 27.5 ± 0.8 barn. According to the definition of cadmium cut-off energy at 0.55 eV, the present resonance integral cross section has been determined to be I{sub o,Sc} = 12.4 ± 0.7 barn. The present results are compared with literature values and discussed.

Activation cross section and isomeric cross section ratios for the (n ,2 n ) reaction on 153Eu

Science.gov (United States)

Luo, Junhua; Jiang, Li; Li, Suyuan

2017-10-01

The 153Eu(n ,2 n ) m1,m2,g152Eu cross section was measured by means of the activation technique at three neutron energies in the range 13-15 MeV. The quasimonoenergetic neutron beam was formed via the 3H(d ,n ) 4He reaction, in the Pd-300 Neutron Generator at the Chinese Academy of Engineering Physics (CAEP). The activities induced in the reaction products were measured using high-resolution γ-ray spectroscopy. The cross section of the population of the second high-spin (8-) isomeric state was measured along with the reaction cross section populating both the ground (3-) and the first isomeric state (0-). Cross sections were also evaluated theoretically using the numerical code TALYS-1.8, with different level density options at neutron energies varying from the reaction threshold to 20 MeV. Results are discussed and compared with the corresponding literature.

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-Induced Charged Particle Studies at LANSCE

Science.gov (United States)

Lee, Hye Young; Haight, Robert C.

2014-09-01

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

Activation cross section and isomeric cross-section ratio for the (n,2n) reaction on {sup 132,134}Ba

Energy Technology Data Exchange (ETDEWEB)

Luo, Junhua [Hexi Univ., Zhangye (China). School of Physics and Electromechanical Engineering; Hexi Univ., Zhangye (China). Inst. of New Energy; Wu, Chunlei; Jiang, Li [Chinese Academy of Engineering Physics, Mianyang (China). Inst. of Nuclear Physics and Chemistry; Li, Suyuan [Hexi Univ., Zhangye (China). Inst. of New Energy

2017-07-01

Cross sections of the {sup 132}Ba(n,2n){sup 131m,g}Ba and {sup 134}Ba(n,2n){sup 133m,g}Ba reactions and their isomeric cross section ratios σ{sub m}/σ{sub g} have been measured by means of the activation technique at three neutron energies in the range 13-15 MeV. BaCO{sub 3} samples and Nb monitor foils were activated together to determine the reaction cross section and the incident neutron flux. The quasimonoenergetic neutrons beam were produced via the {sup 3}H(d,n){sup 4}He reaction at the Pd-300 Neutron Generator of the Chinese Academy of Engineering Physics (CAEP). The activities induced in the reaction products were measured using high-resolution γ ray spectroscopy. The pure cross section of the ground-state was derived from the absolute cross section of the metastable state and the residual nuclear decay analysis. Cross sections were also evaluated theoretically using the numerical nuclear model code, TALYS-1.8 with different level density options at neutron energies varying from the reaction threshold to 20 MeV. Results are discussed and compared with the corresponding literature.

Neutron-capture-activation cross sections of 9496Zr and 98100Mo at thermal and 30 keV energy

International Nuclear Information System (INIS)

Wyrick, J.M.; Poenitz, W.P.

1982-01-01

Neutron-capture cross sections of 94 96 Zr and 98 100 Mo were measured relative to the standard-capture cross section of gold at thermal and 30 keV neutron energies using the activation technique. The reported values are based upon available decay-scheme information

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-10-01

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

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

EASY-2005.1, European Neutron Activation System

International Nuclear Information System (INIS)

Forrest, R.A.; Sublet, Jean-Christophe

2008-01-01

1 - Description of program or function: The EASY-2005 (European Activation System) consists of a wide range of codes, data and documentation all aimed at satisfying the objective of calculating the response of materials irradiated in a neutron flux. The main difference from the previous version is the upper energy limit which has increased from 20 to 60 MeV. It is designed to investigate both fusion devices and accelerator based materials test facilities that will act as intense sources of high energy neutrons causing significant activation of the surrounding materials. The very general nature of the calculational method and the data libraries means that it is applicable (with some reservations) to all situations (e.g. fission reactors or neutron sources) where materials are exposed to neutrons below 60 MeV. EASY can be divided into two parts: data and code development tools and user tools and data. The former are required to develop the latter, but EASY users only need to be able to use the inventory code FISPACT and be aware of the contents of the EAF library (the data source). The complete EASY package contains the FISPACT-2005 inventory code, the EAF-2005 library, and the EASY User Interface. The activation package EASY-2005 [1] is the result of significant development to extend the upper energy range from 20 to 60 MeV so that it is capable of being used for IFMIF calculations. The EAF-2005 library contains 62,637 reactions, almost five times more than in EAF-2003 (12,617). Such a large increase means that some mistakes may exist in the new file that need to be corrected. Because of these shortcomings and for other reasons explained below, a maintenance release, EAF-2005.1 has been produced. A deuteron-induced cross section library was also included for the first time, and can be used with EASY to enable calculations of the activation due to deuterons [2]. This library is included in the EASY-2005.1 maintenance release as is a new version of the FISPACT code

Resonant neutron-induced atomic displacements

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-01

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

Activation measurements for fast neutrons. Part D. Evaluation of cosmic-ray-induced 63Ni background in copper

International Nuclear Information System (INIS)

Ruehm, Werner; Rugel, Georg; Faestermann, Thomas

2005-01-01

As a result of a joint collaboration between the University of Utah, LLNL, the Technical University Munich and the Ludwig Maximilians University Munich, it became possible to determine A-bomb induced 63 Ni in pure copper samples from Hiroshima beyond a ground range of 1,000 m (see Chapter 9, Part B). The low 63 Ni activities induced in copper samples due to neutrons from the A-bomb explosion at large distance require, however, a careful discussion of the fraction of 63 Ni produced in these samples due to cosmic radiation. In this section, an analysis of the production of 63 Ni in copper samples due to cosmic radiation is performed. Production due to neutrons, protons, stopped muons, and photonuclear reactions is discussed. It is obvious from Figure 1 (Pfennig et al. 1995) that a variety of reactions induced by neutrons, protons, muons and photons can contribute to the production of 63 Ni in copper. The most important of these processes will be discussed here. Since the cross-sections for the production of 63 Ni in copper samples due to fast and stopped muons were not known, they were determined experimentally. (J.P.N.)

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)

Activation measurements of fast neutron radiative capture for 139La

International Nuclear Information System (INIS)

Luo, Junhua; Han, Jiuning; Liu, Rong; Jiang, Li; Liu, Zhenlai; Sun, Guihua; Ge, Suhong

2013-01-01

The neutron capture cross section of the neutron magic isotope 139 La has been measured relative to that of 27 Al by means of the activation method. The fast neutrons were produced via the 3 H(d,n) 4 He reaction on Pd-300 neutron generator. The natural high-purity La 2 O 3 powder was used as target material. Induced gamma activities were measured by a high-resolution gamma-ray spectrometer with high-purity germanium (HPGe) detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. The new values for E n =13.5±0.2, 14.1±0.2, and 14.8±0.2 MeV are found to be 1.30±0.08, 1.15±0.08 and 0.99±0.07 mb, respectively. Results were discussed and compared with some corresponding values found in the literature. - Highlights: ► D–T neutron source was used to measure cross sections using activation method. ► 27 Al(n,α) 24 Na was used as the monitor for the measurement. ► The cross sections for the (n,γ) reactions on neutron magic isotope 139 La have been measured. ► The data for 139 La(n,γ) 140 La reaction are presented. ► The results were compared with previous data and with evaluation data

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)

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Preparation of 3-7 MeV neutron source and preliminary results of activation cross section measurement

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)

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

International Nuclear Information System (INIS)

Wang Dahai; Mehta, M.K.

1988-07-01

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

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

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

A neutron irradiator to perform nuclear activation

International Nuclear Information System (INIS)

Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

2001-01-01

The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

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

Monte Carlo calculation of the cross-section of single event upset induced by 14MeV neutrons

International Nuclear Information System (INIS)

Li, H.; Deng, J.Y.; Chang, D.M.

2005-01-01

High-density static random access memory may experience single event upsets (SEU) in neutron environments. We present a new method to calculate the SEU cross-section. Our method is based on explicit generation and transport of the secondary reaction products and detailed accounting for energy loss by ionization. Instead of simulating the behavior of the circuit, we use the Monte Carlo method to simulate the process of energy deposition in sensitive volumes. Thus, we do not need to know details about the circuit. We only need a reasonable guess for the size of the sensitive volumes. In the Monte Carlo simulation, the cross-section of SEU induced by 14MeV neutrons is calculated. We can see that the Monte Carlo simulation not only can provide a new method to calculate SEU cross-section, but also can give a detailed description about random process of the SEU

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

Deuteron and neutron induced activation in the Eveda accelerator materials: implications for the accelerator maintenance

Energy Technology Data Exchange (ETDEWEB)

Garcia, M.; Sanz, J.; Garcia, N.; Cabellos, O. [Madrid Univ. Politecnica, C/ Jose Gutierrez Abascal, lnstituto de Fusion Nuclear (Spain); Sauvan, R. [Universidad Nacional de Educacion a Distancia (UNED), Madrid (Spain); Moreno, C.; Sedano, L.A. [CIEMAT-Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, Association Euratom-CIEMAT, Madrid (Spain)

2007-07-01

Full text of publication follows: The IFMIF (International Fusion Materials Irradiation Facility) is an accelerator-based DLi neutron source designed to test fusion reactor candidate materials for high fluence neutrons. Before deciding IFMIF construction, an engineering design and associated experimental data acquisition, defined as EVEDA, has been proposed. Along the EVEDA accelerator, deuteron beam losses collide with the accelerator materials, producing activation and consequent radiations responsible of dose. Calculation of the dose rates in the EVEDA accelerator room is necessary in order to analyze the feasibility for manual maintenance. Dose rates due to the activation produced by the deuteron beam losses interaction with the accelerator materials, will be calculated with the ACAB activation code, using EAF2007 library for deuteron activation cross-sections. Also, dose rates from the activation induced by the neutron source produced by the interaction of deuteron beam losses with the accelerator materials and the deuterium implanted in the structural lattice, will be calculated with the SRIM2006, TMAP7, DROSG2000/NEUYIE, MCNPX and ACAB codes. All calculations will be done for the EVEDA accelerator with the room temperature DTL structure, which is based on copper cavities for the DTL. Some calculations will be done for the superconducting DTL structure, based on niobium cavities for the DTL working at cryogenic temperature. Final analysis will show the dominant mechanisms and major radionuclides contributing to the surface dose rates. (authors)

Neutron-induced photon production in MCNP

International Nuclear Information System (INIS)

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

1983-01-01

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

Neutron-capture cross-section measurements of Xe136 between 0.4 and 14.8 MeV

Science.gov (United States)

Bhike, Megha; Tornow, W.

2014-03-01

Fast-neutron-capture cross-section data on Xe136 have been measured with the activation method between 0.4 and 14.8 MeV. The cross section was found to be of the order of 1 mb at the eleven energies investigated. This result is important to interpret potential neutron-induced backgrounds in the enriched xenon observatory and KamLAND-Zen neutrinoless double-β decay searches that use xenon as both source and detector. A high-pressure sphere filled with Xe136 was irradiated with monoenergetic neutrons produced by the reactions 3H(p ,n)3He, 2H(d ,n)3He, and 3H(d ,n)4He. Indium and gold monitor foils were irradiated simultaneously with the Xe136 to determine the incident neutron flux. The activities of the reaction products were measured with high-resolution γ-ray spectroscopy. The present results are compared to predictions from ENDF/B-VII.1 and TENDL-2012.

Measurement of 14 MeV neutron cross section of {sup 129}I with foil activation method

Energy Technology Data Exchange (ETDEWEB)

Murata, Isao; Nakano, Daisuke; Takahashi, Akito [Osaka Univ., Suita (Japan). Faculty of Engineering

1997-03-01

The {sup 129}I, which is one of the most famous fission products (FPs), is of very important concern from the standpoint of waste transmutation due to its extremely long half life. The accurate reaction cross section data of {sup 129}I induced by 14 MeV neutrons are indispensable when evaluating the performance to transmute it in a fusion reactor. However, there was no available experimental data reported until now. We measured 14 MeV neutron induced reaction cross sections of {sup 129}I to give the reference cross section data for evaluation of transmutation performance and nuclear data, using OKTAVIAN facility of Osaka university, Japan. Since the available amount of {sup 129}I as a sample is quite small, probably less than 1 mg, the foil activation method was adopted in the measurement. The sample was a sealed source of {sup 129}I and the {gamma}-rays from the irradiated sample were measured with a Hp-Ge detector. Several {gamma}-rays peaks which could be expected to be caused by two nuclear reactions of {sup 129}I(n,2n) and {sup 129}I(n,{gamma}) were observed. We confirmed that these peaks corresponded to those of {sup 128}I and {sup 130}I through ascertaining each energy and half life. From the measurement, the cross section of {sup 129}I(n,2n) and the effective production cross section of {sup 130}I produced by the {sup 129}I(n,{gamma}){sup 130}I reaction including the contribution of {sup 129}I(n,{gamma}){sup 130m}I reaction, that were estimated to be 1.1{+-}0.1 b and 0.032{+-}0.003 b, respectively at 14.8 MeV, were obtained with an acceptable accuracy of about 10 %, though the errors caused by the uncertainty of {gamma} decay scheme data still existed. The measured cross sections were compared with the evaluated nuclear data of JENDL-3.2 and ENDF/B-VI. For the {sup 129}I(n,2n) reaction, the evaluations overestimate the cross section by 30-40 %, while for the {sup 129}I(n,{gamma}) reaction, the evaluations underestimate by at least one order of magnitude

Secondary standards (non-activation) for neutron data measurements above 20 MeV

International Nuclear Information System (INIS)

Haight, R.C.

1991-01-01

In addition to H(n,p) scattering and 235,238 U(n,f) reactions, secondary standards for neutron flux determination may be useful for neutron energies above 20 MeV. For experiments where gamma rays are detected, reference gamma-ray production cross sections are relevant. For neutron-induced charged particle production, standard (n,p) and (n,alpha) cross sections would be helpful. Total cross section standards would serve to check the accuracy of these measurements. These secondary standards are desirable because they can be used with the same detector systems employed in measuring the quantities of interest. Uncertainties due to detector efficiency, geometrical effects, timing and length of flight paths can therefore be significantly reduced. Several secondary standards that do not depend on activation techniques are proposed. 14 refs

Code ACTIVE for calculation of the transmutation, induced activity and decay heat in neutron irradiation

International Nuclear Information System (INIS)

Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.

1976-03-01

The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)

The processed neutron activation cross-section data files of the FENDL project. Summary documentation

International Nuclear Information System (INIS)

Ganesan, S.; Pashchenko, A.B.; Lemmle, H.D.; Mann, F.M.

1994-01-01

This document summarises a neutron activation cross-section database which has been processed in two formats for input to MCNP Monte Carlo codes and to REAC transmutation codes. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command. (author)

Measurement of neutron activation cross-sections for elements Co, Ni, Y, Nb, Tm and Au between 12 and 20 MeV

International Nuclear Information System (INIS)

Iwasaki, S.; Matsuyama, S.; Ohkubo, T.; Fukuda, H.; Sakuma, M.; Kitamura, M.; Odano, N.

1995-01-01

Neutron activation cross-sections for cobalt, nickel, yttrium, niobium, thulium and gold have been measured in the neutron energies from 12 to 20 MeV with the reference cross section of NEA 93 Nb(n,2n) 92m Nb at Tohoku Dynamitron Facility. (author)

Table of nuclear reactions and subsequent radioactive dacays induced by 14-MeV neutrons

International Nuclear Information System (INIS)

Tsukada, Kineo

1977-09-01

Compilation of the data on nuclear reactions and subsequent radioactive decays induced by 14-MeV neutrons is presented in tabular form for most of the isotopes available in nature and for some of the artificially-produced isotopes, including the following items: Nuclide (isotopic abundance), type of nuclear reaction, reaction Q-value, reaction product, type of decay, decay Q-value, half-life of reaction product, decay product, maximum reaction cross section, neutron energy for maximum cross section, reaction cross section for 14 MeV neutrons, saturated radioactivity induced by irradiation of a neutron flux of 1 n/cm 2 sec for a mol of atoms, and reference for the cross section. The mass number dependence of (n, γ), (n, 2n), (n, p), (n, d), (n, t), (n, 3 He) and (n, α) reaction cross sections for 14-MeV neutrons is given in figures to show general trends of the cross sections

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)

Development Of A Method For Measurement Of Total Neutron Cross Sections Based On The Neutron Transmission Method Using A He-3 Counter On Filtered Neutron Beams At Dalat Research Reactor

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)

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

Fast-neutron-induced fission of 242Pu at nELBE

Directory of Open Access Journals (Sweden)

Kögler Toni

2017-01-01

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

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)

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

Broda, E.

1942-07-01

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

Measurement of thermal neutron cross-section and resonance integral for the 165Ho(n,γ) 166gHo reaction using electron linac-based neutron source

Science.gov (United States)

Nguyen, Van Do; Pham, Duc Khue; Kim, Tien Thanh; Kim, Guinyun; Lee, Manwoo; Kim, Kyung Sook; Kang, Heung-Sik; Cho, Moo-Hyun; Ko, In Soo; Namkung, Won

2011-01-01

The thermal neutron cross-section and the resonance integral of the 165Ho(n,γ) 166gHo reaction have been measured by the activation method using a 197Au(n,γ) 198Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation ( Fg), the thermal neutron self-shielding ( Gth), the resonance neutron self-shielding ( Gepi) effects, and the epithermal neutron spectrum shape factor ( α) were taken into account. The thermal neutron cross-section for the 165Ho(n,γ) 166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the 197Au(n,γ) 198Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the 165Ho(n,γ) 166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the 197Au(n,γ) 198Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1988-01-01

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

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

Mercury mass measurement in fluorescent lamps via neutron activation analysis

International Nuclear Information System (INIS)

Viererbl, L.; Vinš, M.; Lahodová, Z.; Fuksa, A.; Kučera, J.; Koleška, M.; Voljanskij, A.

2015-01-01

Mercury is an essential component of fluorescent lamps. Not all fluorescent lamps are recycled, resulting in contamination of the environment with toxic mercury, making measurement of the mercury mass used in fluorescent lamps important. Mercury mass measurement of lamps via instrumental neutron activation analysis (NAA) was tested under various conditions in the LVR-15 research reactor. Fluorescent lamps were irradiated in different positions in vertical irradiation channels and a horizontal channel in neutron fields with total fluence rates from 3×10 8 cm −2 s −1 to 10 14 cm −2 s −1 . The 202 Hg(n,γ) 203 Hg nuclear reaction was used for mercury mass evaluation. Activities of 203 Hg and others induced radionuclides were measured via gamma spectrometry with an HPGe detector at various times after irradiation. Standards containing an Hg 2 Cl 2 compound were used to determine mercury mass. Problems arise from the presence of elements with a large effective cross section in luminescent material (europium, antimony and gadolinium) and glass (boron). The paper describes optimization of the NAA procedure in the LVR-15 research reactor with particular attention to influence of neutron self-absorption in fluorescent lamps. - Highlights: • Mercury is an essential component of fluorescent lamps. • Fluorescent lamps were irradiated in neutron fields in research reactor. • 203 Hg induced radionuclide activity was measured using gamma spectrometry. • Mercury mass in fluorescent lamps can be measured by neutron activation analysis.

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)

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

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)

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

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)

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)

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)

Neutron threshold activation detectors (TAD) for the detection of fissions

Science.gov (United States)

Gozani, Tsahi; Stevenson, John; King, Michael J.

2011-10-01

Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons (≈3 vs. ≈0.01) that are commonly used as indicators for the presence of fissionable materials. The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons. In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron "flash") where fission prompt neutrons could normally not be detected. The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector material. The technique

Neutron threshold activation detectors (TAD) for the detection of fissions

International Nuclear Information System (INIS)

Gozani, Tsahi; Stevenson, John; King, Michael J.

2011-01-01

Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons (∼3 vs. ∼0.01) that are commonly used as indicators for the presence of fissionable materials. The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons. In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron 'flash') where fission prompt neutrons could normally not be detected. The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector material. The technique

Neutron threshold activation detectors (TAD) for the detection of fissions

Energy Technology Data Exchange (ETDEWEB)

Gozani, Tsahi, E-mail: tgozani@rapiscansystems.com [Rapiscan Laboratories, Inc., 520 Almanor Ave., Sunnyvale, CA 94085 (United States); Stevenson, John; King, Michael J. [Rapiscan Laboratories, Inc., 520 Almanor Ave., Sunnyvale, CA 94085 (United States)

2011-10-01

Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons ({approx}3 vs. {approx}0.01) that are commonly used as indicators for the presence of fissionable materials. The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons. In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron 'flash') where fission prompt neutrons could normally not be detected. The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector

Proceedings of a specialists' meeting on neutron activation cross sections for fission and fusion energy applications

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

Measurement of secondary neutron emission double-differential cross sections for {sup 9}Be induced by 21.65 ± 0.07 MeV neutrons

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.

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

Directory of Open Access Journals (Sweden)

Ruskov Ivan

2018-01-01

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

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

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

International Nuclear Information System (INIS)

Jonah, S.A.

2013-01-01

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

Measurement of 0.8 and 1.5 GeV proton induced neutron production cross sections at 0deg

International Nuclear Information System (INIS)

Shigyo, Nobuhiro; Kunieda, Satoshi; Watanabe, Takehito; Ishibashi, Kenji; Satoh, Daiki; Meigo, Shin-ichiro

2004-01-01

Neutron-production double-differential cross sections at 0deg were measured for proton-induced reactions on Fe and Pb targets at 0.8 and 1.5 GeV. The experiment was performed at the π2 beam line of the 12 GeV proton synchrotron in High Energy Accelerator Research Organization (KEK). Neutrons were measured by time-of-flight technique with two different flight path lengths, i.e. 3.5 and 5.0 m at 0.8 and 1.5 GeV, respectively. NE213 liquid organic scintillators 12.7 cm in diameter and 12.7 cm in thickness were set at 0deg as neutron detector. For the improvement of the energy resolution, the scintillator was connected with three Hamamatsu H2431 photomultipliers 5.1 cm in diameter. The neutron detection efficiencies were obtained by the SCINFUL-QMD code. The experimental data were compared with the calculation results of the intranuclear-cascade-evaporation (INC/E) and the quantum-molecular-dynamics (QMD) models. (author)

Development of Cold Neutron Activation Station at HANARO Cold Neutron Source

International Nuclear Information System (INIS)

Sun, G. M.; Hoang, S. M. T.; Moon, J. H.; Chung, Y. S.; Cho, S. J.; Lee, K. H.; Park, B. G.; Choi, H. D.

2012-01-01

A new cold neutron source at the HANARO Research Reactor had been constructed in the framework of a five-year project, and ended in 2009. It has seven neutron guides, among which five guides were already allocated for a number of neutron scattering instruments. A new two-year project to develop a Cold Neutron Activation Station (CONAS) was carried out at the two neutron guides since May 2010, which was supported by the program of the Ministry of Education, Science and Technology, Korea. Fig. 1 shows the location of CONAS. CONAS is a complex facility including several radioanalytical instruments utilizing neutron capture reaction to analyze elements in a sample. It was designed to include three instruments like a CN-PGAA (Cold Neutron - Prompt Gamma Activation Analysis), a CN-NIPS (Cold Neutron - Neutron Induced Pair Spectrometer), and a CN-NDP (Cold Neutron - Neutron-induced prompt charged particle Depth Profiling). Fig. 2 shows the conceptual configuration of the CONAS concrete bioshield and the instruments. CN-PGAA and CN-NIPS measure the gamma-rays promptly emitted from the sample after neutron capture, whereas CN-NDP is a probe to measure the charged particles emitted from the sample surface after neutron capture. For this, we constructed two cold neutron guides called CG1 and CG2B guides from the CNS

Neutron activation analysis in Romania

International Nuclear Information System (INIS)

Apostolescu, St.

1985-01-01

The following basic nuclear facilities are used for neutron activation analysis: a 2000 KW VVR-S Nuclear Reactor, a U-200 Cyclotron, a 30 MeV Betatron, several 14 MeV neutron generators and a king size High Voltage tandem Van de'Graaff accelerator. The main domains of application of the thermal neutron activation analysis are: geology and mining, processing of materials, environment and biology, achaeology. Epithermal neutron activation analysis has been used for determination of uranium and thorium in ores with high Th/U ratios or high rare earth contents. One low energy accelerator, used as 14.1 Mev neutron source, is provided with special equipmen for oxigen and low mass elements determination. An useful alternating way to support fast neutron activation analysis is an accurate theoretical description of the fast neutron induced reactions based on the statistical model (Hauser-Feubach STAPRE code) and the preequilibrium decay geometry dependent model. A gravitational sample changer has been installed at the end of a beam line of the Cyclotron, which enables to perform charged particles activation analysis for protein determination in grains

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

Science.gov (United States)

Pritychenko, B.; Mughabghab, S. F.

2012-12-01

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

Neutron cross section measurements at n-TOF for ADS related studies

Science.gov (United States)

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.

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

Neutron cross section measurements at n-TOF for ADS related studies

CERN Document Server

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.

Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

International Nuclear Information System (INIS)

Chisela, F.

1986-01-01

The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

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 ...

Summary Report from the Consultants' Meeting on International Neutron Cross-Sections Standards: Extending and Updating

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)

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)

Preliminary engineering assessment of the HCLL and HCPB Neutron Activation System

Energy Technology Data Exchange (ETDEWEB)

Calderoni, Pattrick; Leichtle, Dieter [Fusion for Energy, Barcelona, (Spain); Angelone, Maurizio [ENEA, Unita Tecnica Fusione, Frascati, (Italy); Klix, Axel [KIT, Eggenstein-Leopoldshafen, (Germany)

2015-07-01

The Neutron Activation System (NAS) is one of the four types of neutronics sensors considered for the testing of the HCLL and HCPB Test Blanket Module (TBM) in ITER. It measures the absolute neutron flux intensity with information on the neutron spectrum in selected positions of the TBM. The working principle of the NAS is as follows: the system moves small activation probes (capsules) into selected positions in the TBM (irradiation ends) by means of pneumatic transport with pressurized helium gas; the capsules are irradiated for a selected period, depending on their materials composition (several tens of seconds up to the full plasma pulse length); immediately after the irradiation they are extracted and transported to a gamma spectrometer by means of the same pneumatic transport system; the gamma spectrometer determines the induced gamma activity; the neutron flux and neutron fluence is calculated from the measured gamma activity and the known activation cross section of the materials in the activation probe; after the measurement the capsule is sent either to a disposal or storage (for later measurement). This paper summarizes the results of the feasibility assessment of the TBM NAS in the conceptual design phase, including design justification, identification of requirements based on the expected operating conditions in ITER and preliminary engineering assessment of the activation materials, irradiation ends integration in the modules design and the counting station. (authors)

Basic of Neutron NDA

Energy Technology Data Exchange (ETDEWEB)

Trahan, Alexis Chanel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-15

The objectives of this presentation are to introduce the basic physics of neutron production, interactions and detection; identify the processes that generate neutrons; explain the most common neutron mechanism, spontaneous and induced fission and (a,n) reactions; describe the properties of neutron from different sources; recognize advantages of neutron measurements techniques; recognize common neutrons interactions; explain neutron cross section measurements; describe the fundamental of ³He detector function and designs; and differentiate between passive and active assay techniques.

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

ACT-1000. Group activation cross-section library for WWER-1000 type reactors

Energy Technology Data Exchange (ETDEWEB)

Zolotarev, K I; Pashchenko, A B [National Research Centre - A.I. Leipunsky Institute for Physics and Power Engineering, Obninsk (Russian Federation)

2001-10-01

The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10{sup -9}-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (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

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

Experimental Measurement of the Ratio of the Reaction Cross Section (n,2n) for the Natural Mixtures of Lead and Iron Isotopes with 14 MeV Neutrons by the Method of Moderated Neutron-Neutron Coincidences

International Nuclear Information System (INIS)

Panteleev, Ts.Ts.; Penchev, O.I.; Trifonov, A.I.; Troshev, T.M.; Christov, V.I.

1986-01-01

Neutron data are widely applied in nuclei physics and into practice as well. Data obtained by means of different measurement methods are of interest for increasing the accuracy and reliability of the recommended values for the cross sections of neutron interactions, with substance. The activation analysis method gives a possibility to obtain data about cross section interactions of 14 MeV neutrons with the nuclei σ (n,2n), σ( n,p), σ (n,pn), σ (n,α), etc. A serious shortcoming of this measuring method is the necessity of applying express methods of analysis of induced activities - restrictions connected with the life-times of the reaction products. It is also necessary to comply with the requirements for high accuracy in the data about the decay schemes and the absolute intensities of the γ-transitions in the investigated nuclei. The investigations directly measuring the output of the reaction, products from the targets, placed into the neutron beam, do not possess the shortcomings of the activation method but require serious demands toward the detecting apparatuses (ionization chambers, semiconducting and scintillation detectors, proportional counters, etc.). These demands are connected with the heavy background conditions of work at the beams for measurements of whatever partial cross sections. During the experimental measurements of the reactions by neutron emission for registration, it is necessary to slow them down to thermal energies, since only in this energy interval there exist sufficiently effective without threshold detectors. We have elaborated the experimental set-up allowing to measure the neutron multiplicity in an interaction. In the present work we have also used it for a relative measurement the the reaction cross section σ (n,2n) for Pe and Pb nuclei in their natural isotope mixtures. As it is known, this reaction has an important application for the blanket materials of thermonuclear reactors and for tritium fuel regeneration problem. The

Determination of Thermal Neutron Capture Cross Sections Using Cold Neutron Beams at the Budapest PGAA-NIPS Facilities

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)

Fission cross section measurements at intermediate energies

International Nuclear Information System (INIS)

Laptev, Alexander

2005-01-01

The activity in intermediate energy particle induced fission cross-section measurements of Pu, U isotopes, minor actinides and sub-actinides in PNPI of Russia is reviewed. The neutron-induced fission cross-section measurements are under way in the wide energy range of incident neutrons from 0.5 MeV to 200 MeV at the GNEIS facility. In number of experiments at the GNEIS facility, the neutron-induced fission cross sections were obtained for many nuclei. In another group of experiments the proton-induced fission cross-section have been measured for proton energies ranging from 200 to 1000 MeV at 100 MeV intervals using the proton beam of PNPI synchrocyclotron. (author)

Cross sections for d-{sup 3}H neutron interactions with samarium isotopes

Energy Technology Data Exchange (ETDEWEB)

Luo, Junhua; He, Long [Hexi Univ., Zhangye (China). School of Physics and Electromechanical Engineering; Wu, Chunlei; Jiang, Li [Chinese Academy of Engineering Physics, Mianyang (China). Inst. of Nuclear Physics and Chemistry

2016-11-01

The cross sections for (n,x) reactions on samarium isotopes were measured at (d-T) neutron energies of 13.5 and 14.8 MeV with the activation technique. Samples were activated along with Nb and Al monitor foils to determine the incident neutron flux. Theoretical calculations of excitation functions were performed using the nuclear model codes TALYS-1.6 and EMPIRE-3.2 Malta with default parameters, at neutron energies varying from the reaction threshold to 20 MeV. The results were discussed and compared with experimental data found in the literature. At neutron energies 13.5 and 14.8 MeV, the cross sections of the {sup 149}Sm(n,p){sup 149}Pm reaction are reported for the first time. The cross sections of the {sup 150}Sm(n,p){sup 150}Pm, {sup 144}Sm(n,p){sup 144}Pm, {sup 152}Sm(n,α){sup 149}Nd and {sup 144}Sm(n,α){sup 141}Nd reactions at different neutron energies reported in the present work can be added as new data in the nuclear databases.

Generation of laser-induced fast neutron and its application

International Nuclear Information System (INIS)

Cha, Hyung Ki; Lee, S.; Kwon, D.; Nam, S.; Park, S.; Rhee, Y.; Jung, Y.; Lee, K.; Cha, Y.; Kwon, S.; Lim, C.; Han, J.; Park, S.; Chung, C.

2012-04-01

The supply of high-efficiency neutron source is still problematic even though a fast neutron source is being accepted increasingly for industrial applications. Radioisotopes and a neutron tube are typically being used, but their neutron flux, lifetime, and price are the limiting factors for more diverse applications. As ultra high power, short pulse laser technologies have been developed, a neutron source generated via laser induced nuclear reaction comes to the fore. The laser induced neutron source has a high peak flux in comparison to the traditional neutron source and is like a point source with its diameter less than 1 mm. These properties can be utilized effectively for the analysis of pulsed fast neutron activation or the studies of a fast neutron material damage and/or recover. The purpose of R and D here is to develop a robust neutron source with a yield of 107 neutrons/s during 1st R and D stage ('07 ∼ '09) and to construct a stable laser neutron source in longer operation and to demonstrate its usefulness for a neutron activation analysis of explosive materials and a neutron impact analysis of crystalline in the second R and D stage ('10 ∼ '11)

Experimental cross sections for light-charged particle production induced by neutrons with energies between 25 and 65 MeV incident on aluminum

International Nuclear Information System (INIS)

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

2001-01-01

Experimental double-differential cross sections (d 2 σ/dΩdE) for fast neutron-induced proton, deuteron, triton, and alpha-particle production on aluminum are reported, at several incident neutron energies between 25 and 65 MeV, for outgoing particle energies above the experimental energy thresholds. Angular distributions were measured at laboratory angles between 20 deg. and 160 deg. . Reliable extrapolated spectra are derived for very forward (2.5 deg. and 10 deg. ) and very backward angles (170 deg. and 177.5 deg. ). Based on these experimental data, energy-differential (dσ/dE), angle-differential (dσ/dΩ), and total production cross sections (σ T ) are reported for each outgoing particle

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.)

Neutron capture cross sections of $^{70,72,73,74,76}$ Ge at n_TOF EAR-1

CERN Multimedia

We propose to measure the (n;$\\gamma$ ) cross sections of the isotopes $^{70;72;73;74;76}$Ge. Neutron induced reactions on Ge are of importance for the astrophysical slow neutron capture process, which is responsible for forming about half of the overall elemental abundances heavier than Fe. The neutron capture cross section on Ge affects the abundances produced in this process for a number of heavier isotopes up to a mass number of A = 90. Additionally, neutron capture on Ge is of interest for low background experiments involving Ge detectors. Experimental cross section data presently available for Ge (n;$\\gamma$ ) are scarce and cover only a fraction of the neutron energy range of interest. (n;$\\gamma$ ) cross sections will be measured in the full energy range from 25 meV to about 200 keV at n TOF EAR-1.

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

Energy Technology Data Exchange (ETDEWEB)

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

1988-01-01

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

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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)

Measurements of the thermal neutron cross-section and resonance integral for the 108Pd(n,γ)109Pd reaction

Science.gov (United States)

Hien, Nguyen Thi; Kim, Guinyun; Kim, Kwangsoo; Do, Nguyen Van; Khue, Pham Duc; Thanh, Kim Tien; Shin, Sung-Gyun; Cho, Moo-Hyun

2018-06-01

The thermal neutron capture cross-section (σ0) and resonance integral (I0) of the 108Pd(n,γ)109Pd reaction have been measured relative to that of the monitor reaction 197Au(n,γ)198Au. The measurements were carried out using the neutron activation with the cadmium ratio method. Both the samples and monitors were irradiated with and without cadmium cover of 0.5 mm thickness. The induced activities of the reaction products were measured with a well calibrated HPGe γ-ray detector. In order to improve the accuracy of the results, the necessary corrections for the counting losses were made. The thermal neutron capture cross-section and resonance integral of the 108Pd(n,γ)109Pd reaction were determined to be σ0,Pd = 8.68 ± 0.41 barn and I0,Pd = 245.6 ± 24.8 barn, respectively. The obtained results are compared with literature values and discussed.

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

International Nuclear Information System (INIS)

Strohmaier, Brigitte

1995-01-01

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

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.

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

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

Measurements of DT and DD neutron yields by neutron activation on TFTR

International Nuclear Information System (INIS)

Barnes, C.W.; Larson, A.R.; LeMunyan, G.

1994-01-01

A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants. and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma,) accuracy of the measurements: also agreeing are yields from silicon foils using the ACTL library cross-section. While the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n,n) 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

Measurements of DT and DD neutron yields by neutron activation on TFTR

International Nuclear Information System (INIS)

Barnes, C.W.; Larson, A.R.; LeMunyan, G.

1995-03-01

A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants, and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma) accuracy of the measurements; also agreeing are yields from silicon foils using the ACTL library cross-section, while the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n.n') 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

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

International Nuclear Information System (INIS)

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

1976-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1976-07-04

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

Activation cross section data file, (1)

International Nuclear Information System (INIS)

Yamamuro, Nobuhiro; Iijima, Shungo.

1989-09-01

To evaluate the radioisotope productions due to the neutron irradiation in fission of fusion reactors, the data for the activation cross sections ought to be provided. It is planning to file more than 2000 activation cross sections at final. In the current year, the neutron cross sections for 14 elements from Ni to W have been calculated and evaluated in the energy range 10 -5 to 20 MeV. The calculations with a simplified-input nuclear cross section calculation system SINCROS were described, and another method of evaluation which is consistent with the JENDL-3 were also mentioned. The results of cross section calculation are in good agreement with experimental data and they were stored in the file 8, 9 and 10 of ENDF/B format. (author)

Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping

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.

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].

Nondestructive neutron activation analysis of silicon carbide

Energy Technology Data Exchange (ETDEWEB)

Vandergraaf, T. T.; Wikjord, A. G.

1973-10-15

Instrumentel neutron activation analysis was used to determine trace constituents in silicon carbide. Four commercial powders of different origin, an NBS reference material, and a single crystal were characterized. A total of 36 activation species were identified nondestructively by high resolution gamma spectrometry; quantitative results are given for 12 of the more predominant elements. The limitations of the method for certain elements are discussed. Consideration is given to the depression of the neutron flux by impurities with large neutron absorption cross sections. Radiation fields from the various specimens were estimated assuming all radionuclides have reached their saturation activities. (auth)

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.)

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.

Determination of europium content in Li_2SiO_3(Eu) by neutron activation analysis using Am-Be neutron source

International Nuclear Information System (INIS)

Naik, Yeshwant; Tapase, Anant Shamrao; Mhatre, Amol; Datrik, Chandrashekhar; Tawade, Nilesh; Kumar, Umesh; Naik, Haladhara

2016-01-01

Circulardiscs of Li_2SiO_3 doped with europium were prepared and a new activation procedure for the neutron dose estimation in a breeder blanket of fusion reactor is described. The amount of europium in the disc was determined by neutron activation analysis (NAA) using an isotopic neutron source. The average neutron absorption cross section for the reaction was calculated using neutron distribution of the Am-Be source and available neutron absorption cross section data for the "1"5"1Eu(n,γ)"1"5"2"mEu reaction, which was used for estimation of europium in the pallet. The cross section of the elements varies with neutron energy, and the flux of the neutrons in each energy range seen by the nuclei under investigation also varies. Neutron distribution spectrum of the Am-Be source was worked out prior to NAA and the effective fractional flux for the nuclear reaction considered for the flux estimation was also determined. - Highlights: • Lithium meta-silicate is breeder materials for a fusion reactor. • Europium is used for neutron dose estimation in a breeder blanket. • It is important to determine amount of europium in lithium meta-silicate. • Amount of europium in lithium meta-silicate was determined by neutron activation and off-line gamma spectrometry.

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

Science.gov (United States)

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

2018-04-22

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

Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents

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.)

Precise 238U(n,2n)237U reaction cross-section measurements using the activation facility at TUNL

Science.gov (United States)

Krishichayan, Fnu; Bhike, M.; Tornow, W.

2014-09-01

Accurate neutron-induced 238U(n,2n)237U reaction data are required for many practical applications, especially in the field of nuclear energy, including advanced heavy water reactors, where 238U is used as the breeding material to regenerate the fissile material 239Pu. Precise (n,2n) cross-section measurements of 238U are underway at TUNL with mono-energetic neutrons in the 8.0 to 14.0 MeV energy range in steps of 0.25 MeV using the activation technique. After activation of the 0.5 inch diameter and 442 mg 238U foil, the activity of the 208 keV characteristic γ-line is tracked for 6 weeks with a high efficient HPGe clover detector to determine the initial activity needed for the cross-section determination. Results of the cross-section measurements, determined relative to 27Al and 197Au neutron activation monitor foils, and the comparison with theoretical models will be presented during the meeting.

Neutron capture cross section of ^243Am

Science.gov (United States)

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)

Measurement of double differential cross sections of secondary neutrons in the incident energy range 9-13 MeV

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

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

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

Generation of laser-induced fast neutron and its application

International Nuclear Information System (INIS)

Cha, Hyung Ki; Kwon, D. H.; Nam, S. M.

2010-04-01

The supply of high-efficiency neutron source is still problematic even though a fast neutron source is being accepted increasingly for industrial applications. Radioisotopes and a neutron tube are typically being used, but their neutron flux, lifetime, and price are the limiting factors for more diverse applications. As ultra high power, short pulse laser technologies have been developed, a neutron source generated via laser induced nuclear reaction comes to the fore. The laser induced neutron source has a high peak flux in comparison to the traditional neutron source and is like a point source with its diameter less than 1 mm. These properties can be utilized effectively for the analysis of pulsed fast neutron activation or the studies of a fast neutron material damage and/or recover. The purpose of R and D here is to develop a robust neutron source with a yield of 10 7 neutrons/s, and to carry out a preliminary research for application study in the next research stage

Radioactivity induced by neutrons: Enrico Fermi and a thermodynamic approach to radiative capture

Science.gov (United States)

De Gregorio, Alberto

2006-07-01

When Fermi learned that slow neutrons are much more effective than fast ones in inducing radioactivity, he explained this phenomenon by mentioning the well-known scattering cross section between neutrons and protons. At this early stage, he did not refer to the capture cross section by target nuclei. At the same time a thermodynamic approach to neutron-proton capture was being discussed by physicists: neutron capture was interpretated as the reverse of deuteron photodissociation and detailed balance among neutrons, protons, deuterons, and radiation was invoked. This thermodynamic approach might underlie Fermi's early explanation of the great efficiency of slow neutrons. Fermi repeatedly used a thermodynamic approach that had been used in describing some of the physical properties of conductors by Richardson and had been influential in Fermi's youth.

The investigation of fast neutron Threshold Activation Detectors (TAD)

International Nuclear Information System (INIS)

Gozani, T; King, M J; Stevenson, J

2012-01-01

The detection of fast neutrons is usually done by liquid hydrogenous organic scintillators, where the separation between the ever present gamma rays and neutrons is achieved by the pulse shape discrimination (PSD). In many practical situation the detection of fast neutrons has to be carried out while the intense source (be it neutrons, gamma rays or x-rays) that creates these neutrons, for example by the fission process, is present. This source, or ''flash'', usually blinds the neutron detectors and temporarily incapacitates them. By the time the detectors recover the prompt neutron signature does not exist. Thus to overcome the blinding background, one needs to search for processes whereby the desired signature, such as fission neutrons could in some way be measured long after the fission occurred and when the neutron detector is fully recovered from the overload. A new approach was proposed and demonstrated a good sensitivity for the detection of fast neutrons in adverse overload situations where normally it could not be done. A temporal separation of the fission event from the prompt neutrons detection is achieved via the activation process. The main idea, called Threshold Activation Detection (or detector)-TAD, is to find appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation, and then measure the radioactively decaying activation products (typically beta and γ-rays) well after the source pulse has ended. The activation material should possess certain properties: a suitable half-life; an energy threshold below which the numerous source neutrons will not activate it (e.g. about 3 MeV); easily detectable activation products and has a usable cross section for the selected reaction. Ideally the substance would be part of the scintillator. There are several good candidates for TAD. The first one we have selected is based on fluorine. One of the major advantages of this element is the fact that it is a major

The investigation of fast neutron Threshold Activation Detectors (TAD)

Science.gov (United States)

Gozani, T.; King, M. J.; Stevenson, J.

2012-02-01

The detection of fast neutrons is usually done by liquid hydrogenous organic scintillators, where the separation between the ever present gamma rays and neutrons is achieved by the pulse shape discrimination (PSD). In many practical situation the detection of fast neutrons has to be carried out while the intense source (be it neutrons, gamma rays or x-rays) that creates these neutrons, for example by the fission process, is present. This source, or ``flash'', usually blinds the neutron detectors and temporarily incapacitates them. By the time the detectors recover the prompt neutron signature does not exist. Thus to overcome the blinding background, one needs to search for processes whereby the desired signature, such as fission neutrons could in some way be measured long after the fission occurred and when the neutron detector is fully recovered from the overload. A new approach was proposed and demonstrated a good sensitivity for the detection of fast neutrons in adverse overload situations where normally it could not be done. A temporal separation of the fission event from the prompt neutrons detection is achieved via the activation process. The main idea, called Threshold Activation Detection (or detector)-TAD, is to find appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation, and then measure the radioactively decaying activation products (typically beta and γ-rays) well after the source pulse has ended. The activation material should possess certain properties: a suitable half-life; an energy threshold below which the numerous source neutrons will not activate it (e.g. about 3 MeV); easily detectable activation products and has a usable cross section for the selected reaction. Ideally the substance would be part of the scintillator. There are several good candidates for TAD. The first one we have selected is based on fluorine. One of the major advantages of this element is the fact that it is a major

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

Science.gov (United States)

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

2017-09-01

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

Improved activation cross sections for vanadium and titanium

International Nuclear Information System (INIS)

Muir, D.W.; Arthur, E.D.

1983-01-01

Vanadium alloys such as V-20Ti and V-Cr-Ti are attractive candidates for use as structural materials in fusion-reactor blankets. The virtual absence of long-lived activation products in these alloys suggest the possibility of reprocessing on an intermediate time scale. We have employed the modern Hauser-Feshbach nuclear-model code GNASH to calculate cross sections for neutron-activation reactions in 50 V and 51 V, to allow a more accurate assessment of induced radioactivity in vanadium alloys. In addition, cross sections are calculated for the reactions 46 Ti(n,2n) and 45 Ti(n,2n) in order to estimate the production of 44 Ti, a 1.2-MeV gamma-ray source with a half-life of 47 years

Measurement of fission cross section for {sup 232}Th(n,f){sup 131}{sub Z}X (Z = 50, 51, 52, 53) reaction induced by neutrons around 14 MeV

Energy Technology Data Exchange (ETDEWEB)

Lan, Chang-lin; Qiu, Yi-jia; Wang, Qiang; Zhang, Zheng-wei; Zhang, Qian; Tan, Jun-cai; Fang, Kai-hong [Lanzhou University, School of Nuclear Science and Technology, Lanzhou (China); Lai, Cai-feng [China Academy of Engineering Physics, Institute of Nuclear Physics and Chemistry, Mianyang (China)

2017-06-15

The fission cross sections of {sup 232}Th(n,f){sup 131m,g}Sn, {sup 232}Th(n,f){sup 131}Sb, {sup 232}Th(n,f){sup 131m,g}Te, {sup 232}Th(n,f){sup 131}I fission reactions induced by 14 MeV neutrons were measured precisely with the neutron activation technique. The neutron flux was monitored by accompanying α particle in the irradiation and the neutron energies were determined by the cross section ratio of {sup 90}Zr(n,2n){sup 89}Zr to {sup 93}Nb(n,2n){sup 92m}Nb reaction. The values of the cross sections of {sup 232}Th(n,f){sup 131m,g}Sn were analyzed, and the cross sections of {sup 232}Th(n,f){sup 131}Sb were deduced to be 6.5±0.7, 6.3±0.6, 6.1±0.6 mb at 14.1±0.3, 14.5±0.3 and 14.8±0.3 MeV, respectively. The values of the cross sections of {sup 232}Th(n,f){sup 131g}Te were deduced to be 1.8 ± 0.1, 1.5 ± 0.1 and 1.4±0.1 mb at 14.1±0.3, 14.5±0.3 and 14.8±0.3 MeV, respectively. The values of the cross sections of {sup 232}Th(n,f){sup 131}I were given as 1.8±0.2, 1.6±0.2, 1.5±0.1 mb at 14.1±0.3, 14.5±0.3 and 14.8±0.3 MeV, respectively. (orig.)

Measurement of 235U fission spectrum-averaged cross sections and neutron spectrum adjusted with the activation data

International Nuclear Information System (INIS)

Kobayashi, Katsuhei; Kobayashi, Tooru

1992-01-01

The 235 U fission spectrum-averaged cross sections for 13 threshold reactions were measured with the fission plate (27 cm in diameter and 1.1 cm thick) at the heavy water thermal neutron facility of the Kyoto University Reactor. The Monte Carlo code MCNP was applied to check the deviation from the 235 U fission neutron spectrum due to the room-scattered neutrons, and it was found that the resultant spectrum was close to that of 235 U fission neutrons. Supplementally, the relations to derive the absorbed dose rates with the fission plate were also given using the calculated neutron spectra and the neutron Kerma factors. Finally, the present values of the fission spectrum-averaged cross sections were employed to adjust the 235 U fission neutron spectrum with the NEUPAC code. The adjusted spectrum showed a good agreement with the Watt-type fission neutron spectrum. (author)

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

International Nuclear Information System (INIS)

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

2009-12-01

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

ORLIB: a computer code that produces one-energy group, time- and spatially-averaged neutron cross sections

International Nuclear Information System (INIS)

Blink, J.A.; Dye, R.E.; Kimlinger, J.R.

1981-12-01

Calculation of neutron activation of proposed fusion reactors requires a library of neutron-activation cross sections. One such library is ACTL, which is being updated and expanded by Howerton. If the energy-dependent neutron flux is also known as a function of location and time, the buildup and decay of activation products can be calculated. In practice, hand calculation is impractical without energy-averaged cross sections because of the large number of energy groups. A widely used activation computer code, ORIGEN2, also requires energy-averaged cross sections. Accordingly, we wrote the ORLIB code to collapse the ACTL library, using the flux as a weighting function. The ORLIB code runs on the LLNL Cray computer network. We have also modified ORIGEN2 to accept the expanded activation libraries produced by ORLIB

Multi-group transport methods for high-resolution neutron activation analysis

International Nuclear Information System (INIS)

Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

2009-01-01

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

Charged-particle magnetic-quadrupole spectrometer for neutron induced reactions

International Nuclear Information System (INIS)

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

1975-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

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

Novel fast-neutron activation counter for high repetition rate measurements

International Nuclear Information System (INIS)

Mahmood, S.; Springham, S. V.; Zhang, T.; Rawat, R. S.; Tan, T. L.; Krishnan, M.; Beg, F. N.; Lee, S.; Schmidt, H.; Lee, P.

2006-01-01

A fast-neutron beryllium activation counter has been constructed for neutron measurements on a high repetition rate deuterium plasma focus. Beryllium activation is especially suitable for measurements of DD neutron yields. The cross section for the relevant reaction, 9 Be(n,α) 6 He, results in a maximum sensitivity at the characteristic energy of the DD neutrons (∼2.5 MeV) and practically no sensitivity to neutrons with energies 6 He enabled the shot-to-shot neutron yield from the plasma focus to be measured for repetition rates from 0.2 to 3 Hz (and for a range of deuterium gas pressures). With careful analysis, the shot-to-shot yield can be measured up to a maximum repetition rate of 3 Hz, beyond which the pileup of counts from the previous shots reduces the accuracy of the measurements to an unacceptable level. This new beryllium activation counter has been cross-checked against an indium activation counter to obtain absolute neutron yields. At a charging voltage of 12.5 kV (bank energy of 2.2 kJ), the average neutron yield was found to be (7.9±0.7)x10 7 per shot (standard deviation of 4x10 7 ). It was found that activation of the plasma focus construction materials (especially aluminum) must be taken into account

Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping.

Science.gov (United States)

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.

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)

Neutron activation of building materials used in the reactor shield

International Nuclear Information System (INIS)

Hernandez, A.T.; Perez, G.; D'Alessandro, K.

1993-01-01

Cuban concretes and their main components (mineral aggregates and cement) were investigated through long-lived activation products induced by neutrons from a reactor. The multielemental content in the materials studied was obtained by neutron activation analysis in an IBR-2 reactor and gamma activation analysis in an MT-25 microtron from Join Institute of Nuclear Research of Dubna. After irradiation of building materials for 30 years by a neutron flow of unitary density, induced radioactivity was calculated according to experimental data. The comparative evaluation of different concretes aggregates and two types of cement related to the activation properties is discussed

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

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)

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

FY17 Status Report on NEAMS Neutronics Activities

Energy Technology Data Exchange (ETDEWEB)

Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Jung, Y. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Smith, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-09-30

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H) capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.

Modern g,d,p,n-induced activation-transmutation systems

International Nuclear Information System (INIS)

Sublet, J.Ch.

2010-01-01

Document available in abstract form only. Full text of publication follows: The European Activation System (EASY) includes as the source of nuclear data the European Activation File (EAF) and as its engine the FISPACT activation-transmutation code. The latest version of the EAF, EAF-2010, contains cross-section data for gamma-, deuteron- and proton-induced reactions in addition to the traditional neutron-induced data. The main reason for the addition of these data to EAF is to enable activation-transmutation calculations to be performed for even more nuclear facilities, including 'accelerator'-driven devices with incident upper energy limit of 60 or 200 MeV. EAF-2010 has benefited from the generation and maintenance of comprehensive activation files in the past and the development of the processing code SAFEPAQ-II and model code TALYS. TALYS is the source for all gamma-, proton- and deuteron-induced data and a fair share of the neutron-induced data. Cross-section validation exercises against both experimental data and systematic, which were started in 1995, enable a comprehensive assessment of the data. Although EAF-2010 is certainly the most-validated activation neutron cross-section library in the world, currently less than 3% of all the reactions can be compared with experimental information, and even then only for a very limited, and not always application-relevant, energy range. As with EAF-2010, -2003, -2005 and -2007 results of integral experiments have been used to correct, adjust and validate data. This can be done using SAFEPAQ II by inputting the measured effective cross-sections. Validation using integral data has been performed by means of direct comparison with measurements of various materials under relevant particle spectra. A tool has recently been developed which is important now that the libraries contain so much TALYS-calculated data. Statistical analysis of cross-sections (SACS) is used to look for trends in the library data for a particular

Measurements of fission cross-sections and of neutron production rates; Mesures de sections efficaces de fission et du nombre de neutrons prompts emis par fission

Energy Technology Data Exchange (ETDEWEB)

Billaud, P; Clair, C; Gaudin, M; Genin, R; Joly, R; Leroy, J L; Michaudon, A; Ouvry, J; Signarbieux, C; Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

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 {sup 10}B layers, was used as the neutron flux monitor. b) Measurement of the fission cross section of {sup 235}U. We intend to measure the variation of the neutron induced fission cross section of {sup 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 {sup 235}U (20 deposits of 25 cm diameter). The relative fission data were corrected for the neutron spectrum measured with a set of BF{sub 3} proportional counters. c) Mean number {nu} of neutrons emitted in neutron induced fission. We measured the value of {nu} for several fissile isotopes in the case of fission induced by 14 MeV neutrons. The 14 MeV neutrons were produced by D (t, n) {alpha} reaction by means of a 300 kV Cockcroft Walton generator. (author)Fren. [French] a) Mesures de sectionficaces de fission a basse energie. Nous avons mesure et analyse la variation de la section efficace de fission de divers isotopes fissiles pour des neutrons d'energie inferieure a 0,025 eV. Le monochromateur est constitue par un spectrometre a cristal auquel est associe un selecteur mecanique destine a eliminer les diffractions de Bragg d'ordre superieur au premier. Le materiau fissile est contenu dans une chambre a fission sous forme de depots realises par peinture; une chambre d'ionisation a depots minces de B{sub 10

Statistical Model Analysis of (n, α Cross Sections for 4.0-6.5 MeV Neutrons

Directory of Open Access Journals (Sweden)

Khuukhenkhuu G.

2016-01-01

Full Text Available The statistical model based on the Weisskopf-Ewing theory and constant nuclear temperature approximation is used for systematical analysis of the 4.0-6.5 MeV neutron induced (n, α reaction cross sections. The α-clusterization effect was considered in the (n, α cross sections. A certain dependence of the (n, α cross sections on the relative neutron excess parameter of the target nuclei was observed. The systematic regularity of the (n, α cross sections behaviour is useful to estimate the same reaction cross sections for unstable isotopes. The results of our analysis can be used for nuclear astrophysical calculations such as helium burning and possible branching in the s-process.

Microscopic integral cross section measurements in the Be(d,n) neutron spectrum for applications in neutron dosimetry, radiation damage and the production of long-lived radionuclides

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

The CERN n_TOF Facility: Neutron Beams Performances for Cross Section Measurements

CERN Document Server

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.

Neutrino-induced neutral-current reaction cross sections for r-process nuclei

CERN Document Server

Langanke, K

2002-01-01

Neutrino-induced reactions play an important role during and after the r-process, if the latter occurs in an environment with extreme neutrino fluxes such as the neutrino-driven wind model or neutron star mergers. Recently we have evaluated the charged-current neutrino-nucleus cross sections relevant for r-process simulations. We extend our approach here to the neutral-current cross sections. Our tabulation considers neutron-rich nuclei with neutron numbers N=41-135 and charge numbers Z=21-82 and lists total as well as partial neutron spallation cross sections. The calculations have been performed within the random phase approximation considering multipole transitions with J<=3 and both parities. The supernova neutrino spectrum is described by a Fermi-Dirac distribution with various temperature parameters between T=2.8 MeV and T=10 MeV and with the degeneracy parameters alpha=0 and alpha=3.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Differential α-production cross sections of iron and nickel for 4.3 to 14.1 MeV Neutrons

International Nuclear Information System (INIS)

Baba, Mamoru; Ito, Nobuo; Matsuyama, Isamu

1994-01-01

The cross section data for neutron-induced α-production are of prime importance in the evaluation of the radiation damage and nuclear heating in fusion and fast reactors. For the evaluation, energy and angular doubly differential cross sections are also required to calculate primary knock-on atom spectra. However, the experimental (n, xα) data are few and discrepant, therefore, the new experimental data are required urgently to improve the accuracy of the (n, xα) cross section data. The authors have measured the double differential (n, xα) cross sections of Fe and Ni in the neutron energy range of 4.3-14.1 MeV using a specially developed gridded ionization chamber. The present work was undertaken as a part of IAEA Coordinated Research Program for neutron-induced He production cross sections. The gridded ionization chamber and the experimental method were reported previously. Three-signals from the common cathode and two anodes were accumulated as two sets of two-dimensional data. The experimental two-dimensional data for the anode and cathode signals were transformed into the double differential cross sections. The results of the double differential cross sections, angular distributions, angle-integrated spectra in the center of mass system and total α-production cross sections are shown. (K.I.)

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

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

CERN Multimedia

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

Low doses of neutrons induce changes in gene expression

International Nuclear Information System (INIS)

Woloschak, G.E.; Chang-Liu, C.M.; Panozzo, J.; Libertin, C.R.

1993-01-01

Studies were designed to identify genes induced following low-dose neutron but not following γ-ray exposure in fibroblasts. Our past work had shown differences in the expression of β-protein kinase C and c-fos genes, both being induced following γ-ray but not neutron exposure. We have identified two genes that are induced following neutron, but not γ-ray, exposure: Rp-8 (a gene induced by apoptosis) and the long terminal repeat (LTR) of the human immunodeficiency (HIV). Rp-8 mRNA induction was demonstrated in Syrian hamster embryo fibroblasts and was found to be induced in cells exposed to neutrons administered at low (0.5 cGy/min) and at high dose rate (12 cGy/min). The induction of transcription from the LTR of HIV was demonstrated in HeLa cells bearing a transfected construct of the chloramphenicol acetyl transferase (CAT) gene driven by the HIV-LTR promoter. Measures of CAT activity and CAT transcripts following irradiation demonstrated an unresponsiveness to γ rays over a broad range of doses. Twofold induction of the HIV-LTR was detected following neutron exposure (48 cGy) administered at low (0.5 cGy/min) but not high (12 cGy/min) dose rates. Ultraviolet-mediated HIV-LTR induction was inhibited by low-dose-rate neutron exposure

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)

Measurement of the scattering cross section of slow neutrons on liquid parahydrogen from neutron transmission

Science.gov (United States)

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.

Neutron cross-section measurements at the nTOF facility at CERN

CERN Document Server

Colonna, N

2004-01-01

A neutron Time-of-Flight facility (n_TOF) has recently become operative at CERN. 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 n_TOF facility is here described, together with the main features of the experimental apparata used for cross-section measurements. The results of the first measurement campaign, which have confirmed the innovative aspects of the facility, are presented. The measurement plan of the n_TOF collaboration, in particular with regard to implications to ADS, is briefly discussed.

Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

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

Measurement of the neutron-capture cross section on 63,65Cu between 0.4 and 7.5 MeV

Science.gov (United States)

Bray, Isabel; Bhike, Megha; Krishichayan, (None); Tornow, W.

2015-10-01

Copper is currently being used as a cooling and shielding material in most experimental searches for 0 ν β β decay. In order to accurately interpret background events in these experiments, the cross section of neutron-induced reactions on copper must be known. The purpose of this work was to measure the cross section of the 63,65Cu(n, γ)64,66Cu reactions. Data were collected through the activation method at a range of energies from approximately 0.4 MeV to 7.5 MeV, employing the neutron production reactions 3H(p,n)3Heand2H(d,n)3He. Previous data were limited to energies below approximately 3 MeV. The results are compared to predictions from the nuclear data libraries ENDF/B-VII.1 and TENDL-2014.

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

Science.gov (United States)

Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

2016-12-01

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract.

Neutron source investigations in support of the cross section program at the Argonne Fast-Neutron Generator

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

Neutron and proton transmutation-activation cross section libraries to 150 MeV for application in accelerator-driven systems and radioactive ion beam target-design studies

International Nuclear Information System (INIS)

Koning, A.J.; Chadwick, M.B.; MacFarlane, R.E.; Mashnik, S.; Wilson, W.B.

1998-05-01

New transmutation-activation nuclear data libraries for neutrons and protons up to 150 MeV have been created. These data are important for simulation calculations of radioactivity, and transmutation, in accelerator-driven systems such as the production of tritium (APT) and the transmutation of waste (ATW). They can also be used to obtain cross section predictions for the production of proton-rich isotopes in (p,xn) reactions, for radioactive ion beam (RIB) target-design studies. The nuclear data in these libraries stem from two sources: for neutrons below 20 MeV, we use data from the European activation and transmutation file, EAF97; For neutrons above 20 MeV and for protons at all energies we have isotope production cross sections with the nuclear model code HMS-ALICE. This code applies the Monte Carlo Hybrid Simulation theory, and the Weisskopf-Ewing theory, to calculate cross sections. In a few cases, the HMS-ALICE results were replaced by those calculated using the GNASH code for the Los Alamos LA150 transport library. The resulting two libraries, AF150.N and AF150.P, consist of 766 nuclides each and are represented in the ENDF6-format. An outline is given of the new representation of the data. The libraries have been checked with ENDF6 preprocessing tools and have been processed with NJOY into libraries for the Los Alamos transmutation/radioactivity code CINDER. Numerous benchmark figures are presented for proton-induced excitation functions of various isotopes compared with measurements. Such comparisons are useful for validation purposes, and for assessing the accuracy of the evaluated data. These evaluated libraries are available on the WWW at: http://t2.lanl.gov/. 21 refs

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

Nuclear data relevant to single event upsets in semiconductor memories induced by cosmic-ray neutrons and protons

International Nuclear Information System (INIS)

Watanabe, Yukinobu

2008-01-01

The role of nuclear data is examined in the study of single event upset (SEU) phenomena in semiconductor memories caused by cosmic-ray neutrons and protons. Neutron and proton SEU cross sections are calculated with a simplified semi-empirical model using experimental heavy-ion SEU cross-sections and a dedicated database of neutron and proton induced reactions on 28 Si. Some impacts of the nuclear reaction data on SEU simulation are analyzed by investigating relative contribution of secondary ions and neutron elastic scattering to SEU and influence of simultaneous multiple ions emission on SEU. (author)

Mercury mass measurement in fluorescent lamps via neutron activation analysis

Science.gov (United States)

Viererbl, L.; Vinš, M.; Lahodová, Z.; Fuksa, A.; Kučera, J.; Koleška, M.; Voljanskij, A.

2015-11-01

Mercury is an essential component of fluorescent lamps. Not all fluorescent lamps are recycled, resulting in contamination of the environment with toxic mercury, making measurement of the mercury mass used in fluorescent lamps important. Mercury mass measurement of lamps via instrumental neutron activation analysis (NAA) was tested under various conditions in the LVR-15 research reactor. Fluorescent lamps were irradiated in different positions in vertical irradiation channels and a horizontal channel in neutron fields with total fluence rates from 3×108 cm-2 s-1 to 1014 cm-2 s-1. The 202Hg(n,γ)203Hg nuclear reaction was used for mercury mass evaluation. Activities of 203Hg and others induced radionuclides were measured via gamma spectrometry with an HPGe detector at various times after irradiation. Standards containing an Hg2Cl2 compound were used to determine mercury mass. Problems arise from the presence of elements with a large effective cross section in luminescent material (europium, antimony and gadolinium) and glass (boron). The paper describes optimization of the NAA procedure in the LVR-15 research reactor with particular attention to influence of neutron self-absorption in fluorescent lamps.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-06

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

On the research activities in reactor and neutron physics using the first egyptian research reactor

International Nuclear Information System (INIS)

Hassan, A.M.

2000-01-01

A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

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 resonance parameters for 206Pb and of the thermal neutron capture cross section for 206Pb and 209Bi

International Nuclear Information System (INIS)

Borella, A.

2005-01-01

Chapter 1 describes the motivation of the measurements (accelerator driven systems, stellar nucleosynthesis, neutron induced reactions on 206 Pb), the present status of the neutron capture data for 206 Pb and 209 Bi and the structure of this work. In Chapter 2 the basic reaction theory underlying this work is described. The neutron induced reaction mechanism and formalism are explained. The parameterisation of the cross section in terms of R-matrix theory is discussed and we put particular emphasis on the statistical behaviour of the resonance parameters and the impact of the angular distribution of gamma rays following neutron capture. The relation between experimental observables and the resonance parameters is discussed together with general comments related to resonance shape analysis. Chapter 3 is focused on the determination of resonance parameters for 206 Pb. We performed high-resolution transmission and capture measurements at the Time-Of-Flight (TOF) facility GELINA of the IRMM at Geel (B) and determined the resonance parameters. For nuclei like 206 Pb, where the total width is dominated by Γ n , the capture area allows to determine G . Transmission measurements were carried out to determine Γ n , and the statistical factor g of resonances. Before performing a Resonance Shape Analysis (RSA) on the transmission and capture data, we verified the neutron flux and resolution at GELINA. We also compared the characteristics of GELINA with those of the n-TOF facility at CERN. A special emphasis is placed on the total energy detection technique using C 6 D 6 detectors. This technique was applied for the determination of the capture cross section. To reduce systematic bias effects on the capture cross section, the response of the detectors was determined by Monte Carlo simulations, which has been validated by experiments. Using these response functions the partial capture cross sections for individual resonances of 206 Pb have been deduced, by unfolding the

Isotonic and isotopic dependence of the radiative neutron capture cross-section on the neutron excess

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

Analysis of cosmic ray neutron-induced single-event phenomena

International Nuclear Information System (INIS)

Tukamoto, Yasuyuki; Watanabe, Yukinobu; Nakashima, Hideki

2003-01-01

We have developed a database of cross sections for the n+ 28 Si reaction in the energy range between 2 MeV and 3 GeV in order to analyze single-event upset (SEU) phenomena induced by cosmic-ray neutrons in semiconductor memory devices. The data are applied to calculations of SEU cross sections using the Burst Generation Rate (BGR) model including two parameters, critical charge and effective depth. The calculated results are compared with measured SEU cross-sections for energies up to 160 MeV, and the reaction products that provide important effects on SEU are mainly investigated. (author)

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.

Current quests in nucleosynthesis: present and future neutron-induced reaction measurements

Directory of Open Access Journals (Sweden)

Praena J.

2014-03-01

Full Text Available We present some open questions in nucleosynthesis focused on the measurement of relevant neutron capture cross-sections and on new experimental methods. We review the recent 63Ni(n,γ experiment carried out at the n_TOF facility at CERN and its astrophysical implications as well as future experiments and opportunities at n_TOF. We argue some improvements in the measurement of cross-sections by activation arising from a new method for the generation of stellar neutron spectra. We show preliminary results of the experimental validation of the method. We discuss the astrophysical implications of the 181Ta(n,γ stellar cross-section measured with this method. Finally, we describe challenging experiments consisting of in situ radioactive ion beams and stellar neutron beams.

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

International Nuclear Information System (INIS)

Pashchenko, A.B.

1996-12-01

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

Total reaction cross sections and neutron-removal cross sections of neutron-rich light nuclei measured by the COMBAS fragment-separator

Science.gov (United States)

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.

Measurement of neutron activation cross sections in the energy range between 2 and 7 MeV by using a Ti-deuteron target and a deuteron gas target

Energy Technology Data Exchange (ETDEWEB)

Senga, T.; Sakane, H.; Shibata, M.; Yamamoto, H.; Kawade, K. [Nagoya Univ. (Japan); Kasugai, Yoshimi; Ikeda, Yujiro; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-03-01

Using a Ti-deuteron target in the neutron energy range between 2 and 4.5 MeV and a deuteron gas target between 4.5 and 7 MeV, mono-energetic neutrons could be generated enough for activation cross section measurements. The KN-3750 Van de Grraff accelerator at Nagoya University and the Fusion Neutronics Source (FNS) at Japan Atomic Energy Research Institute (JAERI) were used. Preliminary results of activation cross sections were obtained for reactions of {sup 27}Al(n,p){sup 27}Mg, {sup 47}Ti(n,p){sup 47}Sc, {sup 58}Ni(n,p){sup 58}Co. The evaluation data of JENDL-3.2 showed reasonable agreement with our results. (author)

Neutron induced bystander effect among zebrafish embryos

Science.gov (United States)

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

2015-12-01

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

Neutron capture cross-section measurements for 238U between 0.4 and 1.4 MeV

Science.gov (United States)

Krishichayan, Fnu; Finch, S. W.; Howell, C. R.; Tonchev, A. P.; Tornow, W.

2017-09-01

Neutron-induced radiative-capture cross-section data of 238U are crucial for fundamental nuclear physics as well as for Stewardship Science, for advanced-fuel-cycle calculations, and for nuclear astrophysics. Based on different techniques, there are a large number of 238U(n, γ) 239U cross-section data available in the literature. However, there is a lack of systematic and consistent measurements in the 0.1 to 3.0 MeV energy range. The goal of the neutron-capture project at TUNL is to provide accurate 238U(n, γ) 239U cross-section data in this energy range. The 238U samples, sandwiched between gold foils of the same size, were irradiated for 8-14 hours with monoenergetic neutrons. To avoid any contribution from thermal neutrons, the 238U and 197Au targets were placed inside of a thin-walled pill-box made of 238U. Finally, the whole pill-box was wrapped in a gold foil as well. After irradiation, the samples were gamma-counted at the TUNL's low-background counting facility using high-efficient HPGe detectors. The 197Au monitor foils were used to calculate the neutron flux. The experimental technique and 238U(n, γ) 239U cross-section results at 6 energies will be discussed during the meeting.

A Bragg curve counter with an internal production target for the measurement of the double-differential cross-section of fragment production induced by neutrons at energies of tens of MeV

International Nuclear Information System (INIS)

Sanami, T.; Hagiwara, M.; Oishi, T.; Hosokawa, M.; Kamada, S.; Tanaka, Su.; Iwamoto, Y.; Nakashima, H.; Baba, M.

2009-01-01

A Bragg curve counter equipped with an internal production target was developed for the measurements of double-differential cross-sections of fragment production induced by neutrons at energies of tens of MeV. The internal target permitted a large detection solid angle and thus the registration of processes at low production rates. In this specific geometry, the detection solid angle depends on the emission angle and the range of the particle. Therefore the energy, atomic number, and angle of trajectory of the particle have to be taken into account for the determination of the solid angle. For the selection of events with tracks confined within a defined cylindrical volume around the detector axis, a segmented anode was applied. The double-differential cross-sections for neutron-induced production of lithium, beryllium, and boron fragments from a carbon target were measured at 0 deg. for 65 MeV neutrons. The results are in good agreement with theoretical calculation using PHITS code with GEM and ISOBAR model.

Improved Fission Neutron Data Base for Active Interrogation of Actinides

Energy Technology Data Exchange (ETDEWEB)

Pozzi, Sara; Czirr, J. Bart; Haight, Robert; Kovash, Michael; Tsvetkov, Pavel

2013-11-06

This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems both with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).

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

International Nuclear Information System (INIS)

Panini, G.C.

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1987-06-01

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

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

Energy Technology Data Exchange (ETDEWEB)

Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Trkov, Andrej [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); IAEA, Vienna International Centre, PO Box 100, A-1400 Vienna (Austria); Jaćimović, Radojko [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Gregoire, Gilles; Destouches, Christophe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St. Paul-Lez-Durance (France)

2016-12-21

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

International Nuclear Information System (INIS)

Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

2016-01-01

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

Bayesian statistics applied to neutron activation data for reactor flux spectrum analysis

International Nuclear Information System (INIS)

Chiesa, Davide; Previtali, Ezio; Sisti, Monica

2014-01-01

Highlights: • Bayesian statistics to analyze the neutron flux spectrum from activation data. • Rigorous statistical approach for accurate evaluation of the neutron flux groups. • Cross section and activation data uncertainties included for the problem solution. • Flexible methodology applied to analyze different nuclear reactor flux spectra. • The results are in good agreement with the MCNP simulations of neutron fluxes. - Abstract: In this paper, we present a statistical method, based on Bayesian statistics, to analyze the neutron flux spectrum from the activation data of different isotopes. The experimental data were acquired during a neutron activation experiment performed at the TRIGA Mark II reactor of Pavia University (Italy) in four irradiation positions characterized by different neutron spectra. In order to evaluate the neutron flux spectrum, subdivided in energy groups, a system of linear equations, containing the group effective cross sections and the activation rate data, has to be solved. However, since the system’s coefficients are experimental data affected by uncertainties, a rigorous statistical approach is fundamental for an accurate evaluation of the neutron flux groups. For this purpose, we applied the Bayesian statistical analysis, that allows to include the uncertainties of the coefficients and the a priori information about the neutron flux. A program for the analysis of Bayesian hierarchical models, based on Markov Chain Monte Carlo (MCMC) simulations, was used to define the problem statistical model and solve it. The first analysis involved the determination of the thermal, resonance-intermediate and fast flux components and the dependence of the results on the Prior distribution choice was investigated to confirm the reliability of the Bayesian analysis. After that, the main resonances of the activation cross sections were analyzed to implement multi-group models with finer energy subdivisions that would allow to determine the

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)

Method and apparatus for determination of temperature, neutron absorption cross section and neutron moderating power

Science.gov (United States)

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.

Light charged particle production induced by fast neutrons (En=25-65 MeV) on 209Bi

International Nuclear Information System (INIS)

Raeymackers, Erwin; Slypen, Isabelle; Benck, Sylvie; Meulders, Jean-Pierre; Nica, Ninel; Corcalciuc, Valentin

2002-01-01

This paper presents the experimental set-up and data reduction procedures regarding the measurement of double-differential cross sections for light charged particle production in fast neutron induced reactions (n, px), (n, dx), (n, tx) and (n, αx) on bismuth in the incident neutron energy range 25-65 MeV and at laboratory angles from 20deg to 160deg. preliminary double-differential and energy-differential cross sections for hydrogen isotopes are presented. (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)

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

International Nuclear Information System (INIS)

Gardner, D.G.

1975-01-01

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

Verification of dosimetry cross sections above 10 MeV based on measurement of activation reaction rates in fission neutron field

International Nuclear Information System (INIS)

Odano, Naoteru; Miura, Toshimasa; Yamaji, Akio.

1996-01-01

To validate the dosimetry cross sections in fast neutron energy range, activation reaction rates were measured for 5 types of dosimetry cross sections which have sensitivity in the energy rage above 10 MeV utilizing JRR-4 reactor of JAERI. The measured reaction rates were compared with the calculations reaction rates by a continuous energy monte carlo code MVP. The calculated reaction rates were based on two dosimetry files, JENDL Dosimetry File and IRDF-90.2. (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

Source Correlated Prompt Neutron Activation Analysis for Material Identification and Localization

Science.gov (United States)

Canion, Bonnie; McConchie, Seth; Landsberger, Sheldon

2017-07-01

This paper investigates the energy spectrum of photon signatures from an associated particle imaging deuterium tritium (API-DT) neutron generator interrogating shielded uranium. The goal is to investigate if signatures within the energy spectrum could be used to indirectly characterize shielded uranium when the neutron signature is attenuated. By utilizing the correlated neutron cone associated with each pixel of the API-DT neutron generator, certain materials can be identified and located via source correlated spectrometry of prompt neutron activation gamma rays. An investigation is done to determine if fission neutrons induce a significant enough signature within the prompt neutron-induced gamma-ray energy spectrum in shielding material to be useful for indirect nuclear material characterization. The signature deriving from the induced fission neutrons interacting with the shielding material was slightly elevated in polyethylene-shielding depleted uranium (DU), but was more evident in some characteristic peaks from the aluminum shielding surrounding DU.

Neutron-induced reactions on U and Th - A new approach via AMS

International Nuclear Information System (INIS)

Wallner, A.; Capote, R.; Christl, M.; Fifield, L.K.; Srncik, M.; Tims, S.; Hotchkis, M.; Krasa, A.; Lachner, J.; Lippold, J.; Plompen, A.; Semkova, V.; Steier, P.; Winkler, S.

2014-01-01

Recent studies exhibit discrepancies at keV and MeV energies between major nuclear data libraries for 238 U(n,γ), 232 Th(n,γ) and also for (n,xn) reactions. We have extended our initial (n,γ) measurements on 235,238 U to higher neutron energies and to additional reaction channels. Neutron-induced reactions on 232 Th and 238 U were measured by a combination of the activation technique and atom counting of the reaction products using accelerator mass spectrometry (AMS). Natural thorium and uranium samples were activated with quasi-monoenergetic neutrons at IRMM. Neutron capture data were produced for neutron energies between 0.5 and 5 MeV. Fast neutron-induced reactions were studied in the energy range from 17 to 22 MeV. Preliminary data indicate a fair agreement with data libraries; however at the lower band of existing data. This approach represents a complementary method to on-line particle detection techniques and also to conventional decay counting. (authors)

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

International Nuclear Information System (INIS)

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

1981-10-01

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

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.

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)

A compact fast-neutron producing target for high resolution cross section measurements

NARCIS (Netherlands)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1984-08-01

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

Beryllium neutron activation detector for pulsed DD fusion sources

International Nuclear Information System (INIS)

Talebitaher, A.; Springham, S.V.; Rawat, R.S.; Lee, P.

2011-01-01

A compact fast neutron detector based on beryllium activation has been developed to perform accurate neutron fluence measurements on pulsed DD fusion sources. It is especially well suited to moderate repetition-rate ( 9 Be(n,α) 6 He cross-section, energy calibration of the proportional counters, and numerical simulations of neutron interactions and beta-particle paths using MCNP5. The response function R(E n ) is determined over the neutron energy range 2-4 MeV. The count rate capability of the detector has been studied and the corrections required for high neutron fluence measurements are discussed. For pulsed DD neutron fluencies >3×10 4 cm -2 , the statistical uncertainty in the fluence measurement is better than 1%. A small plasma focus device has been employed as a pulsed neutron source to test two of these new detectors, and their responses are found to be practically identical. Also the level of interfering activation is found to be sufficiently low as to be negligible.

Neutron Scattering Differential Cross Sections for 12C

Science.gov (United States)

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).

Neutron activation analysis of certified samples by the absolute method

Science.gov (United States)

Kadem, F.; Belouadah, N.; Idiri, Z.

2015-07-01

The nuclear reactions analysis technique is mainly based on the relative method or the use of activation cross sections. In order to validate nuclear data for the calculated cross section evaluated from systematic studies, we used the neutron activation analysis technique (NAA) to determine the various constituent concentrations of certified samples for animal blood, milk and hay. In this analysis, the absolute method is used. The neutron activation technique involves irradiating the sample and subsequently performing a measurement of the activity of the sample. The fundamental equation of the activation connects several physical parameters including the cross section that is essential for the quantitative determination of the different elements composing the sample without resorting to the use of standard sample. Called the absolute method, it allows a measurement as accurate as the relative method. The results obtained by the absolute method showed that the values are as precise as the relative method requiring the use of standard sample for each element to be quantified.

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)

Neutron-induced fission of uranium isotopes up to 100 MeV

International Nuclear Information System (INIS)

Lestone, J.P.; Gavron, A.

1994-01-01

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

Cosmic-ray-induced ship-effect neutron measurements and implications for cargo scanning at borders

Energy Technology Data Exchange (ETDEWEB)

Kouzes, Richard T. [Pacific Northwest National Laboratory, MS K7-36, P.O. Box 999, Richland, WA 99352 (United States)], E-mail: richard.kouzes@pnl.gov; Ely, James H.; Seifert, Allen; Siciliano, Edward R.; Weier, Dennis R.; Windsor, Lindsay K.; Woodring, Mitchell L. [Pacific Northwest National Laboratory, MS K7-36, P.O. Box 999, Richland, WA 99352 (United States); Borgardt, James; Buckley, Elise; Flumerfelt, Eric; Oliveri, Anna; Salvitti, Matt [Juniata College Physics Department, 1700 Moore St., Huntingdon, PA 16652 (United States)

2008-03-11

Neutron measurements are used as part of the interdiction process for illicit nuclear materials at border crossings. Even though the natural neutron background is small, its variation can impact the sensitivity of detection systems. The natural background of neutrons that is observed in monitoring instruments arises almost entirely from cosmic-ray-induced cascades in the atmosphere and the surrounding environment. One significant source of variation in the observed neutron background is produced by the 'ship effect' in large quantities of cargo that transit past detection instruments. This paper reports on results from measurements with typical monitoring equipment of ship effect neutrons in various materials. One new result is the 'neutron shadow shielding' effect seen with some low neutron density materials.

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

International Nuclear Information System (INIS)

Maslov, V.M.

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-11-01

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

Estimated neutron-activation data for TFTR. Part II. Biological dose rate from sample-materials activation

International Nuclear Information System (INIS)

Ku, L.; Kolibal, J.G.

1982-06-01

The neutron induced material activation dose rate data are summarized for the TFTR operation. This report marks the completion of the second phase of the systematic study of the activation problem on the TFTR. The estimations of the neutron induced activation dose rates were made for spherical and slab objects, based on a point kernel method, for a wide range of materials. The dose rates as a function of cooling time for standard samples are presented for a number of typical neutron spectrum expected during TFTR DD and DT operations. The factors which account for the variations of the pulsing history, the characteristic size of the object and the distance of observation relative to the standard samples are also presented

Angular distribution and cross section measurements of 64Zn(n,α)61Ni reaction for neutron energy 5 MeV

International Nuclear Information System (INIS)

Chen Yingtang; Chen Zemin; Qi Huiquan; Li Mingtao

1995-01-01

A twin gridded ionization chamber with dual parameter data acquisition system is used to study neutron induced charged particle emission reaction. The angular distribution and cross section of α-particles from the 64 Zn(n,α) 61 Ni reaction are measured at neutron energy 5 MeV

European Collaboration for High-Resolution Measurements of Neutron Cross Sections between 1 MeV and 250 MeV

CERN Multimedia

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...

Neutron induced fission of U isotopes up to 100 MeV

International Nuclear Information System (INIS)

Lestone, J.P.; Gavron, A.

1993-01-01

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

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.

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Thermal neutron imaging in an active interrogation environment

International Nuclear Information System (INIS)

Vanier, P.E.; Forman, L.; Norman, D.R.

2009-01-01

We have developed a thermal-neutron coded-aperture imager that reveals the locations of hydrogenous materials from which thermal neutrons are being emitted. This imaging detector can be combined with an accelerator to form an active interrogation system in which fast neutrons are produced in a heavy metal target by means of excitation by high energy photons. The photo-induced neutrons can be either prompt or delayed, depending on whether neutronemitting fission products are generated. Provided that there are hydrogenous materials close to the target, some of the photo-induced neutrons slow down and emerge from the surface at thermal energies. These neutrons can be used to create images that show the location and shape of the thermalizing materials. Analysis of the temporal response of the neutron flux provides information about delayed neutrons from induced fission if there are fissionable materials in the target. The combination of imaging and time-of-flight discrimination helps to improve the signal-to-background ratio. It is also possible to interrogate the target with neutrons, for example using a D-T generator. In this case, an image can be obtained from hydrogenous material in a target without the presence of heavy metal. In addition, if fissionable material is present in the target, probing with fast neutrons can stimulate delayed neutrons from fission, and the imager can detect and locate the object of interest, using appropriate time gating. Operation of this sensitive detection equipment in the vicinity of an accelerator presents a number of challenges, because the accelerator emits electromagnetic interference as well as stray ionizing radiation, which can mask the signals of interest.

Activation cross section measurement at neutron energy from 13.3 to 14.9 MeV using FNS facility

Energy Technology Data Exchange (ETDEWEB)

Kasugai, Yoshimi; Ikeda, Yujiro; Uno, Yoshitomo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yamamoto, Hiroshi; Kawade, Kiyoshi [Nagoya Univ. (Japan)

2001-03-01

Sixty activation cross sections have been measured in the neutron energy between 13.4 and 14.9 MeV using intense D-T neutrons source (Fusion Neutronics Source, FNS) at JAERI. The following reactions are included in this work: (1) 32 reactions mainly for lanthanide isotopes, (2) 19 reactions for short-lived products (the half-lives are from 1 s to 20 min) and (3) 9 (n, n{alpha}) reactions. The experimental results were compared with the data reported previously and the evaluated data of ENDF/B-VI Rev. 4, JENDL-3.2 and FENDL/A-2.0. The present data for the (n, p) and (n, {alpha}) reactions were compared with the values estimated by using the empirical formulae proposed by our group in order to validate the systematics for the reactions for the lanthanide isotopes. Systematic trend of (n, n{alpha}) reactions were discussed based on the present data. (author)

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

Neutron activation analysis in minerals prospecting

International Nuclear Information System (INIS)

Gomez, H.; Duque O, J.

1988-01-01

One method multielemental analysis in geological samples has been developed by neutron activation analysis without using standards and by eliminating many of the error sources of the absolute method. It uses the ratio of the activities induced by mass unit, between the element in the sample and one cobalt monitor. The detection limits are good for more than thirty elements in many prospecting programs, with a standard deviation less than 7%. The neutron flux used is 2x10 11 nxcm -2 .S -1 and the HPGE detector has a relative efficiency of 20% and an energy resolution of 1.9 KeV in 1332 KeV photopeak

Neutronic investigation and activation calculation for CFETR HCCB blankets

Science.gov (United States)

Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

2017-12-01

The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

Fissile materials in solution concentration measured by active neutron interrogation

International Nuclear Information System (INIS)

Romeyer Dherbey, J.; Passard, Ch.; Cloue, J.; Bignan, G.

1993-01-01

The use of the active neutron interrogation to measure the concentration of plutonium contained in flow solutions is particularly interesting for fuel reprocessing plants. Indeed, this method gives a signal which is in a direct relation with the fissile materials concentration. Moreover, it is less sensitive to the gamma dose rate than the other nondestructive methods. Two measure methods have been evolved in CEA. Their principles are given into details in this work. The first one consists to detect fission delayed neutrons induced by a 252 Cf source. In the second one fission prompt neutrons induced by a neutron generator of 14 MeV are detected. (O.M.)

Activation neutron detector

International Nuclear Information System (INIS)

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

1976-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-05-21

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

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

Deuteron-induced activation data in EAF for IFMIF calculations

International Nuclear Information System (INIS)

Forrest, R.; Cook, I.

2006-01-01

The main type of activation calculations needed for fusion technology deals with the interaction of neutrons with materials. The road map for development of fusion as an electricity producing technology is based on ITER and IFMIF followed by DEMO. IFMIF is a materials testing facility that will enable materials planned to be used in DEMO to be irradiated to very high fluences, so providing the database of material properties required for the licensing of DEMO. IFMIF will use intense beams of high energy deuterons striking a flowing lithium target to produce the neutron field. Although the neutron spectrum is a good match to those produced in a D-T fusion device, there is a significant high energy tail extending up to 55 MeV. These high energy neutrons were the motivation for increasing the upper energy limit in the neutron-induced part of EAF-2005 so that activation calculations could be made in IFMIF. The deuterons themselves will also make a contribution to activation especially in the target where they strike the lithium but also due to beam losses in the accelerator. It was realised that because of corrosion in the lithium loop there is the potential for a wide range of elements to be present in the target region and it is therefore necessary to have a complete library of deuteron-induced cross section data, just as in the neutron case. A preliminary library based on model calculations with TALYS using global parameters was used to construct a deuteron-induced library and this was released as part of the maintenance release of EAF-2005.1 at the beginning of this year. This data library has been used with an updated version of the inventory code FISPACT to calculate the activation in the lithium target due to reactions of the deuterons with the corrosion products. These calculations show that deuterons are much more important than neutrons (about a factor of 70) in activating the elements other than lithium. This work shows the importance of the effect and means

Measurement of thermal neutron cross-sections and resonance integrals for sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As by using sup 2 sup 4 sup 1 Am-Be isotopic neutron source

CERN Document Server

Karadag, M; Tan, M; Oezmen, A

2003-01-01

Thermal neutron cross-sections and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions were measured by the activation method. The experimental samples with and without a cylindrical Cd shield case in 1 mm wall thickness were irradiated in an isotropic neutron field of the sup 2 sup 4 sup 1 Am-Be neutron source. The induced activities in the samples were measured by high-resolution gamma-ray spectrometry with a calibrated reverse-electrode germanium detector. Thermal neutron cross-sections for 2200 m/s neutrons and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions have been obtained relative to the reference values, sigma sub 0 =13.3+-0.1 b and I sub 0 =14.0+-0.3 b for the sup 5 sup 5 Mn(n,gamma) sup 5 sup 6 Mn reaction as a single comparator. The necessary correction factors for gamma attenuation, thermal neutron and resonance neutron self-shielding effects were taken into...

The total neutron cross-section of Nb at different temperatures for neutrons with energies below 1 eV

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-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.)

Neutron induced activity in fuel element components

International Nuclear Information System (INIS)

Kjellbert, N.

1978-03-01

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

A PC-program for the calculation of neutron flux and element contents using the ki-method of neutron activation analysis

International Nuclear Information System (INIS)

Boulyga, E.G.; Boulyga, S.F.

2000-01-01

A computer program is described, which calculates the induced activities of isotopes after irradiation in a known neutron field, thermal and epithermal neutron fluxes from the measured induced activities and from nuclear data of 2-4 monitor nuclides as well as the element concentrations in samples irradiated together with the monitors. The program was developed for operation in Windows 3.1 (or higher). The application of the program for neutron activation analysis allows to simplify the experimental procedure and to reduce the time. The program was tested by measuring different types of standard reference materials at the FRJ-2 (Research Centre, Juelich, Germany) and Triga (University Mainz, Germany) reactors. Comparison of neutron flux parameters calculated by this program with those calculated by a VAX program developed at the Research Centre, Juelich was done. The results of testing seem to be satisfactory. (author)

Development of low-activation design method for reduction of radioactive waste (2). Precise neutron flux and activation estimation of nuclear power plants using MATXSLIB-J33T10

International Nuclear Information System (INIS)

Uematsu, Mikio; Hayashi, Katsumi; Nemezawa, Shigeki; Ogata, Tomohiro; Nakata, Mikihiro; Kinno, Masaharu; Yamaguchi, Katsuyoshi; Saito, Minoru; Hasegawa, Akira

2008-01-01

We have been developing low-activation concrete for biological shielding wall of nuclear power plants, for the purpose of reducing large amount of radioactive waste. Based on measurement of Eu and Co content in various aggregate candidates, limestone and electro-fused alumina were selected as the most feasible aggregate for low activation concrete. Induced activity in shielding wall was calculated for both low activation concrete and ordinary concrete using neutron flux obtained from DORT two-dimensional calculation made for typical ABWR and APWR models. We have prepared new cross section library named 'MATXSLIB-J33T10 that has multi-group structure in thermal energy. The library was processed from evaluated cross section library JENDL 3.3 by using NJOY 99.83. Activation cross section library for ORIGEN-79 code is prepared for each activation calculation case by collapsing JENDL-3.3 originated 183-group constants into 3-group activation cross section using 183-group neutron flux. One-group activation cross section was also prepared in the same manner for ORIGEN2 calculation. The ΣD/C value results for low-activation concrete was sufficiently low comparing to the ordinary concrete. By using the developed low-activation concrete, activation level of biological shielding wall concrete will be effectively decreased. The use of the developed low-activation concrete will contribute to economization of nuclear power plants decommissioning by reducing large amount of radioactive concrete waste. (author)

Neutron induced alpha production from carbon between 18 and 22 MeV

International Nuclear Information System (INIS)

Stevens, A.P.

1976-10-01

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

Neutron capture studies of {sup 206}Pb at a cold neutron beam

Energy Technology Data Exchange (ETDEWEB)

Schillebeeckx, P.; Kopecky, S.; Quetel, C.R.; Tresl, I.; Wynants, R. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); Belgya, T.; Szentmiklosi, L. [Institute for Energy Security and Environmental Safety, Centre for Energy Research, Budapest (Hungary); Borella, A. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); SCK CEN, Mol (Belgium); Mengoni, A. [Nuclear Data Section, International Atomic Energy Agency (IAEA), Wagramerstrasse 5, PO Box 100, Vienna (Austria); Agenzia Nazionale per le Nuove Tecnologie, l' Energia e lo Sviluppo Economico Sostenibile (ENEA), Bologna (Italy)

2013-11-15

Gamma-ray transitions following neutron capture in {sup 206}Pb have been studied at the cold neutron beam facility of the Budapest Neutron Centre using a metallic sample enriched in {sup 206}Pb and a natural lead nitrate powder pellet. The measurements were performed using a coaxial HPGe detector with Compton suppression. The observed {gamma} -rays have been incorporated into a decay scheme for neutron capture in {sup 206}Pb. Partial capture cross sections for {sup 206}Pb(n, {gamma}) at thermal energy have been derived relative to the cross section for the 1884 keV transition after neutron capture in {sup 14}N. From the average crossing sum a total thermal neutron capture cross section of 29{sup +2}{sub -1} mb was derived for the {sup 206}Pb(n, {gamma}) reaction. The thermal neutron capture cross section for {sup 206}Pb has been compared with contributions due to both direct capture and distant unbound s-wave resonances. From the same measurements a thermal neutron-induced capture cross section of (649 {+-} 14) mb was determined for the {sup 207}Pb(n, {gamma}) reaction. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

Holmqvist, B; Wiedling, T

1969-06-15

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

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

International Nuclear Information System (INIS)

Carlsson, Magnus

2004-06-01

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

Measurements of neutron cross section of the {sup 243}Am(n,{gamma}){sup 244}Am reaction

Energy Technology Data Exchange (ETDEWEB)

Hatsukawa, Yuichi; Shinohara, Nobuo; Hata, Kentaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

The effective thermal neutron cross section of {sup 243}Am(n,{gamma}){sup 244}Am reaction was measured by the activation method. Highly-purified {sup 243}Am target was irradiated in an aluminum capsule by using a research reactor JRR-3M. The tentative effective thermal neutron cross sections are 3.92 b, and 84.44 b for the production of {sup 244g}Am and {sup 244m}Am, respectively. (author)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Calculation of the neutron noise induced by periodic deformations of a large sodium-cooled fast reactor core

International Nuclear Information System (INIS)

Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.

2014-01-01

The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)

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

Measurements of the total neutron cross-sections of poly- and mono-germanium crystals at neutron energies below 1 eV

International Nuclear Information System (INIS)

Maayouf, R.M.A.; Abdel-Kawy, A.; Abbas, Y.; Habib, N.; Adib, M.; Hamouda, I.

1983-12-01

Total neutron cross-section measurements have been performed for poly and mono-germanium crystals in the energy range from 2 meV-1eV. The measurements were performed using two TOF and a double axis crystal spectrometer installed at the ET-RR-1 reactor. The obtained 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 Ge and the analysis of its neutron diffraction pattern. The incoherent and thermal diffuse scattering cross-sections of Ge were estimated from the analysis of the total cross-section data obtained for Ge mono-crystal

Simple, empirical approach to predict neutron capture cross sections from nuclear masses

Science.gov (United States)

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

A neutron activation detector

International Nuclear Information System (INIS)

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

1973-01-01

The present invention concerns a neutron activation detector made from a moulded and hardened composition. According to the invention, that composition contains an activable substance constituted by at least two chemical elements and/or compounds of at least two chemical elements. Each of these chemical elements is capable of reacting with the neutrons forming radio-active isotopes with vatious levels of energy during desintegration. This neutron detector is mainly suitable for measuring integral thermal neutron and fast neutron fluxes during irradiation of the sample, and also for measuring the intensities of neutron fields [fr

The progress of neutron induced prompt gamma analysis technique in 1988-2002

International Nuclear Information System (INIS)

Liu Yuren; Jing Shiwei

2003-01-01

The new development of the neutron induced prompt gamma-ray analysis (NIPGA) technology in 1988-2002 are described. The pulse fast-thermal neutron activation analysis method, which utilizes the inelastic reaction and capture reaction jointly is employed to measure the elemental content in the material more efficiently. The lifetime of the neutron generator is more than 10000 h and the capability of HPGe, TeZeCd and MCA (multi-channel analyser) reaches the high level. At the same time, Monte Carlo Library least-square method is used to solve the nonlinearity problem in the PGNAA (Prompt Gamma Neutron Activation Analysis)

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

Production of neutron-rich nuclei in fission induced by neutrons generated by the p+ sup 1 sup 3 C reaction at 55 MeV

CERN Document Server

Stroe, L; Andrighetto, A; Tecchio, L B; Dendooven, P; Huikari, J; Pentillä, H; Peraejaervi, K; Wang, Y

2003-01-01

Cross-sections for the production of neutron-rich nuclei obtained by neutron-induced fission of natural uranium have been measured. The neutrons were generated by bombarding a sup 1 sup 3 C target with 55 MeV protons. The results, position of the maximum in the (Z, A)-plane, width and magnitude, are very comparable with those where the neutrons are generated by bombardment of natural sup 1 sup 2 C graphite with 50 MeV deuterons. Depending on the geometry of the converter/target assembly the isotope yields, however, are a factor of 2-3 lower due to less efficient production of neutrons per primary projectile, especially at small forward angles. (orig.)

The importance of using the mixed neutron flux in activation analysis of D-3He fueled reactors

International Nuclear Information System (INIS)

Khater, H.Y.; Sawan, M.E.

1992-01-01

This paper reports on the D-D and D-T secondary reactions in D- 3 He reactors which provide the neutron source term for most of the radioactivity produced in the structure of the reactor. radionuclides are produced as a result of neutron interactions with their parent nuclides. The amount of activity produced by any radionuclide depends on the number of its parent atoms present at any given time. One approach to account for the activity induced by both neutron sources in any activation analysis is to add their individual contributions. Performing two separate calculations for the D-D and D-T neutron flux components and adding their contributions yields conservative results due to underestimating the destruction of the parent atoms. The overestimation is more pronounced for short and intermediate lived nuclides, long operation time, large neutron flux and large destruction cross section for the parent atoms. In the steel first wall of a typical d- 3 He reactor, adding the individual contributions of the tow neutron sources results in overestimating the activities produced by most of the radioactive isotopes of Ag, Lu, Ta, W and Re. After 30 years of reactor operation, the activity of 187 W, which is a major source of safety concern in case of an accident, is more than an order of magnitude higher than its value if the mixed neutron flux is used. The activity of 188 Re, which is an important source of offsite does in case of accidental release, is overestimated by more than a factor of two

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Neutron Library (ENDL82) in the transmittal format

International Nuclear Information System (INIS)

Howerton, R.J.; Dye, R.E.; Perkins, S.T.

1982-01-01

There are four main libraries of data included within the system described. They are ENDL (Evaluated Neutron Data Library), ECPL (Evaluated Charged-Particle Data Library), ACTL (Evaluated Neutron-Induced Activation Cross-Section Library), and EGDL (Evaluated Photon Interaction Data Library). The first three deal with nuclear processes induced by neutrons or light charged particles (Z less than or equal to 2, A less than or equal to 4). The fourth (EGDL) contains the data appropriate to photons with energies between 100 eV and 100 MeV that interact with atoms of the elements in their ground state, i.e., cold targets. EGDL does not contain data for photonuclear reactions

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

Neutron spectometers

International Nuclear Information System (INIS)

Poortmans, F.

1977-01-01

Experimental work in the field of low-energy neutron physics can be subdivided into two classes: 1)Study of the decay process of the compound-nucleus state as for example the study of the capture gamma rays and of the neutron induced fission process; 2)Study of the reaction mechanism, mainly by measuring the reaction cross-sections and resonance parameters. These neutron cross-sections and resonance parameters are also important data required for many technological applications especially for reactor development programmes. In general, the second class of experiments impose other requirements on the neutron spectrometer than the first class. In most cases, a better neutron energy resolution and a broader neutron energy range are required for the study of the reaction mechanism than for the study of various aspects of the decay process. (author)

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)

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Cross-sections of 197Au(n,α)198Au and 63Cu(n,α)64Cu induced by 252Cf neutrons

International Nuclear Information System (INIS)

Mandal, Ranjita; Sengupta, D.; Roy, Malobika; Naik, Mamta; Bhoraskar, V.N.

2014-01-01

Analytical work, employing nuclear techniques, is normally carried out through (n,α) reaction because of the availability of neutrons either from reactors or laboratory sources such as Sb-Be, Am-Be, Ra-Be, Po-Be, 252 Cf, etc. The laboratory neutron sources are though portable and adaptable to a particular experimental arrangement, suffer from the disadvantage of slow neutron yield (except 252 Cf). In this set up since the neutrons available are monoenergetic, it was thought appropriate to initiate a program to measure cross-sections of a few nuclear reactions which have practical applications. Earlier studies on cross-section measurement of the reaction 197 Au(n,α) 198 Au and 63 Cu(n,α) 64 Cu has been carried out using different sources, monitors and techniques

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)

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)

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

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

International Nuclear Information System (INIS)

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

1984-05-01

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

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.)

Total cross section measurement of radioactive isotopes with a thin beam neutron spectrometer

International Nuclear Information System (INIS)

Razbudej, V.F.; Vertebnyj, V.P.; Padun, G.S.; Muravitskij, A.V.

1975-01-01

The method for measuring the neutron total cross sections of radioactive isotopes by a time-of-flight spectrometer with a narrow (0.17 mm in diameter) beam of thermal neutrons is described. The distinguishing feature of this method is the use of capillary samples with a small amount of substance (0.05-1.0 mg). The energy range is 0.01-0.3 eV. The total cross sections of irradiated samples of sub(153)Eu and sub(151)Eu are measured. From them are obtained the cross sections of sub(152)Eu (Tsub(1/2)=12.4 g) and of sub(154)E (Tsub(1/2)=8.6 yr); they equal 11400+-1400 and 1530+-190 barn at E=0.0253 eV. The cross section of the sub(152)Eu absorption for the thermal spectrum (T=333 K) is determined by the activation method; it is 8900+-1200 barn

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

Parameterization of neutron production double-differential cross section above several tens-MeV by the use of moving source model

International Nuclear Information System (INIS)

Kitsuki, Hirohiko; Shigyo, Nobuhiro; Ishibashi, Kenji

2000-01-01

The moving source model based on the Maxwell-like energy distribution with Gaussian shape terms are employed for analyzing the neutron emission spectra from proton-induced spallation reaction. The parallelization of the double differential cross section is made for the experimental and calculated neutron data in the energy region from several-tens MeV to 3 GeV. (author)

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Measurement of neutron-production double-differential cross sections for continuous neutron-incidence reaction up to 100 MeV

International Nuclear Information System (INIS)

Kunieda, Satoshi; Watanabe, Takehito; Shigyo, Nobuhiro; Ishibashi, Kenji; Satoh, Daiki; Nakamura, Takashi; Haight, Robert C.

2004-01-01

The inclusive measurements of neutron-incident neutron-production double-differential cross sections in intermediate energy range is now being carried out. Spallation neutrons are used as incident particles. As a part of this, the experiment was performed by using of NE213 liquid organic scintillators to detect outgoing-neutrons. Incident-neutron energy was determined by time-of-flight technique, and outgoing-neutron energy spectrum was derived by unfolding light-output spectrum of NE213 with response functions calculated by SCINFUL-R. Preliminary cross sections were obtained up to about 100 MeV, and were compared with calculations by the GNASH code. It is hoped to get pure measurements by using measured response functions for our detectors used in this study. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Carlsson, Magnus

2004-06-01

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

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)

Delayed neutron yield from fast neutron induced fission of 238U

International Nuclear Information System (INIS)

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

2002-01-01

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

Determination of manganese in some pyrolusite ores of Myanmar by neutron activation analysis

International Nuclear Information System (INIS)

Myint, U.; Swe, M.T.

1994-01-01

Manganese in pyrolusite ores from various regions of Myanmar was determined by thermal neutron activation analysis using an Am(Be) neutron source. The induced activities of 56 Mn were monitored by a γ-counting technique. (author) 2 refs.; 1 tab

Determination of manganese in some pyrolusite ores of Myanmar by neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Myint, U; Swe, M T [Yangon Univ. (Myanmar). Dept. of Chemistry

1994-05-17

Manganese in pyrolusite ores from various regions of Myanmar was determined by thermal neutron activation analysis using an Am(Be) neutron source. The induced activities of [sup 56]Mn were monitored by a [gamma]-counting technique. (author) 2 refs.; 1 tab.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Measurement of 241Am Ground State Radiative Neutron Capture Cross Section with Cold Neutron Beam. Progress Report on Research Contract HUN14318 for the CRP on Minor Actinide Neutron Reaction Data (MANREAD)

International Nuclear Information System (INIS)

Belgya, T.; Szentmiklosi, L.; Kis, Z.; Nagy, N.M.; Konya, J.

2012-01-01

The ground state cross section of 242 Am has been measured with beams of cold neutrons at the Budapest Research Reactor using the X-ray emission of the decay product of 242 Pu. This methodology avoids the uncertainty caused by resonance neutrons in the pile activations. The target was characterized with gamma and X-ray spectrometry. The obtained ground state cross section is 540 ± 32 b, which is at the low end of the most recent literature values, but agrees with most of them within their uncertainty. (author)

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)

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-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

Passive detectors for neutron fluence measurement

International Nuclear Information System (INIS)

Holt, P.D.

1985-01-01

The use of neutron activation detectors (slow neutron detectors and threshold detectors) and fission track detectors for radiological protection purposes, principally in criticality dosimetry, dosimetry of pulsed accelerators and calibration of neutron fluxes is discussed. References are given to compilations of cross sections. For the determination of the activity induced, either beta ray or gamma ray counting may be used. For beta-ray counting, thin foils are usually necessary which result in low neutron sensitivity. When fission track detectors are used, it is necessary to know the efficiency of track registration. Alternatively, a detector-counter system may be calibrated by exposure to a known flux of monoenergetic neutrons. Usually, the sensitivity of activation detectors is low because small foils are used. For criticality dosimetry, calibration work and shielding studies on accelerators, low sensitivity is acceptable. However, there are some instances where, by the use of long integration times, or very large quantities of detector material with gamma ray detection, neutron fluences in operational areas have been measured. (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

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)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

[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

Assessment and comparison of different multigroup neutron cross section libraries for dosimetry purposes

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)

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)

FY16 Status Report on NEAMS Neutronics Activities

Energy Technology Data Exchange (ETDEWEB)

Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Smith, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jung, Y. S. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-09-30

The goal of the NEAMS neutronics effort is to develop a neutronics toolkit for use on sodium-cooled fast reactors (SFRs) which can be extended to other reactor types. The neutronics toolkit includes the high-fidelity deterministic neutron transport code PROTEUS and many supporting tools such as a cross section generation code MC²-3, a cross section library generation code, alternative cross section generation tools, mesh generation and conversion utilities, and an automated regression test tool. The FY16 effort for NEAMS neutronics focused on supporting the release of the SHARP toolkit and existing and new users, continuing to develop PROTEUS functions necessary for performance improvement as well as the SHARP release, verifying PROTEUS against available existing benchmark problems, and developing new benchmark problems as needed. The FY16 research effort was focused on further updates of PROTEUS-SN and PROTEUS-MOCEX and cross section generation capabilities as needed.

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

International Nuclear Information System (INIS)

Mori, Takamasa

1985-05-01

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

Multicomponent activation detector measurements of reactor neutron spectra

International Nuclear Information System (INIS)

Sandberg, J.; Aarnio, P. A.; Routti, J. T.

1984-01-01

Information on the neutron flux is required in many applications of research reactors, such as activation analysis or radiation damage measurements. Flux spectrum measurements are commonly carried out with activation foils. The reaction types used are threshold reactions in the fast energy region, resonance reactions in the intermediate region and neutron capture reactions with l/v-cross section in the thermal region. It has been shown that it is possible to combine several detector elements into homogeneous multicomponent detectors. The activities of all detector reaction products can be determined with a single gamma spectrum measurement. The multicomponent principle sets some restrictions on the choice of detector reactions, for example, each product nuclide may be produced in one reaction only. Separate multicomponent threshold and resonance detectors were designed for the fast and intermediate regions, respectively. The detectors were fabricated in polyethylene irradiation capsules or quartz glass ampoules, and they were irradiated in a cadmium cover. The detectors were succesfully used in the irradiation ring and in the core of a Triga reactor. The intermediate and fast neutron spectrum was unfolded with the least-squares unfolding program LOUHI. According to the preliminary results multicomponent activation detectors might constitute a convenient means for carrying out routine neutron spectrum measurements in research reactors. (orig.)

Data reduction for a high-throughput neutron activation analysis system

International Nuclear Information System (INIS)

Bowman, W.W.

1979-01-01

To analyze samples collected as part of a geochemical survey for the National Uranium Resource Evaluation program, Savannah River Laboratory has installed a high-throughput neutron activation analysis system. As part of that system, computer programs have been developed to reduce raw data to elemental concentrations in two steps. Program RAGS reduces gamma-ray spectra to lists of photopeak energies, peak areas, and statistical errors. Program RICHES determines the elemental concentrations from photopeak and delayed-neutron data, detector efficiencies, analysis parameters (neutron flux and activation, decay, and counting times), and spectrometric and cross-section data from libraries. Both programs have been streamlined for on-line operation with a minicomputer, each requiring approx. 64 kbytes of core. 3 tables

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

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

Neutron Capture Cross Sections of Zr and La: Probing Neutron Exposure and Neutron Flux in Red Giant Stars

CERN Document Server

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.

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)

Surrogate Measurements of Actinide (n,2n) Cross Sections with NeutronSTARS

Energy Technology Data Exchange (ETDEWEB)