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Sample records for fission cross sections

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.
Neutron-induced fission cross sections

International Nuclear Information System (INIS)

Weigmann, H.

1991-01-01

In the history of fission research, neutron-induced fission has always played the most important role. The practical importance of neutron-induced fission rests upon the fact that additional neutrons are produced in the fission process, and thus a chain reaction becomes possible. The practical applications of neutron-induced fission will not be discussed in this chapter, but only the physical properties of one of its characteristics, namely (n,f) cross sections. The most important early summaries on the subject are the monograph edited by Michaudon which also deals with the practical applications, the earlier review article on fission by Michaudon, and the review by Bjornholm and Lynn, in which neutron-induced fission receives major attention. This chapter will attempt to go an intermediate way between the very detailed theoretical treatment in the latter review and the cited monograph which emphasizes the applied aspects and the techniques of fission cross-section measurements. The more recent investigations in the field will be included. Section II will survey the properties of cross sections for neutron-induced fission and also address some special aspects of the experimental methods applied in their measurement. Section Ill will deal with the formal theory of neutron-induced nuclear reactions for the resolved resonance region and the region of statistical nuclear reactions. In Section IV, the fission width, or fission transmission coefficient, will be discussed in detail. Section V will deal with the broader structures due to incompletely damped vibrational resonances, and in particular will address the special case of thorium and neighboring isotopes. Finally, Section VI will briefly discuss parity violation effects in neutron-induced fission. 74 refs., 14 figs., 3 tabs
Measurements of fission cross-sections. Chapter 4

International Nuclear Information System (INIS)

James, G.D.

1981-01-01

The steps involved in the measurement of fission cross sections are summarized and the range of techniques available are considered. Methods of fission detection are described with particular emphasis on the neutron energy dependent properties of the fission process and the details of fragment energy loss which can lead to energy-dependent changes in detector efficiency. Selected examples of fission cross-section measurements are presented and methods of data reduction, storage, analysis and evaluation, are examined. Finally requested accuracies for fission cross section data are compared to estimated available accuracies. (U.K.)
Fission fragment angular distributions and fission cross section validation

International Nuclear Information System (INIS)

Leong, Lou Sai

2013-01-01

The present knowledge of angular distributions of neutron-induced fission is limited to a maximal energy of 15 MeV, with large discrepancies around 14 MeV. Only 238 U and 232 Th have been investigated up to 100 MeV in a single experiment. The n-TOF Collaboration performed the fission cross section measurement of several actinides ( 232 Th, 235 U, 238 U, 234 U, 237 Np) at the n-TOF facility using an experimental set-up made of Parallel Plate Avalanche Counters (PPAC), extending the energy domain of the incident neutron above hundreds of MeV. The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. I will show the methods we used to reconstruct the full angular resolution by the tracking of fission fragments. Below 10 MeV our results are consistent with existing data. For example in the case of 232 Th, below 10 MeV the results show clearly the variation occurring at the first (1 MeV) and second (7 MeV) chance fission, corresponding to transition states of given J and K (total spin and its projection on the fission axis), and a much more accurate energy dependence at the 3. chance threshold (14 MeV) has been obtained. In the spallation domain, above 30 MeV we confirm the high anisotropy revealed in 232 Th by the single existing data set. I'll discuss the implications of this finding, related to the low anisotropy exhibited in proton-induced fission. I also explore the critical experiments which is valuable checks of nuclear data. The 237 Np neutron-induced fission cross section has recently been measured in a large energy range (from eV to GeV) at the n-TOF facility at CERN. When compared to previous measurements, the n-TOF fission cross section appears to be higher by 5-7 % beyond the fission threshold. To check the relevance of n-TOF data, we simulate a criticality experiment performed at Los Alamos with a 6 kg sphere of 237 Np. This
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)
Modelisation of the fission cross section

International Nuclear Information System (INIS)

Morariu, Claudia

2013-03-01

The neutron cross sections of four nuclear systems (n+ 235 U, n+ 233 U, n+ 241 Am and n+ 237 Np) are studied in the present document. The target nuclei of the first case, like 235 U and 239 Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237 Np and 241 Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author) [fr
Fission cross sections in the intermediate energy region

International Nuclear Information System (INIS)

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

1991-01-01

Until recently there has been very little cross section data for neutron-induced fission in the intermediate energy region, primarily because no suitable neutron source has existed. At Los Alamos, the WNR target-4 facility provides a high-intensity source of neutrons nearly ideal for fission measurements extending from a fraction of a MeV to several hundred MeV. This paper summarizes the status of fission cross section data in the intermediate energy range (En > 30 MeV) and presents our fission cross section data for 235 U and 238 U compared to intranuclear cascade and statistical model predictions
Fission cross sections in the intermediate energy region

Energy Technology Data Exchange (ETDEWEB)

Lisowski, P.W.; Gavron, A.; Parker, W.E.; Ullmann, J.L.; Balestrini, S.J. (Los Alamos National Lab., NM (USA)); Carlson, A.D.; Wasson, O.A. (National Inst. of Standards and Technology, Gaithersburg, MD (USA)); Hill, N.W. (Oak Ridge National Lab., TN (USA))

1991-01-01

Until recently there has been very little cross section data for neutron-induced fission in the intermediate energy region, primarily because no suitable neutron source has existed. At Los Alamos, the WNR target-4 facility provides a high-intensity source of neutrons nearly ideal for fission measurements extending from a fraction of a MeV to several hundred MeV. This paper summarizes the status of fission cross section data in the intermediate energy range (En > 30 MeV) and presents our fission cross section data for {sup 235}U and {sup 238}U compared to intranuclear cascade and statistical model predictions.
Fission cross section measurements of actinides at LANSCE

Energy Technology Data Exchange (ETDEWEB)

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

2010-01-01

Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications. By combining measurement at two LANSCE facilities, Lujan Center and the Weapons Neutron Research center (WNR), differential cross sections can be measured from sub-thermal energies up to 200 MeV. Incident neutron energies are determined using the time-of-flight method, and parallel-plate ionization chambers are used to measure fission cross sections relative to the {sup 235}U standard. Recent measurements include the {sup 233,238}U, {sup 239,242}Pu and {sup 243}Am neutron-induced fission cross sections. In this paper preliminary results for cross section data of {sup 243}Am and {sup 233}U will be presented.
238U subthreshold neutron induced fission cross section

International Nuclear Information System (INIS)

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

1976-01-01

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

International Nuclear Information System (INIS)

Hambsch, F.J.

2004-01-01

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

Energy Technology Data Exchange (ETDEWEB)

Ohkawachi, Yasushi; Ohki, Shigeo; Wakabayashi, Toshio [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1998-03-01

Fission cross section ratios of minor actinide nuclides (Am-241, Am-243) relative to U-235 in the fast neutron energy region have been measured using a back-to-back (BTB) fission chamber at YAYOI fast neutron source reactor. A small BTB fission chamber was developed to measure the fission cross section ratios in the center of the core at YAYOI reactor. Dependence of the fission cross section ratios on neutron spectra was investigated by changing the position of the detector in the reactor core. The measurement results were compared with the fission cross sections in the JENDL-3.2, ENDF/B-VI and JEF-2.2 libraries. It was found that calculated values of Am-241 using the JENDL-3.2, ENDF/B-VI and JEF-2.2 data are lower by about 15% than the measured value in the center of the core (the neutron average energy is 1.44E+6(eV)). And, good agreement can be seen the measured value and calculated value of Am-243 using the JENDL-3.2 data in the center of the core (the neutron average energy is 1.44E+6)(eV), but calculated values of Am-243 using the ENDF/B-VI and JEF-2.2 data are lower by 11% and 13% than the measured value. (author)
Interference analysis of fission cross section

International Nuclear Information System (INIS)

Toshkov, S.A.; Yaneva, N.B.

1976-01-01

The formula for the reaction cross-section based on the R-matrix formalism considering the interference between the two neighbouring resonances, referred to the same value of total momentum was used for the analysis of the cross-section of resonance neutron induced fission of 230Pu. The experimental resolution and thermal motion of the target nuclei were accounted for numerical integration
Prediction of fission mass-yield distributions based on cross section calculations

International Nuclear Information System (INIS)

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

2005-01-01

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-01-01

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

CERN Multimedia

Wiescher, M; Cox, J; Dahlfors, M

2002-01-01

A measurement of the neutron induced fission cross sections of $^{237}$Np, $^{241},{243}$Am and of $^{245}$Cm is proposed for the n_TOF neutron beam. Two sets of fission detectors will be used: one based on PPAC counters and another based on a fast ionization chamber (FIC). A total of 5x10$^{18}$ protons are requested for the entire fission measurement campaign.
Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

International Nuclear Information System (INIS)

Gauld, I.C.

2005-01-01

U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k eff ) to determine the net importance of cross sections to k eff . The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: 151 Sm, 103 Rh, 155 Eu, 150 Sm, 152 Sm, 153 Eu, 154 Eu, and 143 Nd
Nuclear fission and neutron-induced fission cross-sections

CERN Document Server

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

2013-01-01

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

Systematics of fission cross sections at the intermediate energy region

Energy Technology Data Exchange (ETDEWEB)

Fukahori, Tokio; Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

The systematics was obtained with fitting experimental data for proton induced fission cross sections of Ag, {sup 181}Ta, {sup 197}Au, {sup 206,207,208}Pb, {sup 209}Bi, {sup 232}Th, {sup 233,235,238}U, {sup 237}Np and {sup 239}Pu above 20 MeV. The low energy cross section of actinoid nuclei is omitted from systematics study, since the cross section has a complicated shape and strongly depends on characteristic of nucleus. The fission cross sections calculated by the systematics are in good agreement with experimental data. (author)
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)
Fission cross section measurements for minor actinides

Energy Technology Data Exchange (ETDEWEB)

Fursov, B. [IPPE, Obninsk (Russian Federation)

1997-03-01

The main task of this work is the measurement of fast neutron induced fission cross section for minor actinides of {sup 238}Pu, {sup 242m}Am, {sup 243,244,245,246,247,248}Cm. The task of the work is to increase the accuracy of data in MeV energy region. Basic experimental method, fissile samples, fission detectors and electronics, track detectors, alpha counting, neutron generation, fission rate measurement, corrections to the data and error analysis are presented in this paper. (author)
MCNP6 Fission Cross Section Calculations at Intermediate and High Energies

OpenAIRE

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

2013-01-01

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

International Nuclear Information System (INIS)

Bouland, O.; Lynn, J.E.; Talou, P.

2014-01-01

This paper discusses two common approximations used to calculate average fission cross sections over the compound energy range: the disregard of the W II factor and the Porter-Thomas hypothesis made on the double barrier fission width distribution. By reference to a Monte Carlo-type calculation of formal R-matrix fission widths, this work estimates an overall error ranging from 12% to 20% on the fission cross section in the case of the 239 Pu fissile isotope in the energy domain from 1 to 100 keV with very significant impact on the competing capture cross section. This work is part of a recent and very comprehensive formal R-matrix study over the Pu isotope series and is able to give some hints for significant accuracy improvements in the treatment of the fission channel. (authors)
Cross sections of the lumped fission products for the AMZ library

International Nuclear Information System (INIS)

Ono, S.; Corcueca, R.P.; Nascimento, J.A.

1985-01-01

The preparation of the lumped fission product cross section for the AMZ library is described. For this purpose 100 nuclides were selected. The cross sections for each nuclide were generated by the NJOY code with evaluated nuclear data from ENDF/B-V, complemented with ENDF/B-IV data. A comparison is performed between the data obtained and the lumped fission product cross section of JFS-II [pt
Tables of RCN-2 fission-product cross section evaluation

International Nuclear Information System (INIS)

Gruppelaar, H.

1979-05-01

This report (continuation of ECN-13 and ECN-33) describes the third part of the RCN-2 evaluation of neutron cross sections for fission product nuclides in KEDAK format. It contains evaluated data for nine nuclides, i.e. 142 Nd, 143 Nd, 144 Nd, 145 Nd, 146 Nd, 147 Nd, 148 Nd, 150 Nd and 147 Pm. Most emphasis has been given to the evaluation of the radiative capture cross section, in order to provide a data base for adjustment calculations using results of integral measurements. Short evaluation reports are given for this cross section. The evaluated capture cross sections are compared with recent experimental differential and integral data. Graphs are given of the capture cross sections at neutron energies above 1 keV, in which also adjusted point cross sections, based upon integral STEK and CFRMF data have been plotted. Moreover, the results are compared with those of the well-known ENDF/B-IV evaluation for fission product nucleides. Finally, evaluation summaries are given, which include tables of other important neutron cross sections, such as the total, elastic scattering and inelastic scattering cross sections
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)
Status update on the NIFFTE high precision fission cross section measurement program

International Nuclear Information System (INIS)

Laptev, Alexander B.; Tovesson, Fredrik; Burgett, Eric; Greife, Uwe; Grimes, Steven; Heffner, Michael D.; Hertel, Nolan E.; Hill, Tony; Isenhower, Donald; Klay, Jennifer L.; Kornilov, Nickolay; Kudo, Ryuho; Loveland, Walter; Massey, Thomas; McGrath, Chris; Pickle, Nathan; Qu, Hai; Sharma, Sarvagya; Snyder, Lucas; Thornton, Tyler; Towell, Rusty S.; Watson, Shon

2010-01-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) program has been underway for nearly two years. The program's mission is to measure fission cross sections of the primary fissionable and fissile materials ( 235 U, 239 Pu, 238 U) as well as the minor actinides across energies from approximately 50 keV up to 20 MeV with an absolute uncertainty of less than one percent while investigating energy ranges from below an eV to 600 MeV. This basic nuclear physics data is being reinvestigated to support the next generation power plants and a fast burner reactor program. Uncertainties in the fast, resolved and unresolved resonance regions in plutonium and other transuranics are extremely large, dominating safety margins in the next generation nuclear power plants and power plants of today. This basic nuclear data can be used to support all aspects of the nuciear renaissance. The measurement campaign is utilizing a Time Projection Chamber or TPC as the tool to measure these cross sections to these unprecedented levels. Unlike traditional fission cross section measurements using time-of-flight and a multiple fission foil configurations in which fission cross sections in relation to that of 235 U are performed, the TPC project uses time-of-flight and hydrogen as the benchmark cross section. Using the switch to hydrogen, a simple, smooth cross section that can be used which removes the uncertainties associated with the resolved and unresolved resonances in 235 U.
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)
Fission-neutron displacement cross sections in metals

International Nuclear Information System (INIS)

Takamura, Saburo; Aruga, Takeo; Nakata, Kiyotomo

1985-01-01

The sensitivity damage rates for 22 metals were measured after fission-spectrum neutron irradiation at low temperature and the experimental damage rates were compared with the theoretical calculation. The relation between the theoretical displacement cross section and the atomic weight of metals can be written by two curves; one is for fcc and hcp metals, and another is for bcc metals. On the other hand, the experimental displacement cross section versus atomic weight is shown approximately by a curve for both fcc and bcc metals, and the cross section for hcp metals deviates from the curve. The defect production efficiency is 0.3-0.4 for fcc metals and 0.6-0.8 for bcc metals. (orig.)
High-energy Neutron-induced Fission Cross Sections of Natural Lead and Bismuth-209

CERN Document Server

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

2011-01-01

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

International Nuclear Information System (INIS)

Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

1997-01-01

Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)
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)
Fission cross section measurements at the LLL 100-MeV linac

International Nuclear Information System (INIS)

Browne, J.C.

1975-01-01

The fission cross section for 235 U was measured from thermal energy to 20 MeV in several steps. First, the cross section was measured from 8 MeV to 20 MeV relative to the n,p scattering cross section and then from thermal to one MeV relative to 6 Li(n,α). In addition, a measurement of the ratio of the fission cross sections of 235 U and 238 U relative to 235 U has been completed in the range 1 keV to 30 MeV for 233 U and 100 keV to 30 MeV for 238 U. Statistical uncertainties are less than 4 percent. (U.S.)
ENDF/B-5 fission product cross section evaluations

International Nuclear Information System (INIS)

Schenter, R.E.; England, T.R.

1979-12-01

Cross section evaluations were made for the 196 fission product nuclides on the ENDF/B-5 data files. Most of the evaluations involve updating the capture cross sections of the important absorbers for fast and thermal reactor systems. This included updating thermal values, resonance integrals, resonance parameter sets, and fast capture cross sections. For the fast capture results generalized least-squares calculations were made with the computer code FERRET. Input for these cross section adjustments included nuclear models calculations and both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, 4000. Comparisons of these evaluations with recent capture measurements are shown. 15 figures, 10 tables
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
Comparison of 235U fission cross sections in JENDL-3.3 and ENDF/B-VI

International Nuclear Information System (INIS)

Kawano, Toshihiko; Carlson, Allan D.; Matsunobu, Hiroyuki; Nakagawa, Tsuneo; Shibata, Keiichi

2002-01-01

Comparisons of evaluated fission cross sections for 235 U in JENDL-3.3 and ENDF/B-VI are carried out. The comparisons are made for both the differential and integral data. The fission cross sections as well as the fission ratios are compared with the experimental data in detail. Spectrum averaged cross sections are calculated and compared with the measurements. The employed spectra are the 235 U prompt fission neutron spectrum, the 252 Cf spontaneous fission neutron spectrum, and the neutron spectrum produced by a 9 Be(d, xn) reaction. For 235 U prompt fission neutron spectrum, the ENDF/B-VI evaluation reproduces experimental averaged cross sections. For 252 Cf and 9 Be(d, xn) neutron spectra, the JENDL-3.3 evaluation gives better results than ENDF/B-VI. (author)
Evaluations of fission product capture cross sections for ENDF/B-V

International Nuclear Information System (INIS)

Schenter, R.E.; Johnson, D.L.; Mann, F.M.; Schmittroth, F.

1979-01-01

Capture cross section evaluations were made for the 36 most important fission product absorbers in a fast reactor system. These evaluations were obtained by use of a generalized least-squares approach with calculations being performed with the computer code FERRET. These results will provide the major revisions to the ENDF/B-IV Fission Product Cross Section File which will be released as part of ENDF/B-V. Input for the cross section adjustment calculations included both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, and 4000. Comparisons of these evaluations with recent capture measurements are presented. 14 figures

Preparation of lumped fission product (FP) cross sections for a multigroup library

International Nuclear Information System (INIS)

Ono, S.; Corcuera, R.P.

1984-01-01

A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author) [pt
Measuring Cross-Section and Estimating Uncertainties with the fissionTPC

Energy Technology Data Exchange (ETDEWEB)

Bowden, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Manning, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sangiorgio, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Seilhan, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-01-30

The purpose of this document is to outline the prescription for measuring fission cross-sections with the NIFFTE fissionTPC and estimating the associated uncertainties. As such it will serve as a work planning guide for NIFFTE collaboration members and facilitate clear communication of the procedures used to the broader community.
A compact multi-plate fission chamber for the simultaneous measurement of 233U capture and fission cross-sections

Directory of Open Access Journals (Sweden)

Bacak M.

2017-01-01

Full Text Available 233U plays the essential role of fissile nucleus in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section which is about one order of magnitude lower than the fission cross-section on average. Therefore, the accuracy in the measurement of the 233U capture cross-section essentially relies on efficient capture-fission discrimination thus a combined setup of fission and γ-detectors is needed. At CERN n_TOF the Total Absorption Calorimeter (TAC coupled with compact fission detectors is used. Previously used MicroMegas (MGAS detectors showed significant γ-background issues above 100 eV coming from the copper mesh. A new measurement campaign of the 233U capture cross-section and alpha ratio is planned at the CERN n_TOF facility. For this measurement, a novel cylindrical multi ionization cell chamber was developed in order to provide a compact solution for 14 active targets read out by 8 anodes. Due to the high specific activity of 233U fast timing properties are required and achieved with the use of customized electronics and the very fast ionizing gas CF4 together with a high electric field strength. This paper describes the new fission chamber and the results of the first tests with neutrons at GELINA proving that it is suitable for the 233U measurement.
Comparison of {sup 235}U fission cross sections in JENDL-3.3 and ENDF/B-VI

Energy Technology Data Exchange (ETDEWEB)

Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan); Carlson, Allan D. [National Institute of Standards and Technology (United States); Matsunobu, Hiroyuki [Data Engineering, Inc., Fujisawa, Kanagawa (Japan); Nakagawa, Tsuneo; Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Talou, Patrick; Young, Philip G.; Chadwick, Mark B. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2002-01-01

Comparisons of evaluated fission cross sections for {sup 235}U in JENDL-3.3 and ENDF/B-VI are carried out. The comparisons are made for both the differential and integral data. The fission cross sections as well as the fission ratios are compared with the experimental data in detail. Spectrum averaged cross sections are calculated and compared with the measurements. The employed spectra are the {sup 235}U prompt fission neutron spectrum, the {sup 252}Cf spontaneous fission neutron spectrum, and the neutron spectrum produced by a {sup 9}Be(d, xn) reaction. For {sup 235}U prompt fission neutron spectrum, the ENDF/B-VI evaluation reproduces experimental averaged cross sections. For {sup 252}Cf and {sup 9}Be(d, xn) neutron spectra, the JENDL-3.3 evaluation gives better results than ENDF/B-VI. (author)
What can be learnt from the channel analysis of the 232Th neutron fission cross section

International Nuclear Information System (INIS)

Abou Yehia, H.; Jary, J.; Trochon, J.; Boldeman, J.W.; Musgrove, A.R. de L.

1979-10-01

Channel analyses of the neutron fission cross section of 232 Th have been made in two laboratories. The calculated fission cross sections and fission fragment anisotropies are compared with the experimental data. Despite some differences in the methods used, the conclusions on the physical aspects of the fission process are very similar
Determination of the neutron-induced fission cross section of 242Pu

International Nuclear Information System (INIS)

Koegler, Toni Joerg

2016-01-01

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

International Nuclear Information System (INIS)

Tamagno, Pierre

2015-01-01

The work presented here aims to improve models used in the fission cross section evaluation. The results give insights for a significant breakthrough in this field and yielded large extensions of the evaluation code CONRAD. Partial cross sections are inherently strongly correlated together as of the competition of the related reactions must yield the total cross section. Therefore improving fission cross section benefits to all partial cross sections. A sound framework for the simulation of competitive reactions had to be settled in order to further investigate on the fission reaction; this was implemented using the TALYS reference code as guideline. After ensuring consistency and consistency of the framework, focus was made on fission. Perspective resulting from the use of macroscopic-microscopic models such as the FRDM and FRLDM were analyzed; these models have been implemented and validated on experimental data and benchmarks. To comply with evaluation requirements in terms of computation time, several specific numerical methods have been used and parts of the program were written to run on GPU. These macroscopic-microscopic models yield potential energy surfaces that can be used to extract a one-dimensional fission barrier. This latter can then be used to obtained fission transmission coefficients that can be used in a Hauser-Feshbach model. This method has been finally tested for the calculation of the average fission cross section for 239 Pu(n,f). (author) [fr
Determination of minor actinides fission cross sections by means of transfer reactions

Energy Technology Data Exchange (ETDEWEB)

Jurado, B.; Aiche, M.; Barreau, G.; Boyer, S.; Czajkowski, S.; Dassie, D.; Grosjean, C.; Guiral, A.; Haas, B.; Osmanov, B.; Petit, M. [CENBG - UMR 5795 CNRS/IN2P3-Univ. Bordeaux 1- Le Haut Vigneau, 33175 Gradignan (France); Berthoumieux, E.; Gunsing, F.; Perrot, L.; Theisen, Ch. [CEN Saclay, DSM/DAPNIA/SPhN, 91191 Gif-sur-Yvette cedex (France); Bauge, E. [CEA, SPhN, BP12 91680 Bruyeres-le-Chatel (France); Michel-Sendis, F. [IPN, 15 rue G. Clemenceau, 91406 Orsay cedex (France); Billebaud, A. [LPSC, 53 Avenue des Martyrs, 38026 Grenoble cedex (France); Wilson, J. N. [IPN, 15 rue G. Clemenceau, 91406 Orsay cedex (France); LPSC, 53 Avenue des Martyrs, 38026 Grenoble cedex (France); Ahmad, I.; Greene, J.P.; Janssens, R. V. F. [ANL, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2005-07-01

We present an original method that allows to determine neutron-induced cross sections of very short-lived minor actinides. This indirect method, based on the use of transfer reactions, has already been applied with success for the determination of the neutron-induced fission and capture cross section of {sup 233}Pa, a key nucleus in the {sup 232}Th - {sup 233}U fuel cycle. A recent experiment using this technique has been performed to determine the neutron-induced fission cross sections of {sup 242,243,244}Cm and {sup 241}Am which are present in the nuclear waste of the current U-Pu fuel cycle. These cross sections are highly relevant for the design of reactors capable to incinerate minor actinides. The first results will be illustrated. (authors)
Study of fission cross sections induced by nucleons and pions using the cascade-exciton model CEM95

International Nuclear Information System (INIS)

Yasin, Z.; Shahzad, M. I.

2007-01-01

Nucleon and pion-induced fission cross sections at intermediate and at higher energies are important in current nuclear applications, such as accelerator driven-systems (ADS), in medicine, for effects on electronics etc. In the present work, microscopic fission cross sections induced by nucleons and pions are calculated using the cascade-exciton model code CEM95 for different projectile-target combinations; at various energies and the computed cross sections are compared with the experimental data found in literature. A new approach is used to compute the fission cross sections in which a change of the ratio of the level density parameter in fission to neutron emission channels was taken into account with the change in the incident energy of the projectile. We are unable to describe well the fission cross sections without using this new approach. Proton induced fission cross sections are calculated for targets 1 97Au, 2 08Pb, 2 09Bi, 2 38U and 2 39Pu in the energy range from 20 MeV to 2000 MeV. Neutron induced fission cross sections are computed for 2 38U and 2 39Pu in the energy range from 20 MeV to 200 MeV. Negative pion induced cross sections for fission are calculated for targets 1 97Au and 2 08Pb from 50 MeV to 2500 MeV energy range. The calculated cross sections are essential to build a data library file for accelerator driven systems just like was built for conventional nuclear reactors. The computed values exhibited reasonable agreement with the experimental values found in the literature across a wide range of beam energies
Status of recent fast capture cross section evaluations for important fission product nuclides

International Nuclear Information System (INIS)

Gruppelaar, H.

1982-01-01

A comparison is made between recent evaluations of fission-product cross sections as given in the CNEN/CEA, ENDF/B-IV, ENDF/V-V, JENDL-1, RCN-2 and RCN-3 data libraries. The intercomparison is restricted to 24 important fission products in a fast power reactor. The evaluation methods used to obtain the various data files are reviewed and possible shortcomings are indicated. A survey is given of the experimental data based used in the various evaluations. Some graphs are included showing the new ENDF/B-V and RCN-3 fastcapture cross-section evaluations. Further intercomparisons are made by means of multi-group and one-group cross sections. It is shown that lumped fission-product cross sections calculated from the most recent versions of the data files are in quite good agreement with each other. This review concludes with a discussion on observed discrepancies and requests for new measurements. 78 references
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)
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)
Fission cross section calculations for 209Bi target nucleus based on fission reaction models in high energy regions

Directory of Open Access Journals (Sweden)

Kaplan Abdullah

2015-01-01

Full Text Available Implementation of projects of new generation nuclear power plants requires the solving of material science and technological issues in developing of reactor materials. Melts of heavy metals (Pb, Bi and Pb-Bi due to their nuclear and thermophysical properties, are the candidate coolants for fast reactors and accelerator-driven systems (ADS. In this study, α, γ, p, n and 3He induced fission cross section calculations for 209Bi target nucleus at high-energy regions for (α,f, (γ,f, (p,f, (n,f and (3He,f reactions have been investigated using different fission reaction models. Mamdouh Table, Sierk, Rotating Liquid Drop and Fission Path models of theoretical fission barriers of TALYS 1.6 code have been used for the fission cross section calculations. The calculated results have been compared with the experimental data taken from the EXFOR database. TALYS 1.6 Sierk model calculations exhibit generally good agreement with the experimental measurements for all reactions used in this study.
Measurement of the 235U/238U fission cross section ratio in the 235U fission neutron spectrum

International Nuclear Information System (INIS)

Azimi-Garakani, D.; Bagheri-Darbandi, M.

1983-06-01

Fission cross section ratio of 235 U to 238 U has been measured in the fast neutron field generated by the 235 U fission plate installed on the thermal column of the Tehran Research Reactor (TRR) with a Makrofol solid state nuclear track detector. The experiments were carried out with a set of total six enriched 235 U and depleted 238 U deposits with different masses and Makrofol films of 0.025mm and 0.060mm thicknesses. The chemically etched tracks were counted by an optical microscope. No significant differences were observed with the thin and the thick films. The results showed that the average fission cross section ratio is 3.83+-0.25. (author)
Consultants’ Meeting on Recommended Input Parameters for Fission Cross-Section Calculations. Summary Report

International Nuclear Information System (INIS)

Capote Noy, Roberto; Simakov, Stanislav; Goriely, Stephane; Hilaire, Stephane; Iwamoto, Osamu; Kawano, Toshihiko; Koning, Arjan

2014-12-01

A Consultants’ Meeting on “Recommended Input Parameters for Fission Cross-Section Calculations” was held at IAEA Headquarters, Vienna, Austria to define the scope, deliverables and appropriate work programme of a possible Coordinated Research Project (CRP) on the subject. Presentations are available online at https://www-nds.iaea.org/indexmeeting-crp/CM-RIPL-fission/. A new CRP was endorsed to recommend a comprehensive set of fission input parameters needed for the modelling of fission cross sections. Special attention will be given to the modelling of photon and nucleon induced reactions on actinides with emphasis on incident energies below 30 MeV. The goals and detailed deliverables of the planned CRP were proposed. A Hauser-Feshbach code intercomparison was recommended. (author)
Determination of extra-push energies for fusion from differential fission cross-section measurements

International Nuclear Information System (INIS)

Ramamurthy, V.S.; Kapoor, S.S.

1993-01-01

Apparent discrepancies between values of extra-push energies for fusion of two heavy nuclei derived through measurements of fusion evaporation residue cross sections and of differential fission cross sections have been reported by Keller et al. We show here that with the inclusion of the recently proposed preequilibrium fission decay channel in the analysis, there is no inconsistency between the two sets of data in terms of the deduced extra-push energies
Comparison of fission and capture cross sections of minor actinides

International Nuclear Information System (INIS)

Nakagawa, Tsuneo; Iwamoto, Osamu

2003-01-01

The fission and capture cross sections of minor actinides given in JENDL-3.3 are compared with other evaluated data and experimental data. The comparison was made for 32 nuclides of Th-227, 228, 229, 230, 233, 234, Pa-231, 232, 233, U-232, 234, 236, 237, Np-236, 237, 238, Pu-236, 237, 238, 242, 244, Am-241, 242, 242m, 243, Cm-242, 243, 244, 245, 246, 247 and 248. Given in the present report are figures of these cross sections and tables of cross sections at 0.0253 eV and resonance integrals. (author)
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)
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.

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
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.
Proton-induced fission cross sections on "2"0"8Pb at high kinetic energies

International Nuclear Information System (INIS)

Rodriguez-Sanchez, J.L.; Benlliure, J.; Paradela, C.; Ayyad, Y.; Alvarez-Pol, H.; Cortina-Gil, D.; Pietras, B.; Ramos, D.; Vargas, J.; Taieb, J.; Chatillon, A.; Belier, G.; Boutoux, G.; Gorbinet, T.; Laurent, B.; Martin, J.F.; Pellereau, E.; Casarejos, E.; Rodriguez-Tajes, C.

2014-01-01

Total fission cross sections of "2"0"8Pb induced by protons have been determined at 370 A, 500 A, and 650 A MeV. The experiment was performed at GSI Darmstadt where the combined use of the inverse kinematics technique with an efficient detection setup allowed us to determine these cross sections with an uncertainty below 6%. This result was achieved by an accurate beam selection and registration of both fission fragments in coincidence which were also clearly distinguished from other reaction channels. These data solve existing discrepancies between previous measurements, providing new values for the Prokofiev systematics. The data also allow us to investigate the fission process at high excitation energies and small deformations. In particular, some fundamental questions about fission dynamics have been addressed, which are related to dissipative and transient time effects. (authors)
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...
Evaluation of fission spectra and cross sections by zero-leakage core experiments

International Nuclear Information System (INIS)

Iijima, T.; Mukaiyama, T.

1979-01-01

A series of unit k-infinity core experiments were performed in FCA of JAERI to obtain the information on the equivalence of 239 Pu to 235 U in fast reactors, and to examine the inelastic slowing down cross section of 238 U. Three assemblies were built. Each assembly consists of a test zone (about 44l) of nearly unit k-infinity, a 20% enriched uranium driver and a natural uranium blanket. Assembly IV-1 (first built in 1969 and rebuilt in 1972) is an all uranium system, and Assemblies IV-1-P, IV-1-P' have a plutonium/natural uranium test zone. Three assemblies are nearly the same from the view-point of the slowing down cross section in the main energy region of the neutron spectrum, since 238 U occupies the most part of the composition. The main difference between Assembly IV-1 and the latter two is the difference in the fissile material. Fission rate ratios and k-infinity values were measured to obtain knowledge of the fission spectra and cross sections important for the criticality. In order to evaluate the inelastic slowing down cross section of 238 U, neutron spectra were measured with various methods. The analysis was done with four cross section sets. The agreement of k-infinity values between the experiment and the calculation is unsatisfactory, especially for Pu/NU systems
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.
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.)
Fission cross section and fission fragment angular distribution for oriented nucleus fission by intermediate energy neutrons (epsilon < or approximately 1 Mev)

International Nuclear Information System (INIS)

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

1985-01-01

General analysis is conducted, and formulae for fission cross section and angular distribution of fission fragments of oriented nuclei by fast neutrons are presented. Geometrical coefficients making up the formulae permitting to carry out calculations for target nuclei with spins I=3/2, 5/2, 7/2 at interaction energies epsilon < or approximately 1 MeV are tabulated. Results of demonstrative calculation of fission fragment angular distribution of oriented sup(235)U nuclei by 0.1 <= epsilon <= 1.0 MeV neutrons reveal that angular distribution weakly depends on the set of permeability factors of neutron waves applied in the calculations
Towards the high-accuracy determination of the 238U fission cross section at the threshold region at CERN – n_TOF

Directory of Open Access Journals (Sweden)

Diakaki M.

2016-01-01

Full Text Available The 238U fission cross section is an international standard beyond 2 MeV where the fission plateau starts. However, due to its importance in fission reactors, this cross-section should be very accurately known also in the threshold region below 2 MeV. The 238U fission cross section has been measured relative to the 235U fission cross section at CERN – n_TOF with different detection systems. These datasets have been collected and suitably combined to increase the counting statistics in the threshold region from about 300 keV up to 3 MeV. The results are compared with other experimental data, evaluated libraries, and the IAEA standards.
Total and fission cross-sections of 239Pu - statistical study of resonance parameters

International Nuclear Information System (INIS)

Derrien, H.; Blons, J.; Eggermann, C.; Michaudon, A.; Paya, D.; Ribon, P.

1967-01-01

The authors measured the total and fission cross-sections of 239 Pu with the linear accelerator at Saclay as a pulsed source of neutrons. The total cross-section was measured in the range from 4 to 700 eV and the best resolution used was 1.5 ns/m; the fission cross-section was measured between 4 eV and 6 keV, the best resolution having been 6 ns/m. The transmission measurements on five samples were made at the temperature of liquid nitrogen, and comparisons made with supplementary experiments at ambient temperature made it possible to determine the Doppler broadening factor (Δ = η√E). The resonances were identified from 4 to 500 eV in the total cross-section; the average level spacing was of the order of 2.4 eV. It would appear that, in this energy range, nearly all the levels were identified. The resonance parameters were determined by analysis of shape in conjunction with a least-squares programme on an IBM-7094 computer. The existence of a large number of broad resonances corresponding to very large fission widths has been shown to exist. Statistical study of the fission widths actually shows the existence of two families of resonances, one corresponding to a mean Γ f of the order of 45 meV and the other to a mean Γ/f of about 750 meV. The authors were therefore able to postulate a classification of resonances in terms of two spin states, the level population ratio in each family being: (2J 1 +1)/(2J 2 +1) = 1/3; J 1 = 0 corresponds to the broad resonances and J 2 = 1 to the narrow ones. The partial widths for radiative capture fluctuate slightly around a mean value of 40 meV. By using a multilevel programme, the authors were able to investigate the extent to which the existence of large fission widths might give rise to fictitious resonances (quasi-resonances) and perturbations and also to make a statistical study of the resonance parameters. (author) [fr
Actinide Capture and Fission Cross Section Measurements Within the Mini-Inca Project

International Nuclear Information System (INIS)

Letourneau, A.

2006-01-01

Full text of publication follows: The Mini-INCA project is devoted to precise description of the transmutation chain of Actinides within high thermal neutron fluxes. It uses the High Flux Reactor of ILL (Laue Langevin Institute) as an intense thermal neutron source to measure capture and fission cross sections. Two irradiation channels are dedicated for those measurements offering a diversity of fluxes ranging from pure thermal neutrons to 15% epithermal neutrons with intensities as high as 1*10 15 n/cm 2 /s. Standard nuclear techniques for measurements, such as α and γ-spectroscopy of irradiated samples, have been extended in order to stand all constraints due to the irradiation in high fluxes. In particular new types of fission micro-chambers have been developed to follow online the evolution of one actinide and to measure its fission cross section in reference to 235 U(n,F) standard reaction. This type of neutron detector will be used within the MEGAPIE target to on-line characterise the neutron flux and to study the potentiality of such target in terms of incineration. (author)
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)
Method of measurement of cross sections of heavy nuclei fission induced by intermediate energy protons

International Nuclear Information System (INIS)

Kotov, Alexander; Chtchetkovski, Alexander; Fedorov, Oleg; Gavrikov, Yuri; Chestnov, Yuri; Poliakov, Vladimir; Vaishnene, Larissa; Vovchenko, Vil; Fukahori, Tokio

2003-01-01

The purpose of this work is experimental studies of the energy dependence of the fission cross sections of heavy nuclei, nat Pb, 209 Bi, 232 Th, 233 U, 235 U, 238 U, 237 Np and 239 Pu, by protons at the energies from 200 to 1000 MeV. At present experiment the method based on use of the gas parallel plate avalanche counters (PPACs) for registration of complementary fission fragments in coincidence and the telescope of scintillation counters for direct counting of the incident protons on the target has been used. First preliminary results of the energy dependences of proton induced fission cross sections for nat Pb, 209 Bi, 235 U and 238 U are reported. (author)
Measurement of {sup 238}Np fission cross-section by neutrons near thermal point (preliminary results)

Energy Technology Data Exchange (ETDEWEB)

Abramo; vich, S.N.; Andreev, M.F.; Bol`shakov, Y.M. [Institute of Experimental Physics, Arzamas (Russian Federation)] [and others

1995-10-01

Measurements have been carried out of {sup 238}Np fission cross-section by thermal neutrons. The isotope {sup 238}Np was built up through the reaction {sup 238}U(p,n) on an electrostatic accelerator. Extraction and cleaning of the sample were done by ion-exchange chromatography. Fast neutrons were generated on the electrostatic accelerator through the reaction {sup 9}Be(d,n); a polyethylene block was used to slow down neutrons. Registration of fission fragments was performed with dielectric track detectors. Suggesting that the behavior of {sup 238}Np and {sup 238}U. Westscott`s factors are indentical the fission cross-section of {sup 238}Np was obtained: {sigma}{sub fo}=2110 {plus_minus} 75 barn.
Fission cross section of 235U from 1 to 6 MeV

International Nuclear Information System (INIS)

Barton, D.M.; Diven, B.C.; Hansen, G.E.; Jarvis, G.A.; Koontz, P.G.; Smith, R.K.

1976-01-01

The ratio of the neutron-induced fission cross section of 235 U to the neutron-proton scattering cross section was measured in the neutron energy region from 1 to 6 MeV. The neutron source was the T(p,n) reaction produced by a pulsed Van de Graaff proton beam on a thin tritium gas target. The use of monoenergetic neutrons allowed time-of-flight methods to be used to study carefully backgrounds and source characteristics
Using a Time Projection Chamber to Measure High Precision Neutron-Induced Fission Cross Sections

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-06

2014 LANSCE run cycle data will provide a preliminary ²³⁹Pu(n,f) cross section and will quantify uncertainties: PID and Target/beam non-uniformities. Continued running during the 2015 LANSCE run cycle: Thin targets to see both fission fragments and ²³⁹Pu(n,f) cross section and fully quantified uncertainties
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)
Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS

International Nuclear Information System (INIS)

Perkasa, Y. S.; Waris, A.; Kurniadi, R.; Su'ud, Z.

2014-01-01

Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS have been conducted. In this work, fission cross section resulted from MCNP6 prediction will be compared with result from TALYS calculation. MCNP6 with its event generator CEM03.03 and LAQGSM03.03 have been validated and verified for several intermediate and heavy nuclides fission reaction data and also has a good agreement with experimental data for fission reaction that induced by photons, pions, and nucleons at energy from several ten of MeV to about 1 TeV. The calculation that induced within TALYS will be focused mainly to several hundred MeV for actinide and sub-actinide nuclides and will be compared with MCNP6 code and several experimental data from other evaluator
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
Isotopic production cross sections of fission residues in 197Au-on-proton collisions at 800 A MeV

International Nuclear Information System (INIS)

Benlliure, J.; Armbruster, P.; Bernas, M.

2000-02-01

Interactions of 197 Au projectiles at 800 A MeV with protons leading to fission are investigated. We measured the production cross sections and velocities of all fission residues which are fully identified in atomic and mass number by using the in-flight separator FRS at GSI. The new data are compared with partial measurements of the characteristics of fission in similar reactions. Both the production cross sections and the recoil energies are relevant for a better understanding of spallation reactions. (orig.)

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
Reaction and fission cross-sections of 750AMeV 238U ions on Pb, Cu and AI-targets

International Nuclear Information System (INIS)

Hesse, M.; Aumann, T.; Czajkowski, S.; Dessagne, P.; Hanelt, E.; Kozhuharov, C.; Miehe, C.; Pfuetzner, M.; Roehl, C.; Schwab, W.; Stephan, C.; Tassan-Got, L.

1995-09-01

Charge-loss and fission cross-sections of 238 U at 750 A.MeV were measured on Al, Cu and Pb targets. The charge-loss rate was obtained by the attenuation method. Fission was selected by detecting the pair of highly ionizing fragments. Since the neutron-loss cross sections were measured in a parallel experiment for the same projectiles, all cross sections contributing to 238 U collisions on nuclei are available now as function of the target mass number and can be compared with current models. (orig.)
Evaluation of the 235U fission cross-section from 100 eV to 20 MeV

International Nuclear Information System (INIS)

Bhat, M.R.

1976-01-01

The evaluation of the 235 U fission cross section from 100 eV to 20 MeV for ENDF/B-V is described. The evaluated average cross sections from 100 eV to 200 keV are given, and it is proposed to include structure in the cross section in this energy region. Above 200 keV, the cross section is given as a smooth curve, and is recommended as a standard. Preliminary error estimates in the cross section are also given
Measurement of the uranium-235 fission cross section over the neutron energy range 1 to 6 MeV

International Nuclear Information System (INIS)

Barton, D.M.; Diven, B.C.; Hansen, G.E.; Jarvis, G.A.; Koontz, P.G.; Smith, R.K.

1976-01-01

The ratio of the fission cross section of 235 U to the scattering cross section of 1 H was measured in the 1- to 6-MeV range using monoenergetic neutrons from a pulsed 3 H(p,n) 3 He source. In this measurement, solid-state detectors determined fission fragment and recoil proton emissions from back-to-back U(99.7%) and polyethylene disks. Timing permitted discrimination against room-scattered neutron backgrounds. Absolute values for 235 U(n,f) are obtained using the Hopkins-Breit evaluation of the hydrogen-scattering cross section
Measurement of fission cross-section of actinides at n_TOF for advanced nuclear reactors

CERN Document Server

Calviani, Marco; Montagnoli, G; Mastinu, P

2009-01-01

The subject of this thesis is the determination of high accuracy neutron-induced fission cross-sections of various isotopes - all of which radioactive - of interest for emerging nuclear technologies. The measurements had been performed at the CERN neutron time-of-flight facility n TOF. In particular, in this work, fission cross-sections on 233U, the main fissile isotope of the Th/U fuel cycle, and on the minor actinides 241Am, 243Am and 245Cm have been analyzed. Data on these isotopes are requested for the feasibility study of innovative nuclear systems (ADS and Generation IV reactors) currently being considered for energy production and radioactive waste transmutation. The measurements have been performed with a high performance Fast Ionization Chamber (FIC), in conjunction with an innovative data acquisition system based on Flash-ADCs. The first step in the analysis has been the reconstruction of the digitized signals, in order to extract the information required for the discrimination between fission fragm...
A study of the effect of intermediate structure in the fission cross section of 239Pu on self-shielding factors

International Nuclear Information System (INIS)

Ganesan, S.

1978-01-01

A set of energy dependent fission widths of 1 + spin state corresponding to the recommended fission cross sections of Sowerby et al is evaluated by adjustment in the energy region 600 ev to 25 Kev. Corresponding to these mean fission widths of 1 + spin state, the intermediate resonance parameters based on Weigmann's formulation of Struitinsky's double humped fission barrier model are then obtained. Pseudorandom resonances are generated with and without the intermediate structure in the mean fission but leading to the same value of infinite dilution fission cross section. The effect of the intermediate structure on the self shielding factors was then investigated. (author)
Evaluation of fission cross sections and covariances for 233U, 235U, 238U, 239Pu, 240Pu, and 241Pu

International Nuclear Information System (INIS)

Kawano, Toshihiko; Matsunobu, Hiroyuki; Murata, Toru

2000-02-01

A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of 233 U, 235 U, 238 U, 239 Pu, 240 Pu, and 241 Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)
Measurement of the 238U subthreshold fission cross section for incident neutron energies between 0.6 and 100 keV

International Nuclear Information System (INIS)

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

1977-01-01

The neutron-induced 238 U subthreshold fission cross section has been measured in the neutron energy range between 0.6 and 100 keV. A total of 28 fission clusters were identified. The well-known clusters at 721 and 1210 eV appeared resolved into their Class I components. Average 238 U subthreshold fission cross sections were determined and compared with available results in the literature. The measurement is interpreted in terms of fission doorway (Class II levels) arising from the assumption of the existence of a double-humped fission barrier for the ( 238 U + n) compound nucleus at large deformations. On the basis of this model, several fission barrier parameters were determined
Progress report on the 14-MeV fission cross section measurements

International Nuclear Information System (INIS)

1979-01-01

The development of a recoil proton monitor was completed. It will be used to measure the neutron flux in the 14-MeV fisson cross section measurements. Extensive calculations of the efficiency of this monitor were made and compared with the calculations of other authors. It is clear that a major source of uncertainty in the efficiency is the lack of precise knowledge of the angular distribution of the n-p elastic scattering cross section. This leads to a change in efficiency of 3% depending on the form of the angular distribution that is used. A 4πβ-γ coincidence system was assembled to investigate the K-correction in determining the absolute activity of foil sources. Iron foils will be used as secondary flux standards in comparing the 14-MeV neutron flux with the fluxes in other laboratories, so this is an important correction to measure. The target and target holders that will be used in the 14-MeV measurements were designed and constructed. Preparations were completed to measure the angular distribution of the fission fragments produced in neutron-induced fission at 14 MeV. 2 figures
Fission Cross-section Measurements of (233)U, (245)Cm and (241,243)Am at CERN n_TOF Facility

CERN Document Server

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

2011-01-01

Neutron-induced fission cross-sections of minor actinides have been measured using the n_TOF white neutron source at CERN, Geneva, as part of a large experimental program aiming at collecting new data relevant for nuclear astrophysics and for the design of advanced reactor systems. The measurements at n_TOF take advantage of the innovative features of the n_TOF facility, namely the wide energy range, high instantaneous neutron flux and good energy resolution. Final results on the fission cross-section of 233U, 245Cm and 243Am from thermal to 20 MeV are here reported, together with preliminary results for 241Am. The measurement have been performed with a dedicated Fast Ionization Chamber (FIC), a fission fragment detector with a very high efficiency, relative to the very well known cross-section of 235U, measured simultaneously with the same detector.
Thermal-neutron fission cross section of 26. 1-min /sup 235/U/sup m/

Energy Technology Data Exchange (ETDEWEB)

Talbert W.L. Jr.; Starner, J.W.; Estep, R.J.; Balestrini, S.J.; Attrep M. Jr.; Efurd, D.W.; Roensch, F.R.

1987-11-01

The thermal-neutron fission cross section of /sup 235/U/sup m/ has been measured relative to the ground-state cross section. A rapid radiochemical separation procedure was developed to provide sizeable (10/sup 10/ to 10/sup 11/ atom) samples that were reasonably free of the parent /sup 239/Pu. From a series of eight measurements, the value of 1.42 +- 0.04 was obtained for the ratio sigma/sub m//sigma/sub g/.
Thermal-neutron fission cross section of 26.1-min /sup 235/U/sup m/

International Nuclear Information System (INIS)

Talbert, W.L. Jr.; Starner, J.W.; Estep, R.J.; Balestrini, S.J.; Attrep, M. Jr.; Efurd, D.W.; Roensch, F.R.

1987-01-01

The thermal-neutron fission cross section of /sup 235/U/sup m/ has been measured relative to the ground-state cross section. A rapid radiochemical separation procedure was developed to provide sizeable (10/sup 10/ to 10/sup 11/ atom) samples that were reasonably free of the parent /sup 239/Pu. From a series of eight measurements, the value of 1.42 +- 0.04 was obtained for the ratio σ/sub m//σ/sub g/
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
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
Fission cross-section measurements on 233U and minor actinides at the CERN n-TOF facility

International Nuclear Information System (INIS)

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

2010-01-01

Neutron-induced fission cross-sections of minor actinides have been measured at the white neutron source n-TOF at CERN, Geneva. The studied isotopes include 233 U, interesting for Th/U based nuclear fuel cycles, 241, 243 Am and 245 Cm, relevant for transmutation and waste reduction studies in new generation fast reactors (Gen-IV) or Accelerator Driven Systems. The measurements take advantage of the unique features of the n-TOF facility, namely the wide energy range, the high instantaneous neutron flux and the low background. Results for the involved isotopes are reported from ∼30 meV to around 1 MeV neutron energy. The measurements have been performed with a dedicated Fission Ionization Chamber (FIC), relative to the standard cross-section of the 235 U fission reaction, measured simultaneously with the same detector. Results are here reported. (authors)
Fission cross sections of some thorium, uranium, neptunium and plutonium isotopes relative to /sup 235/U

Energy Technology Data Exchange (ETDEWEB)

Meadows, J W

1983-10-01

Earlier results from the measurements, at this Laboratory, of the fission cross sections of /sup 230/Th, /sup 232/Th, /sup 233/U, /sup 234/U, /sup 236/U, /sup 238/U, /sup 237/Np, /sup 239/Pu, /sup 240/Pu, and /sup 242/Pu relative to /sup 235/U are reviewed with revisions to include changes in data processing procedures, alpha half lives and thermal fission cross sections. Some new data have also been included. The current experimental methods and procedures and the sample assay methods are described in detail and the sources of error are presented in a systematic manner. 38 references.
Fission cross sections of {sup 235,238}U and {sup 209}Bi at incident proton energies above 70 MeV

Energy Technology Data Exchange (ETDEWEB)

Obukhov, A I; Rimskij-Korsakov, A A; Eismont, V P [V.G. Khlopin Radium Inst., St. Petersburg (Russian Federation)

1997-06-01

The proton fission cross-section data of {sup 235,238}U and Bi were measured in the V.G. Khlopin Radium Institute over a wide proton energy range. The experimental and calculated data were also compared with experimental neutron values. The proton cross-section of {sup 235,238}U increased up to 60-70 MeV and then decreased. The bismuth proton fission cross-section increased in line with the rise in proton energy up to 1 GeV. (author). 21 refs, 6 figs.
Simultaneous analysis of fission and capture cross section with Adler-Adler resonance formula

International Nuclear Information System (INIS)

Cao Hengdao; Qiu Guochun

1989-01-01

The method of simultaneous analysis of fission and capture cross section for fissile nuclide with Adler-Adler resonance formula and the corresponding computer code are presented. A simple and convenient method to correct parameters μ, γ simultaneously is given in order to acquire optimized parameters. The results are satisfactory
Neutron-induced fission cross sections of uraniums up to 40 MeV

Energy Technology Data Exchange (ETDEWEB)

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

1998-11-01

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

Energy Technology Data Exchange (ETDEWEB)

Wright, R.Q.

1999-01-01

the uncertainty in calculated results and provide a better interpretation of criticality safety margins. Thus, the thrust of the Nuclear Data Task is to obtain high-resolution data in the intermediate energy region and provide fits to the data that utilize the modern RM formalism and covariance information for subsequent use in criticality predictability applications. As a subtask of the Nuclear Data Task, this review of the fission-product cross sections has several objectives. The first objective is a general data status review at various levels for the some 200 fission products. The second objective is a more detailed investigation of the top 20 fission products with regard to thermal- and intermediate-energy capture and scatter cross sections. The third objective is to demonstrate the revision of ENDF/B evaluations utilizing new data and evaluation techniques for 13 fission products. The fourth objective is to make recommendations for improvements, both specific and general in nature.

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)
Fission cross section ratios for 233,234,236U relative to 235U from 0.5 to 400 MeV

International Nuclear Information System (INIS)

Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

1991-01-01

Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of 233, 234, 236 U relative to 235 U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. 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 14.1 MeV to allow us to obtain cross section section values from the ratio data and our values for 235 U(n,f). 6 refs., 1 fig
Use of an oscillation technique to measure effective cross-sections of fissionable samples in critical assemblies

International Nuclear Information System (INIS)

Tretiakoff, O.; Vidal, R.; Carre, J.C.; Robin, M.

1964-01-01

The authors describe the technique used to measure the effective absorption and neutron-yield cross-sections of a fissionable sample. These two values are determined by analysing the signals due to the variation in reactivity (over-all signal) and the local perturbation in the flux (local signal) produced by the oscillating sample. These signals are standardized by means of a set of samples containing quantities of fissionable material ( 235 U) and an absorber, boron, which are well known. The measurements are made for different neutron spectra characterized by lattice parameters which constitute the central zone within which the sample moves. This technique is used to study the effective cross-sections of uranium-plutonium alloys for different heavy-water and graphite lattices in the MINERVE and MARIUS critical assemblies. The same experiments are carried out on fuel samples of different irradiations in order to determine the evolution of effective cross-sections as a function of the spectrum and the irradiations. (authors) [fr
Level density parameter dependence of the fission cross sections of some subactinide nuclei induced by protons with the incident energy up to 250 MeV

International Nuclear Information System (INIS)

Aydin, A.; Yalim, H.A.; Tel, E.; Sarer, B.; Unal, R.; Sarpuen, I.H.; Kaplan, A.; Dag, M.

2009-01-01

This study aims to show the dependence on the choice of the ratio of the level density parameters a f and a n corresponding to the saddle point of fission and equilibrium deformation of nucleus, respectively, of the proton induced fission cross sections of some subactinide targets. The method was employed using different level density parameter ratios for each fission cross section calculation in ALICE/ASH computer code. The ALICE/ASH code calculations were compared both with the available experimental data and with the Prokofiev systematics data. It is found that the fission cross sections dependent heavily on the choice of level density parameter ratio in the fission and neutron emission channels, a f /a n , for some subactinide nuclei. To get a good description of the measured fission cross sections for subactinide nuclei, we used a ratio of the level density parameters in the fission and neutron emission channels, a f /a n , depending both on the target-nucleus and on the energy of the projectile, in agreement with results published in literature.
Fission cross section ratios for 233,234,236U relative to 235U from 0.5 to 400 MeV

International Nuclear Information System (INIS)

Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

1992-01-01

Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of 233,234,236 U relative to 235 U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most of the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. 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 14.1 MeV which will allow us to obtain cross section values from the ratio data and our values for 235 U(n, f). (orig.)
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
Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

International Nuclear Information System (INIS)

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

1977-08-01

The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model
Measurement of isotopic cross sections of the fission fragments produced in 500 AMeV 208Pb + p reaction

International Nuclear Information System (INIS)

Fernandez-Dominguez, B.

2003-03-01

The aim of this work is the study of the fission fragments produced in the spallation reaction 208 Pb + p at 500 AMeV. The fission fragments from Z=23 up to Z=59 have been detected and identified by using the inverse kinematics technique with the high-resolution spectrometer FRS. The production cross sections and the recoil velocities of 430 nuclei have been measured. The measured data have been compared with previous data. The isotopic distributions show a high precision. However, the absolute value of the fission cross section is higher than expected. From the experimental data the characteristics of the average fissioning system have been reconstructed (Z fis , A fis , E* fis ). In addition, the number of post-fission neutrons emitted from the fission fragments, v post , has been determined by using a new method. The experimental data have been compared to the two-steps models describing the spallation reaction. The impact of the model parameters on the observables has been analysed and the reasons Leading to the observed differences between the codes are also presented. This analyse shows a good agreement with the INCL4+ABLA code. (author)
Neutron capture and fission cross section of Americium-243 in the energy range from 5 to 250 keV

International Nuclear Information System (INIS)

Wisshak, K.; Kaeppeler, F.

1983-04-01

The neutron capture and subthreshold fission cross section of 243 Am was measured in the energy range from 5 to 250 keV using 197 Au and 235 U as the respective standards. Neutrons were produced via the 7 Li(p,n) and the T(p,n) reaction with the Karlsruhe 3-MV pulsed Van de Graaff accelerator. Capture events were detected by two MoxonRae detectors with graphite and bismuthgraphite converters, respectively. Fission events were registered by a NE-213 liquid scintillator with pulse-shape discriminator equipment. Flight paths as short as 50-70 mm were used to obtain optimum signal-to-background ratio. After correction for the different efficiency of the individual converter materials the capture cross section could be determined with a total uncertainty of 3-6%. The respective values for the fission cross section are 8-12%. The results are compared to predictions of recent evaluations, which in some cases are severely discrepant. (orig.)
The evaluated neutron cross sections and resonance integrals of fission products with Z = 57-62

International Nuclear Information System (INIS)

Fedorova, A.F.; Pisanko, Zh.I.; Novoselov, G.M.

1976-01-01

Neutron cross sections at a neutron velocity of V=2200 m/s, and resonance integrals for fission products with Z=57-71 are estimated. In obtaining the recommended values the results of the neutron cross sections and resonance integrals for elements used as references were normalized in accordance with the latest adjusted values. In the course of estimation, preference was given to the more accurate methods for obtaining the measured values and to the more recent investigations
The evaluated neutron cross sections and resonance integrals of fission products with Z=63-71

International Nuclear Information System (INIS)

Fedorova, A.F.; Pisanko, Zh.I.; Novoselov, G.M.

1976-01-01

Neutron cross sections at a neutron velocity of V=2200 m/s, and the resonance integrals for fission products with Z=63-71 are estimated. In obtaining the recommended values the results were normalized of the neutron cross sections and resonance integrals for elements used as references in accordance with the latest adjusted values. In the course of estimation, preference was given to the more accurate measuring methods and the more recent investigations. Scientific publications up to 1975 have been used
Measurement of the fission cross-section ratio for 237Np/235U around 14 MeV neutron energies

International Nuclear Information System (INIS)

Desdin, L.; Szegedy, S.; Csikai, J.

1989-01-01

Fission cross-section ratio was determined for 237 Np/ 235 U around 14 MeV neutron energies with a back-to-back ionization chamber. Neutrons were produced by a 180 KV accelerator using T(d,n) 4 He reaction. No significant energy dependence was found in the cross section ratio
A Preliminary Study on Time Projection Chamber Simulation for Fission Cross Section Measurements with Geant4

International Nuclear Information System (INIS)

Kim, Jong Woon; Lee, Youngouk; Kim, Jae Cheon

2014-01-01

We present the details of the TPC simulation with Geant4 and show the results. TPC can provide more information than a fission chamber in that it is possible to distinguish different particle types. Simulations are conducted for uranium and plutonium targets with 20MeV neutrons. The simulation results are compared with the reference and show reasonable results. This is the first phase of study for realizing a TPC in the NFS at RAON, and we have more work to do, such as applying an electric field, signal processing in the simulation, and manufacturing of a TPC. The standard in fission cross section measurement is a fission chamber. It is basically just two parallel plates separated by a few centimeters of gas. A power supply connected to the plates sets up a moderate electric field. The target is deposited onto one of the plates. When fission occurs, the fragments ionize the gas, and the electric field causes the produced electrons to drift to the opposite plate, which records the total energy deposited in the chamber. A Time Projection Chamber (TPC) is a gas ionization detector similar to a fission chamber. However, it can measure the charged particle trajectories in the active volume in three dimensions by adding several readouts on the pad plane (fission chamber has only one readout one a pad plane). The specific ionization for each particle track enables the TPC to distinguish different particle types. A TPC will be used for fission cross section measurements in the Neutron Science Facility (NSF) at RAON. As a preliminary study, we present details of TPC simulation with Geant4 and discuss the results
Fusion-fission probabilities, cross sections, and structure notes of superheavy nuclei

International Nuclear Information System (INIS)

Kowal, Michał; Cap, Tomasz; Jachimowicz, Piotr; Skalski, Janusz; Siwek-Wilczyńska, Krystyna; Wilczyński, Janusz

2016-01-01

Fusion – fission probabilities in the synthesis of heaviest elements are discussed in the context of the latest experimental reports. Cross sections for superheavy nuclei are evaluated using the “Fusion by Diffusion” (FBD) model. Predictive power of this approach is shown for experimentally known Lv and Og isotopes and predictions given for Z = 119, 120. Ground state and saddle point properties as masses, shell corrections, pairing energies, and deformations necessary for cross-section estimations are calculated systematically within the multidimensional microscopic-macroscopic method based on the deformed Woods-Saxon single-particle potential. In the frame of the FBD approach predictions for production of elements heavier than Z = 118 are not too optimistic. For this reason, and because of high instability of superheavy nuclei, we comment on some structure effects, connected with the K-isomerism phenomenon which could lead to a significant increase in the stability of these systems.
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
Fission cross section ratios for sup 233,234,236 U relative to sup 235 U from 0. 5 to 400 MeV

Energy Technology Data Exchange (ETDEWEB)

Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J. (Los Alamos National Lab., NM (USA)); Carlson, A.D.; Wasson, O.A. (National Inst. of Standards and Technology, Gaithersburg, MD (USA)); Hill, N.W. (Oak Ridge National Lab., TN (USA))

1991-01-01

Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of {sup 233, 234, 236}U relative to {sup 235}U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. Those 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 value for {sup 235}U(n,f) at 14.1 MeV to allow us to obtain cross section section values from the ratio data and our values for {sup 235}U(n,f). 6 refs., 1 fig.
Upper Limits of the Fission Cross-Sections of Lead and Bismuth for Li-D Neutrons

Energy Technology Data Exchange (ETDEWEB)

Broda, E.

1945-07-01

This report was written by E. Broda and P.K. Wright at the Cavendish Laboratory (Cambridge) in April 1945 and is about the upper limits of the fission cross sections of lead and bismuth for Li-D neutrons. This report includes the experiment description and the discussion of the results. (nowak)
Standard cross-section data

International Nuclear Information System (INIS)

Carlson, A.D.

1984-01-01

The accuracy of neutron cross-section measurement is limited by the uncertainty in the standard cross-section and the errors associated with using it. Any improvement in the standard immediately improves all cross-section measurements which have been made relative to that standard. Light element, capture and fission standards are discussed. (U.K.)
Summary of fission spectrum workshop held at the National Neutron Cross Section Center, Brookhaven National Laboratory, October 23, 1978

International Nuclear Information System (INIS)

Stewart, L.

1979-03-01

In response to an action by the Standards Subcommittee of the Cross Section Evaluation Working Group, a workshop was convened to determine the status of available information on prompt fission neutron spectra. The experimental data were reviewed and theoretical models were developed. The current ENDF/B fission neutron spectra files were summarized. Further work is currently under way, especially to provide a better theoretical tool to represent energy-dependent fission spectra. 5 references
Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

International Nuclear Information System (INIS)

Leal, Luiz C; Derrien, Herve; Dunn, Michael E; Mueller, Don

2007-01-01

Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

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

International Nuclear Information System (INIS)

Lynn, J.E.

1980-01-01

The lectures discuss the possibilities and realisations of applying nuclear fission theory to the calculation of unknown nuclear data required for applications, principally in the nuclear power field. A brief description of the fundamentals of fission theory, the nature of the potential energy surface in the deformation plane, and of the inertial tensor, is given, and the accuracy of the theoretical calculations is discussed. It is concluded that it is impracticable to obtain required quantities such as neutron cross-sections from such fundamental calculations at present. On the other hand the fundamental theory reveals a wealth of phenomenological aspects of the fission process which can be incorporated into nuclear reaction theory. It is then shown how reaction theory thus extended to take correct account of the structured (''double-humped'') fission barrier can be used to parametrise the barrier by analysis of experimental data, and subsequently to calculate new data. Descriptions of computer programmes and illustrations of the application of the methods to actual physical examples are included in this account. (author)
ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

Energy Technology Data Exchange (ETDEWEB)

Chadwick, M. B. [Los Alamos National Laboratory (LANL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Oblozinsky, Pavel [Brookhaven National Laboratory (BNL); Dunn, Michael E [ORNL; Danon, Y. [Rensselaer Polytechnic Institute (RPI); Kahler, A. [Los Alamos National Laboratory (LANL); Smith, Donald L. [Argonne National Laboratory (ANL); Pritychenko, B [Brookhaven National Laboratory (BNL); Arbanas, Goran [ORNL; Arcilla, r [Brookhaven National Laboratory (BNL); Brewer, R [Los Alamos National Laboratory (LANL); Brown, D A [Brookhaven National Laboratory (BNL); Capote, R. [International Atomic Energy Agency (IAEA); Carlson, A. D. [National Institute of Standards and Technology (NIST); Cho, Y S [Korea Atomic Energy Research Institute; Derrien, Herve [ORNL; Guber, Klaus H [ORNL; Hale, G. M. [Los Alamos National Laboratory (LANL); Hoblit, S [Brookhaven National Laboratory (BNL); Holloway, Shannon T. [Los Alamos National Laboratory (LANL); Johnson, T D [Brookhaven National Laboratory (BNL); Kawano, T. [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Kim, H [Korea Atomic Energy Research Institute; Kunieda, S [Los Alamos National Laboratory (LANL); Larson, Nancy M [ORNL; Leal, Luiz C [ORNL; Lestone, J P [Los Alamos National Laboratory (LANL); Little, R C [Los Alamos National Laboratory (LANL); Mccutchan, E A [Brookhaven National Laboratory (BNL); Macfarlane, R E [Los Alamos National Laboratory (LANL); MacInnes, M [Los Alamos National Laboratory (LANL); Matton, C M [Lawrence Livermore National Laboratory (LLNL); Mcknight, R D [Argonne National Laboratory (ANL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Nobre, G P [Brookhaven National Laboratory (BNL); Palmiotti, G [Idaho National Laboratory (INL); Palumbo, A [Brookhaven National Laboratory (BNL); Pigni, Marco T [ORNL; Pronyaev, V. G. [Institute of Physics and Power Engineering (IPPE), Obninsk, Russia; Sayer, Royce O [ORNL; Sonzogni, A A [Brookhaven National Laboratory (BNL); Summers, N C [Lawrence Livermore National Laboratory (LLNL); Talou, P [Los Alamos National Laboratory (LANL); Thompson, I J [Lawrence Livermore National Laboratory (LLNL); Trkov, A. [Jozef Stefan Institute, Slovenia; Vogt, R L [Lawrence Livermore National Laboratory (LLNL); Van der Marck, S S [Nucl Res & Consultancy Grp, Petten, Netherlands; Wallner, A [University of Vienna, Austria; White, M C [Los Alamos National Laboratory (LANL); Wiarda, Dorothea [ORNL; Young, P C [Los Alamos National Laboratory (LANL)

2011-01-01

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He; Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl; K; Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides (235,238)U and (239)Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es; Fm; and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on (239)Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide
Measurements of integral cross sections in the californium-252 fission neutron spectrum

International Nuclear Information System (INIS)

Alberts, W.G.; Guenther, E.; Matzke, M.; Rassl, G.

1977-01-01

In a low-scattering arrangement cross sections averaged over the californium-252 spontaneous fission neutron spectrum were measured. The reactions 27 Al(n,α) 46 Ti, 47 Ti, 48 Ti(n,p), 54 Fe, 56 Fe(n,p), 58 Ni(n,p), 64 Zn(n,p), 115 In(n,n') were studied in order to obtain a consistent set of threshold detectors used in fast neutron flux density measurements. Overall uncertainties between 2 and 2.5% could be achieved; corrections due to neutron scattering in source and samples are discussed
ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

International Nuclear Information System (INIS)

Palmiotti, G.

2011-01-01

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 418 nuclides; (2) Covariance uncertainty data for 185 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions at higher energies for isotopes of F, Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new Decay Data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide
Fission cross section of 245Cm from 10-3 eV to 104 eV

International Nuclear Information System (INIS)

White, R.M.; Browne, J.C.; Howe, R.E.; Landrum, J.H.; Becker, J.A.

1979-01-01

The neutron-induced fission cross section of 245 Cm measured from .001 eV to 10 keV using the LLL 100-MeV Linac. The resonance data are analyzed with a multilevel-multichannel R-matrix code. The statistical distribution of R-matrix parameters extracted from the analysis are investigated and comparisons are made with previous work. 4 reference
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).
Evaluation of fission cross sections and covariances for {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu

Energy Technology Data Exchange (ETDEWEB)

Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan); Matsunobu, Hiroyuki [Data Engineering, Inc. (Japan); Murata, Toru [AITEL Corporation, Tokyo (JP)] [and others

2000-02-01

A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)
ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

Energy Technology Data Exchange (ETDEWEB)

Chadwick, M.B.; Herman, M.; Author(s): Chadwick,M.B.; Herman,M.; Oblozinsky,P.; Dunn,M.E.; Danon,Y.; Kahler,A.C.; Smith,D.L.; Pritychenko,B.; Arbanas,G.; Arcilla,R.; Brewer,R.; Brown,D.A.; Capote,R.; Carlson,A.D.; Cho,Y.S.; Derrien,H.; Guber,K.; Hale,G.M.; Hoblit,S.; Holloway,S.: Johnson,T.D.; Kawano,T.; Kiedrowski,B.C.; Kim,H.; Kunieda,S.; Larson,N.M.; Leal,L.; Lestone,J.P.; Little,R.C.; McCutchan,E.A.; MacFarlane,R.E.; MacInnes,M.; Mattoon,C.M.; McKnight,R.D.; Mughabghab,S.F.; Nobre,G.P.A.; Palmiotti,G.; Palumbo,A.; Pigni,M.T.; Pronyaev,V.G.; Sayer,R.O.; Sonzogni,A.A.; Summers,N.C.; Talou,P.; Thompson,I.J.; Trkov,A.; Vogt,R.L.; van der Marck,S.C.; Wallner,A.; White,M.C.; Wiarda,D.; Young,P.G.

2011-12-01

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides {sup 235,238}U and {sup 239}Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on {sup 239}Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0
Fission cross section measurement of Am-242m using lead slowing-down spectrometer

Energy Technology Data Exchange (ETDEWEB)

Kai, Tetsuya; Kobayashi, Katsuhei; Yamamoto, Shuji; Fujita, Yoshiaki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.; Kimura, Itsuro; Ohkawachi, Yasushi; Wakabayashi, Toshio

1998-03-01

By making use of double fission chamber and lead slowing-down spectrometer coupled to an electron linear accelerator, fission cross section for the {sup 242m}Am(n,f) reaction has been measured relative to that for the {sup 235}U(n,f) reaction in the energy range from 0.1 eV to 10 keV. The measured result was compared with the evaluated nuclear data appeared in ENDF/B-VI and JENDL-3.2, of which evaluated data were broadened by the energy resolution function of the spectrometer. Although the JENDL-3.2 data seem to be a little smaller than the present measurement, good agreement can be seen in the general shape and the absolute values. The ENDF/B-VI data are larger more than 50 % than the present values above 3 eV. (author)
Curves and tables of neutron cross sections

International Nuclear Information System (INIS)

Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi

1990-07-01

Neutron cross-section curves from the Japanese Evaluated Nuclear Data Library version 3, JENDL-3, are presented in both graphical and tabular form for users in a wide range of application areas in the nuclear energy field. The contents cover cross sections for all the main reactions induced by neutrons with an energy below 20 MeV including; total, elastic scattering, capture, and fission, (n,n'), (n,2n), (n,3n), (n,α), (n,p) reactions. The 2200 m/s cross-section values, resonance integrals, and Maxwellian- and fission-spectrum averaged cross sections are also tabulated. (author)
Measurement of the fission cross-section of {sup 235}U and {sup 239}Pu for thermal neutrons; Mesures des sections de fission de {sup 235}U et de {sup 239}Pu en neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Fraysse, G; Prosdocimi, A; Netter, F; Samour, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

Improved techniques of fast detection have been applied for determining the fission cross-sections of {sup 235}U and {sup 239}Pu with reference to the absorption cross-section of Boron. Monochromatic neutron beams of 0.0322 eV, 0.0626 eV and 0.275 eV have been employed. Use has been made of a Xe-filled gaseous scintillator and of a low-geometry solid state ion chamber. Both measured alpha and fission rates. The results at the reference energy of 0.0253 eV are: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (authors) [French] Des techniques avancees de comptage rapide ont ete mise en oeuvre pour determiner la section efficace de fission de {sup 235}U et de {sup 239}Pu par rapport a celle d'absorption du bore. Des faisceaux de neutrons monochromatiques de 0,0322 eV, 0,0626 eV et 0,275 eV ont ete employes. Les detecteurs utilises sont un scintillateur gazeux rempli de xenon et une chambre d'ionisation a etat solide a basse geometrie. Les deux ont mesure les taux des desintegrations alpha et des fissions. Les resultats a l'energie de reference de 0,0253 eV sont: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (auteurs)
Measurements of the neutron-induced fission cross sections of 240Pu and 242Pu relative to 235U

International Nuclear Information System (INIS)

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

1976-01-01

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

International Nuclear Information System (INIS)

Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.

1978-01-01

Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)
Measurement of the neutron-induced fission cross section of 237Np relative to 235U from 0.02 to 30 MeV

International Nuclear Information System (INIS)

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

1977-01-01

The 237 Np/ 235 U fission cross section ratio has been measured from 0.02 to 30 MeV. Using the threshold method, a value of 1.294 +- 0.019 is obtained for the average cross section ratio in the interval from 1.75 to 4.00 MeV
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
Determination of the fission-neutron averaged cross sections of some high-energy threshold reactions of interest for reactor dosimetry

International Nuclear Information System (INIS)

Arribere, M.A.; Kestelman, A.J.; Korochinsky, S.; Blostein, J.J.

2003-01-01

For three high threshold reactions, we have measured the cross sections averaged over a 235 U fission neutron spectrum. The measured reactions, and corresponding averaged cross sections found, are: 127 I(n,2n) 126 I, (1.36±0.12) mb; 90 Zr(n,2n) 89m Zr, (13.86±0.83) μb; and 58 Ni(n,d+np+pn) 57 Co, (274±15) μb; all referred to the well known standard of (111±3) mb for the 58 Ni(n,p) 58m+g Co averaged cross section. The measured cross sections are of interest in nuclear engineering for the characterization of the fast neutron component in the energy distribution of reactor neutrons. (author)
Experimental study of energy dependence of proton induced fission cross sections for heavy nuclei in the energy range 200-1000 MeV

Energy Technology Data Exchange (ETDEWEB)

Kotov, A.A.; Gavrikov, Yu.A.; Vaishnene, L.A.; Vovchenko, V.G.; Poliakov, V.V.; Fedorov, O.Ya.; Chestnov, Yu.A.; Shchetkovskiy, A.I [Petersburg Nuclear Physics Institute, Gatchina, Leningrad district, Orlova roscha 1, 188300 (Russian Federation); Fukahori, T. [Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki 319-1195 (Japan)

2005-07-01

The results of the total fission cross sections measurements for {sup nat}Pb, {sup 209}Bi, {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu nuclei at the energy proton range 200-1000 MeV are presented. Experiments were carried out at 1 GeV synchrocyclotron of Petersburg Nuclear Physics Institute (Gatchina). The measurement method is based on the registration in coincidence of both complementary fission fragments by two gas parallel plate avalanche counters, located at a short distance and opposite sides of investigated target. The insensitivity of parallel plate avalanche counters to neutron and light charged particles allowed us to place the counters together with target immediately in the proton beam providing a large solid angle acceptance for fission fragment registration and reliable identification of fission events. The proton flux on the target to be studied was determined by direct counting of protons by scintillation telescope. The measured energy dependence of the total fission cross sections is presented. Obtained results are compared with other experimental data as well as with calculation in the frame of the cascade evaporation model. (authors)
Intermediate structure studies of 234U cross sections

International Nuclear Information System (INIS)

James, G.D.; Schindler, R.H.

1976-01-01

Neutron induced fission and total cross sections of 234 U have been measured over the neutron energy range from a few eV to several MeV. Neutron and fission widths for 118 cross section resonances below 1500 eV have been determined and give a class I level spacing of 10.64 + -0.46 eV and a neutron strength function of (0.857 +- 0.108)x10 -4 . These fine structure resonances comprise a narrow intermediate structure resonance in the sub-threshold fission cross section of 234 U. Parameters for the Lorentzian energy dependence of the mean fission width are deduced on the assumption that, relative to this mean, the observed fission widths have a Porter-Thomas distribution. Two large fission widths measured for resonances at 1092.5 eV and 1134 eV may indicate the presence of a second narrow intermediate structure resonance at about this energy. The class II level spacing derived from the observation of 7 resonances below 13 keV is 2.1 +-0.3 keV. Pronounced breaks in the fission cross section at 310 keV, 550 keV and 720 keV are assumed to be due to β-vibrational levels in the second minimum of the Strutinsky potential. Fluctuations due to the presence of class II resonances are strongly evident for each of these vibrational levels. It is shown that the fluctuations near 310 keV are consistent with parameters deduced from the low energy data and this enables parameters for the double humped fission barrier potential to be obtained
Neutron cross sections of 28 fission product nuclides adopted in JENDL-1

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki; Nakagawa, Tsuneo; Igarasi, Sin-iti; Matsunobu, Hiroyuki; Kawai, Masayoshi; Iijima, Shungo.

1981-02-01

This is the final report concerning the evaluated neutron cross sections of 28 fission product nuclides adopted in the first version of Japanese Evaluated Nuclear Data Library (JENDL-1). These 28 nuclides were selected as being most important for fast reactor calculations, and are 90 Sr, 93 Zr, 95 Mo, 97 Mo, 99 Tc, 101 Ru, 102 Ru, 103 Rh, 104 Ru, 105 Pd, 106 Ru, 107 Pd, 109 Ag, 129 I, 131 Xe, 133 Cs, 135 Cs, 137 Cs, 143 Nd, 144 Ce, 144 Nd, 145 Nd, 147 Pm, 147 Sm, 149 Sm, 151 Sm, 153 Eu and 155 Eu. The status of the experimental data was reviewed over the whole energy range. The present evaluation was performed on the basis of the measured data with the aid of theoretical calculations. The optical and statical models were used for evaluation of the smooth cross sections. An improved method was developed in treating the multilevel Breit-Wigner formula for the resonance region. Various physical parameters and the level schemes, adopted in the present work are discussed by comparing with those used in the other evaluations such as ENDF/B-IV, CEA, CNEN-2 and RCN-2. Furthermore, the evaluation method and results are described in detail for each nuclide. The evaluated total, capture and inelastic scattering cross sections are compared with the other evaluated data and some recent measured data. Some problems of the present work are pointed out and ways of their improvement are suggested. (author)
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)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

International Nuclear Information System (INIS)

Soran, P.D.; Seamon, R.E.

1980-05-01

Graphs of all neutron cross sections and photon production cross sections on the Recommended Monte Carlo Cross Section (RMCCS) library have been plotted along with local neutron heating numbers. Values for anti ν, the average number of neutrons per fission, are also given
NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

International Nuclear Information System (INIS)

OH, S.Y.; CHANG, J.; MUGHABGHAB, S.

2000-01-01

Neutron cross section evaluations of the fission-product isotopes, 95 Mo, 99 Tc, 101 Ru, 103 Rh, 105 Pd, 109 Ag, 131 Xe, 133 Cs, 141 Pr, 141 Nd, 147 Sm, 149 Sm, 150 Sm, 151 Sm, 152 Sm, 153 Eu, 155 Gd, and 157 Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of 155 Gd and 157 Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations
Comparison of integral cross section values of several cross section libraries in the SAND-II format

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.

1978-01-01

A comparison of some integral cross section values for several cross section libraries in the SAND-II format is presented. The integral cross section values are calculated with aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross section libraries used have all the SAND-II format. (author)
Graphs of the cross sections in the Alternate Monte Carlo Cross Section library at the Los Alamos Scientific Laboratory

International Nuclear Information System (INIS)

Seamon, R.E.; Soran, P.D.

1980-06-01

Graphs of all neutron cross sections and photon production cross sections on the Alternate Monte Carlo Cross Section (AMCCS) library have been plotted along with local neutron heating numbers. The values of ν-bar, the average number of neutrons per fission, are also plotted for appropriate isotopes
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
Comparison of integral cross section values of several cross section libraries in the SAND-II format

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.

1976-09-01

A comparison of some integral cross-section values for several cross-section libraries in the SAND-II format is presented. The integral cross-section values are calculated with the aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross-section libraries used have all the SAND-II format. Discrepancies between cross-sections in the different libraries are indicated but not discussed
Fission cross sections of plutonium-239 and plutonium-242 relative to uranium-285 from 0. 1 to 10 MeV

Energy Technology Data Exchange (ETDEWEB)

Meadows, J.W.

1978-12-01

Measurement of the ratio of the fission cross sections of /sup 239/Pu and /sup 242/Pu to that of /sup 235/U is reported. The sources of neutrons were the /sup 7/Li(p,n)/sup 7/Be and D(d,n)/sup 3/He reactions. The ratio of the masses of the samples was determined by low geometry alpha counting and from relative thermal fission rates. The results are compared with other measurements. 19 references.
Measurements of thermal fission and capture cross sections of minor actinides within the Mini-INCA project

Energy Technology Data Exchange (ETDEWEB)

Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A.; Panebianco, S.; Veyssiere, Ch. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, 91- Gif sur Yvette (France); Oriol, L. [CEA Cadarache, Dept. d' Etudes des Reacteurs, 13 - Saint Paul lez Durance (France); Chartier, F. [CEA Saclay, Dept. de Physico-Chimie, 91 - Gif sur Yvette (France); Mutti, P. [Institut Laue Langevin, 38 - Grenoble, (France); AlMahamid, I. [Wadsworth Center, New York State Dept. of Health, Albany, NY (United States)

2008-07-01

In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at Cea/DSM to determine optimal conditions for transmutation and incineration of Minor Actinides in high intensity neutron fluxes in the thermal region. Our experimental tool is based on alpha- and gamma-spectroscopy of irradiated samples and microscopic fission-chambers. It can provide both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. The {sup 232}Th, {sup 237}Np, {sup 241}Am, and {sup 244}Cm transmutation chains have been explored in details, showing some discrepancies in comparison with evaluated data libraries but in overall good agreement with recent experimental data. (authors)
Measurements of thermal fission and capture cross sections of minor actinides within the Mini-INCA project

International Nuclear Information System (INIS)

Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A.; Panebianco, S.; Veyssiere, Ch.; Oriol, L.; Chartier, F.; Mutti, P.; AlMahamid, I.

2008-01-01

In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at Cea/DSM to determine optimal conditions for transmutation and incineration of Minor Actinides in high intensity neutron fluxes in the thermal region. Our experimental tool is based on alpha- and gamma-spectroscopy of irradiated samples and microscopic fission-chambers. It can provide both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. The 232 Th, 237 Np, 241 Am, and 244 Cm transmutation chains have been explored in details, showing some discrepancies in comparison with evaluated data libraries but in overall good agreement with recent experimental data. (authors)
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
Measurement of fission cross section with pure Am-243 sample using lead slowing-down spectrometer

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Katsuhei; Yamamoto, Shuji; Kai, T.; Fujita, Yoshiaki; Yamamoto, Hideki; Kimura, Itsuro [Kyoto Univ. (Japan); Shinohara, Nobuo

1997-03-01

By making use of back-to-back type double fission chambers and a lead slowing-down spectrometer coupled to an electron linear accelerator, the fission cross section for the {sup 243}Am(n,f) reaction has been measured relative to that for the {sup 235}U(n,f) reaction in the energy range from 0.1 eV to 10 keV. The measured result was compared with the evaluated nuclear data appeared in ENDF/B-VI and JENDL-3.2, whose evaluated data were broadened by the energy resolution function of the spectrometer. General agreement was seen between the evaluated data and the measurement except that the ENDF/B-VI data were lower in the range from 15 to 60 eV and that the JENDL-3.2 data seemed to be lower above 100 eV. (author)
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)
Interim report on research between Oak Ridge National Laboratory and Japan Nuclear Cycle Development Institute on neutron-capture cross sections by long-lived fission product nuclides

International Nuclear Information System (INIS)

Furutaka, Kazuyoshi; Nakamura, Shoji; Harada, Hideo

2004-03-01

Neutron capture cross sections of long-lived fission products (LLFP) are important quantities as fundamental data for the study of nuclear transmutation of radioactive wastes. Previously obtained thermal-neutron capture gamma-ray data were analyzed to deduce the partial neutron-capture cross sections of LLFPs including 99 Tc, 93 Zr, and 107 Pd for thermal neutrons. By comparing the decay gamma-ray data and prompt gamma-ray data for 99 Tc, the relation between the neutron-capture cross section deduced by the two different methods was studied. For the isotopes 93 Zr and 107 Pd, thermal neutron-capture gamma-ray production cross sections were deduced for the first time. The level schemes of 99 Tc, 93 Zr, and 107 Pd have also been constructed form the analyzed data and compared with previously reported levels. This work has been done under the cooperative program 'Neutron Capture Cross Sections of Long-Lived Fission products (LLFPs)' by Japan Nuclear Cycle Development Institute (JNC) and Oak Ridge National Laboratory (ORNL). (author)
Fission theory and actinide fission data

Energy Technology Data Exchange (ETDEWEB)

Michaudon, A.

1975-06-01

The understanding of the fission process has made great progress recently, as a result of the calculation of fission barriers, using the Strutinsky prescription. Double-humped shapes were obtained for nuclei in the actinide region. Such shapes could explain, in a coherent manner, many different phenomena: fission isomers, structure in near-threshold fission cross sections, intermediate structure in subthreshold fission cross sections and anisotropy in the emission of the fission fragments. A brief review of fission barrier calculations and relevant experimental data is presented. Calculations of fission cross sections, using double-humped barrier shapes and fission channel properties, as obtained from the data discussed previously, are given for some U and Pu isotopes. The fission channel theory of A. Bohr has greatly influenced the study of low-energy fission. However, recent investigation of the yields of prompt neutrons and γ rays emitted in the resonances of {sup 235}U and {sup 239}Pu, together with the spin determination for many resonances of these two nuclei cannot be explained purely in terms of the Bohr theory. Variation in the prompt neutron and γ-ray yields from resonance to resonance does not seem to be due to such fission channels, as was thought previously, but to the effect of the (n,γf) reaction. The number of prompt fission neutrons and the kinetic energy of the fission fragments are affected by the energy balance and damping or viscosity effects in the last stage of the fission process, from saddle point to scission. These effects are discussed for some nuclei, especially for {sup 240}Pu.
NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

Energy Technology Data Exchange (ETDEWEB)

OH,S.Y.; CHANG,J.; MUGHABGHAB,S.

2000-05-11

Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.
Study of U235 neutron fission spectrum by the knowledge of cross sections average over that spectrum

International Nuclear Information System (INIS)

Suarez, P.M.

1997-01-01

A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the 45 Sc(n,2n) 44g Sc and 45 Sc(n,2n) 44m Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, 235 U. The results of the calculations mentioned above lead us to propose an analytical form for the 235 U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole energy range. These two new
ENEA-Bologna production and testing of Jeff-3.1 multi-group cross section libraries for nuclear fission applications

International Nuclear Information System (INIS)

Pescarini, M.; Orsi, R.; Sinitsa, V.

2008-01-01

The ENEA-Bologna Nuclear Data Group produced the JEFF-3.1 VITJEFF31.BOLIB and MATJEFF31. BOLIB fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format, with the same specifications and energy group structure of the Endf/B-VI-3 VITAMIN-B6 American library. Each library, containing 181 nuclide cross section files, was generated from the same set of cross section data files in GENDF format, obtained through the Bondarenko (f-factor) method, with an ENEA-Bologna revised version of the GROUPR module of the NJOY-99.160 system. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the DANTSYS and DOORS systems, can be generated from VITJEFF31.BOLIB and MATJEFF31.BOLIB through, respectively, further data processing with an ENEA-Bologna revised version of the SCAMPI system and with the TRANSX code. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF31.BOLIB validation. (authors)
Measurements of prompt fission neutron spectra and double-differential neutron inelastic-scattering cross sections for 238U and 232Th

International Nuclear Information System (INIS)

Baba, Mamoru; Itoh, Nobuo; Maeda, Kazuto; Hirakawa, Naohiro; Wakabayashi, Hidetaka.

1989-10-01

This report presents the summary of experimental studies of prompt fission neutron spectra and double-differential neutron inelastic-scattering cross sections of 238 U and 232 Th. The experiments were performed at Tohoku University Fast Neutron Laboratory employing a time-of-flight technique and Dynamitron accelerator as the pulsed neutron generator. From the experiments, we obtained the following data for both nuclei; 1. prompt fission neutron spectrum for 2 MeV neutrons, 2. double-differential neutron inelastic-scattering cross sections for 1.2, 2.0, 4.2, 6.1 and 14.1 MeV incident neutrons. Both in experiments and data processing, cares were taken to obtain reliable data by avoiding systematic uncertainty. The experimental data were compared with those by other experiments, evaluations and model calculations. Through the data comparison, some fundamental problems were found in the experiments by previous authors and the evaluations. The present data will provide useful data base for refinement of the evaluated data and theoretical models. (author)
Measurement of the fission cross-section of $^{240}$Pu and $^{242}$Pu at CERN's n_TOF Facility

CERN Multimedia

Pavlik, A F; Gonzalez romero, E M

The n_TOF Collaboration proposes to continue the fission program, already started in 2002-2004, taking advantage of the newly constructed Work Sector Type A, with the measurement of the two isotopes : $^{240}$ Pu and $^{242}$ Pu. They are both of major importance for reactor physics applications and are included in the Nuclear Energy Agency (NEA) High Priority List [1], in the NEA WPEC Subgroup 26 Report on the accuracy of nuclear data for advanced reactor designe [2] and in the EU 6$^{th}$ Framework Programme IP-EUROTRANS/NUDATRA reports [3]. Based on those requests, the measurement of the fission cross-section of the two Pu isotopes is one of the objectives of the project ANDES of the FP7 EURATOM program [4].

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.
Average cross-sections for /n, p/ reactions on calcium in a fission-type reactor spectrum

International Nuclear Information System (INIS)

Bruggeman, A.; Maenhaut, W.; Hoste, J.

1974-01-01

The average cross-section in a fission-type reactor spectrum sigmasub(F) was experimentally determined for the reactions 42 Ca/n,p/ 42 K, 43 Ca/n,p/ 43 K and 44 Ca/n,p/ 44 K. Calcium carbonate samples and fast neutron flux monitors were irradiated with and without cadmium shielding in the Thetis reactor (Institute for Nuclear Sciences, Rijksuniversiteit Gent). The potassium activities induced in the calcium carbonate samples were separated and purified by tetraphenylborate precipitation, after which they were measured with a Ge/Li/-detector of calibrated detection efficiency. On the basis of sigmasub(F)=0.64 mb for the reaction 27 Al/n,α/ 24 Na, the average cross-sections were as follows: 42 Ca/n,p/ 42 K: 2.82+-0.07 mb; 43 Ca/n,p/ 43 K: 1.89+-0.05 mb; 44 Ca/n,p/ 44 K: 0.065+-0.003 mb. (T.G.)
Single-level resonance parameters fit nuclear cross-sections

Science.gov (United States)

Drawbaugh, D. W.; Gibson, G.; Miller, M.; Page, S. L.

1970-01-01

Least squares analyses of experimental differential cross-section data for the U-235 nucleus have yielded single level Breit-Wigner resonance parameters that fit, simultaneously, three nuclear cross sections of capture, fission, and total.
R-matrix analysis of the /sup 239/Pu neutron cross sections

Energy Technology Data Exchange (ETDEWEB)

Saussure, G. de; Perez, R.B.; Macklin, R.L.

1986-03-01

/sup 239/Pu neutron cross-section data in the resolved resonance region were analyzed with the R-Matrix Bayesian Program SAMMY. Below 30 eV the cross sections computed with the multilevel parameters are consistent with recent fission and transmission measurements as well as with older capture and alpha measurements. Above 30 eV no suitable transmission data were available and only fission cross-section measurements were analyzed. However, since the analysis conserves the complete covariance matrix, the analysis can be updated by the Bayes method as transmission measurements become available. To date, the analysis of the fission measurements has been completed up to 300 eV.
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-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.
Status of pseudo fission product cross sections for fast reactors. Results of the SWG 17, International working party on evaluation coordination of the nuclear science committee, NEA- OECD

International Nuclear Information System (INIS)

Gruppelaar, H.; Kloosterman, J.L.; Pijlgroms, B.J.; Rimpault, G.; Smith, P.; Ignatyuk, A.; Koshcheev, V.; Nikolaev, M.; Thsiboulia, A.; Kawai, M.; Nakagawa, T.; Watanabe, T.; Zukeran, A.; Nakajima, Y.; Matsunobu, H.

1998-08-01

Within the framework of the SWG17 benchmark organized by a Working Party of the Nuclear Science Committee of the Nuclear Energy Agency (NEA), a comparison of lumped or pseudo fission product cross sections for fast reactors has been made. Four institutions participated with data libraries based on the JEF2.2, EAF-4.2, BROND-2, FONDL-2.1, ADL-3 and JENDL-3.2 evaluated nuclear data files. Several parameters have been compared with each other: the one-group cross sections and reactivity worths of the lumped nuclide for several partial absorption and scattering cross sections, and the one-group cross sections of the individual fission products. Also graphs of the multi-group cross sections of the lumped nuclide have been compared, as well as graphs of capture cross sections for 27 nuclides. From two contributions based on JEF2.2, it can be concluded that the data processing influences the capture cross section by about 1% and the inelastic scattering cross section by 2%. The differences between the lumped cross sections of the different data libraries are surprisingly small: maximum 6% for capture and 9% for the inelastic scattering. Similar results are obtained for the reactivity effects. Since the reactivity worth of the lumped nuclide is dominated by the capture reaction, the maximum spread in the total reactivity worth is still only 5.3%. There is a systematic difference between total, elastic and capture cross sections of JENDL-3.2 and JEF2.2 of the same order of magnitude. Possible reasons for this discrepancy have been indicated. The one-group capture and inelastic scattering cross sections of most of the important individual fission products differ by less than 10% (root mean square values). Larger differences are observed for unstable nuclides where there is a lack of experimental data. For the (n,2n) group cross sections, which are rather sensitive to the weighting spectrum in the fast energy range, these differences are several tens of percents. The final
Simultaneous fitting of statistical-model parameters to symmetric and asymmetric fission cross sections

International Nuclear Information System (INIS)

Mancusi, D; Charity, R J; Cugnon, J

2013-01-01

The de-excitation of compound nuclei has been successfully described for several decades by means of statistical models. However, accurate predictions require some fine-tuning of the model parameters. This task can be simplified by studying several entrance channels, which populate different regions of the parameter space of the compound nucleus. Fusion reactions play an important role in this strategy because they minimise the uncertainty on the entrance channel by fixing mass, charge and excitation energy of the compound nucleus. If incomplete fusion is negligible, the only uncertainty on the compound nucleus comes from the spin distribution. However, some de-excitation channels, such as fission, are quite sensitive to spin. Other entrance channels can then be used to discriminate between equivalent parameter sets. The focus of this work is on fission and intermediate-mass-fragment emission cross sections of compound nuclei with 70 70 ≲ A ≲ 240. 240. The statistical de-excitation model is GEMINI++. The choice of the observables is natural in the framework of GEMINI++, which describes fragment emission using a fissionlike formalism. Equivalent parameter sets for fusion reactions can be resolved using the spallation entrance channel. This promising strategy can lead to the identification of a minimal set of physical ingredients necessary for a unified quantitative description of nuclear de-excitation.
Model cross section calculations using LAHET

International Nuclear Information System (INIS)

Prael, R.E.

1992-01-01

The current status of LAHET is discussed. The effect of a multistage preequilibrium exciton model following the INC is examined for neutron emission benchmark calculations, as is the use of a Fermi breakup model for light nuclei rather than an evaporation model. Comparisons are made also for recent fission cross section experiments, and a discussion of helium production cross sections is presented
Spallation reaction study for fission products in nuclear waste: Cross section measurements for {sup 137}Cs and {sup 90}Sr on proton and deuteron

Energy Technology Data Exchange (ETDEWEB)

Wang, H., E-mail: wanghe@ribf.riken.jp [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Otsu, H.; Sakurai, H.; Ahn, D.S. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Aikawa, M. [Faculty of Science, Hokkaido University, Sapporo 060-0810 (Japan); Doornenbal, P.; Fukuda, N.; Isobe, T. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Kawakami, S. [Department of Applied Physics, University of Miyazaki, Miyazaki 889-2192 (Japan); Koyama, S. [Department of Physics, University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-0033 (Japan); Kubo, T.; Kubono, S.; Lorusso, G. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Maeda, Y. [Department of Applied Physics, University of Miyazaki, Miyazaki 889-2192 (Japan); Makinaga, A. [Graduate School of Medicine, Hokkaido University, North-14, West-5, Kita-ku, Sapporo 060-8648 (Japan); Momiyama, S. [Department of Physics, University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-0033 (Japan); Nakano, K. [Department of Advanced Energy Engineering Science, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Niikura, M. [Department of Physics, University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-0033 (Japan); Shiga, Y. [Department of Physics, Rikkyo University, 3-34-1 Nishi-Ikebukuro, Toshima, Tokyo 171-8501 (Japan); RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Söderström, P.-A. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); and others

2016-03-10

We have studied spallation reactions for the fission products {sup 137}Cs and {sup 90}Sr for the purpose of nuclear waste transmutation. The spallation cross sections on the proton and deuteron were obtained in inverse kinematics for the first time using secondary beams of {sup 137}Cs and {sup 90}Sr at 185 MeV/nucleon at the RIKEN Radioactive Isotope Beam Factory. The target dependence has been investigated systematically, and the cross-section differences between the proton and deuteron are found to be larger for lighter spallation products. The experimental data are compared with the PHITS calculation, which includes cascade and evaporation processes. Our results suggest that both proton- and deuteron-induced spallation reactions are promising mechanisms for the transmutation of radioactive fission products.
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.
Measurement of isotopic cross sections of the fission fragments produced in 500 AMeV {sup 208}Pb + p reaction; Etude de la production des fragments de fission issus de la reaction {sup 208}Pb + p a 500 AMeV

Energy Technology Data Exchange (ETDEWEB)

Fernandez-Dominguez, B

2003-03-01

The aim of this work is the study of the fission fragments produced in the spallation reaction {sup 208}Pb + p at 500 AMeV. The fission fragments from Z=23 up to Z=59 have been detected and identified by using the inverse kinematics technique with the high-resolution spectrometer FRS. The production cross sections and the recoil velocities of 430 nuclei have been measured. The measured data have been compared with previous data. The isotopic distributions show a high precision. However, the absolute value of the fission cross section is higher than expected. From the experimental data the characteristics of the average fissioning system have been reconstructed (Z{sub fis}, A{sub fis}, E*{sub fis}). In addition, the number of post-fission neutrons emitted from the fission fragments, v{sub post}, has been determined by using a new method. The experimental data have been compared to the two-steps models describing the spallation reaction. The impact of the model parameters on the observables has been analysed and the reasons Leading to the observed differences between the codes are also presented. This analyse shows a good agreement with the INCL4+ABLA code. (author)
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.
The fission cross section ratios and error analysis for ten thorium, uranium, neptunium and plutonium isotopes at 14.74 MeV neutron energy

International Nuclear Information System (INIS)

Meadows, J.W.

1987-03-01

The error information from the recent measurements of the fission cross section ratios of nine isotopes, 230 Th, 232 Th, 233 U, 234 U, 236 U, 238 U, 237 Np, 239 Pu, and 242 Pu, relative to 235 U at 14.74 MeV neutron energy was used to calculate their correlations. The remaining 36 non-trivial and non-reciprocal cross section ratios and their errors were determined and compared to evaluated (ENDF/B-V) values. There are serious differences but it was concluded that the reduction of three of the evaluated cross sections would remove most of them. The cross sections to be reduced are 230 Th - 13%, 237 Np - 9.6% and 239 Pu - 7.6%. 5 refs., 6 tabs
Remarks concerning the accurate measurement of differential cross sections for threshold reactions used in fast-neutron dosimetry for fission reactors

International Nuclear Information System (INIS)

Smith, D.L.

1976-12-01

Some remarks are submitted concerning the measurement of differential cross sections for threshold reactions which are used in fast-neutron dosimetry for fission reactors. The objective is to familiarize the reader with some of the problems associated with these measurements and, in the process, to explain why the existence of large discrepancies in the data sets for many of these reactions is not surprising. Limits to the accuracy which can be expected for these cross sections in the near future--using current technology and available resources--are examined in a general way and recommendations for improving the accuracy of the differential data base for dosimetry reactions are presented
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
Study of the variation with the energy of the fission cross-sections of 233U, 235U, 239Pu for the fast neutrons

International Nuclear Information System (INIS)

Szteinsznaider, D.; Naggiar, V.; Netter, F.

1955-01-01

This measurements have been done while taking the value of the fission cross-sections of 238 U as reference. The neutrons are produced by the reaction 7 Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for 239 Pu: σ f = 2,04 ± 0,12 barns, cross-section constant between 150 and 2000 keV, for 235 U: σ f = 1,15 ± 0,15 barns, cross-section constant between 700 and 1000 keV, for 233 U: σ f = 1,92 ± 0,25 barns, for neutrons of 850 keV. (authors) [fr
Validation of Cross Sections with Criticality Experiment and Reaction Rates: the Neptunium Case

CERN Document Server

Leong, L S; Audouin, L; Berthier, B; Le Naour, C; Stéphan, C; Paradela, C; Tarrío, D; Duran, I

2014-01-01

The Np-237 neutron-induced fission cross section has been recently measured in a large energy range (from eV to GeV) at the n\\_TOF facility at CERN. When compared to previous measurements the n\\_TOF fission cross section appears to be higher by 5-7\\% beyond the fission threshold. To check the relevance of the n\\_TOF data, we considered a criticality experiment performed at Los Alamos with a 6 kg sphere of Np-237, surrounded by uranium highly enriched in U-235 so as to approach criticality with fast neutrons. The multiplication factor k(eff) of the calculation is in better agreement with the experiment when we replace the ENDF/B-VII. 0 evaluation of the Np-237 fission cross section by the n\\_TOF data. We also explored the hypothesis of deficiencies of the inelastic cross section in U-235 which has been invoked by some authors to explain the deviation of 750 pcm. The large modification needed to reduce the deviation seems to be incompatible with existing inelastic cross section measurements. Also we show that t...
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)
Triple-humped fission barrier model for a new {sup 238}U neutron cross-section evaluation and first validations

Energy Technology Data Exchange (ETDEWEB)

Lopez Jimenez, M.J. [CEA/DIF/DPTA/Service de Physique Nucleaire, B.P. 12, F-91680 Bruyeres-le-Chatel (France); Morillon, B. [CEA/DIF/DPTA/Service de Physique Nucleaire, B.P. 12, F-91680 Bruyeres-le-Chatel (France); Romain, P. [CEA/DIF/DPTA/Service de Physique Nucleaire, B.P. 12, F-91680 Bruyeres-le-Chatel (France)]. E-mail: pascal.romain@cea.fr

2005-01-15

A new neutron-induced cross-section evaluation of {sup 238}U from 1 keV up to 200 MeV has been performed using only nuclear reactions models. A new fission penetrability model taking into account a triple humped barrier has been developed. A clear improvement has been observed for K-effective validation tests (up to 30 MeV) with this new evaluation. This improvement is mainly due to a better treatment of the inelastic exit channel.

The fission cross section ratios and error analysis for ten thorium, uranium, neptunium and plutonium isotopes at 14. 74 MeV neutron energy

Energy Technology Data Exchange (ETDEWEB)

Meadows, J.W.

1987-03-01

The error information from the recent measurements of the fission cross section ratios of nine isotopes, /sup 230/Th, /sup 232/Th, /sup 233/U, /sup 234/U, /sup 236/U, /sup 238/U, /sup 237/Np, /sup 239/Pu, and /sup 242/Pu, relative to /sup 235/U at 14.74 MeV neutron energy was used to calculate their correlations. The remaining 36 non-trivial and non-reciprocal cross section ratios and their errors were determined and compared to evaluated (ENDF/B-V) values. There are serious differences but it was concluded that the reduction of three of the evaluated cross sections would remove most of them. The cross sections to be reduced are /sup 230/Th - 13%, /sup 237/Np - 9.6% and /sup 239/Pu - 7.6%. 5 refs., 6 tabs.
Analysis of dependence of fission cross section and angular anisotropy of the 235U fission fragment escape induced by neutrons of intermediate energies (epsilon < or approximately200 keV) on target nucleus orientation

International Nuclear Information System (INIS)

Barabanov, A.L.

1985-01-01

Experimental data on dependence of fission cross section Σsub(f) (epsilon) and angular anisotropy W(epsilon, 0 deg)/W(epsilon, 90 deg) of sup(235)U fission fragment escape by neutrons with energy epsilon=100 and 200 keV on orientation of target nuclei are analyzed. 235 U (Isup(πsub(0))=7/2sup(-)) nuclei were orientated at the expense of interaction of quadrupole nucleus momenta with nonuniform electric field of uranyl-rubidium nitrate crystal at crystal cooling to T=0.2 K. The analysis was carried out with three different sets of permeability factors T(epsilon). Results of the analysis weakly depend on T(epsilon) choice. It is shown that a large number of adjusting parameters (six fissionabilities γsup(f)(Jsup(π), epsilon) and six momenta sub(Jsup(π))) permit to described experimental data on Σsub(f)(epsilon) and W(epsilon, 0 deg)/W(epsilon, 90 deg), obtained at epsilon=200 keV by introducing essential dependence of γsup(f)(Jsup(π), epsilon) and sub(Jsup(π)) on Jsup(π). Estimations of fission cross sections Σsub(f)(epsilon) and angular distribution W(epsilon, n vector) up to T approximately equal to 0.01 K in two geometries of the experiment: the orientation axis is parallel and perpendicular to momentum direction p vector of incident neutrons, are conducted
Burn-up physics in a coupled Hammer-Technion/Cinder-2 system and ENDF/B-V aggregate fission product thermal cross section validation

International Nuclear Information System (INIS)

Santos, A. dos.

1990-01-01

The new methodology developed in this work has the following purposes: a) to implement a burnup capability into the HAMMER-TECHNION/9/computer code by using the CINDER-2/10/computer code to perform the transmutation analysis for the actinides and fission products; b) to implement a reduced version of the CINDER-2 fission product chain structure to treat explicity nearly 99% of all original CINDER-2 fission product absorption in a typical PWR unit cell; c) to treat the effect of the fission product neutron absorption in an unit cell in a multigroup basis; d) to develop a tentative validation procedure for the ENOF/C-V stable and long-lived fission product nuclear data based on the available experimental data/11-14/. The analysis will be performed by using the reduce chain in the coupled system CINDER-2 to generate the time dependent effective four group cross sections for actinides and fission products and CINDER-2 to perform the complete transmutation analysis with its built-in chain structure. (author)
Neutron cross sections for uranium-235 (ENDF/B-IV Release 3)

International Nuclear Information System (INIS)

Lubitz, C.R.

1996-09-01

The resonance parameters in ENDF6 (Release 2) U235 were adjusted to make the average capture and fission cross sections below 900 eV agree with selected differential capture and fission measurements. The measurements chosen were the higher of the credible capture measurements and the lower of the fission results, yielding a higher epithermal alpha. In addition, the 2200 m/s cross sections were adjusted to obtain agreement with the integral value of K1. As a result, criticality calculations for thermal benchmarks, and agreement with a variety of integral parameters, are improved
Evaluation of the (n,xn) and (n,xnf) cross sections for heavy nuclei with the statistical model

International Nuclear Information System (INIS)

Jary, J.

1975-01-01

A method was presented to calculate the (n,xn) and (n,xnf) cross sections for the heavy nuclei having mass numbers of 232 1) without fission, according to the law of conventional statistical models, in the (n,xn) process. Fission can also compete with the emission of neutrons and γ-ray for the nuclei and the excitation energy considered. The fission cross sections of 235 U and 238 U recently evaluated by Sowerby and the fission cross section of 236 U have been used to determine the other parameters needed in the calculation. The fission widths of 239 U and 238 U have been obtained by fitting the first-chance and second-chance fission plateaus of the 238 U cross section. For the fission width of 238 U, good agreement was observed between the authors' results and Landrum and others' experimental data. (Iwase, T.)
Measurement of cross-sections of fission reactions induced by neutrons on actinides from the thorium cycle at n-TOF facility

International Nuclear Information System (INIS)

Ferrant, L.

2005-09-01

In the frame of innovating energy source system studies, thorium fuel cycle reactors are considered. Neutron induced fission cross section on such cycle involved actinides play a role in scenario studies. To feed them, data bases are built with experimental results and nuclear models. For some nuclei, they are not complete or in disagreement. In order to complete these data bases, we have built an original set up, consisting in an alternation of PPACs (Parallel Plate Avalanche Chamber) and ultra - thin targets, which we installed on n-TOF facility. We describe detectors, set up, and the particular care brought to target making and characterization. Fission products in coincidence are detected with precise time measurement and localization with delay line read out method. We contributed, within the n-TOF collaboration, to the CERN brand new intense spallation neutron source characterization, based on time of flight measurement, and we describe its characteristics and performances. We were able to measure such actinide fission cross sections as 232 Th, 234 U, 233 U, 237 Np, 209 Bi, and nat Pb relative to 235 U et 238 U standards, using an innovative acquisition system. We took advantage of the lame accessible energy field, from 0.7 eV to 1 GeV, combined with the excellent energy resolution in this field. Data treatment and analysis advancement are described to enlighten performance and limits of the obtained results. (author)
Surrogate measurement of the 238Pu(n,f) cross section

International Nuclear Information System (INIS)

Ressler, J. J.; Burke, J. T.; Escher, J. E.; Bernstein, L. A.; Bleuel, D. L.; Casperson, R. J.; Gostic, J.; Henderson, R.; Scielzo, N. D.; Thompson, I. J.; Wiedeking, M.; Angell, C. T.; Goldblum, B. L.; Munson, J.; Basunia, M. S.; Phair, L. W.; Beausang, C. W.; Hughes, R. O.; Hatarik, R.; Ross, T. J.

2011-01-01

The neutron-induced fission cross section of 238 Pu was determined using the surrogate ratio method. The (n,f) cross section over an equivalent neutron energy range 5-20 MeV was deduced from inelastic α-induced fission reactions on 239 Pu, with 235 U(α,α ' f) and 236 U(α,α ' f) used as references. These reference reactions reflect 234 U(n,f) and 235 U(n,f) yields, respectively. The deduced 238 Pu(n,f) cross section agrees well with standard data libraries up to ∼10 MeV, although larger values are seen at higher energies. The difference at higher energies is less than 20%.
Spallation reaction study for fission products in nuclear waste: Cross section measurements for 137Cs, 90Sr and 107Pd on proton and deuteron

Directory of Open Access Journals (Sweden)

Wang He

2017-01-01

Full Text Available Spallation reactions for the long-lived fission products 137Cs, 90Sr and 107Pd have been studied for the purpose of nuclear waste transmutation. The cross sections on the proton- and deuteron-induced spallation were obtained in inverse kinematics at the RIKEN Radioactive Isotope Beam Factory. Both the target and energy dependences of cross sections have been investigated systematically. and the cross-section differences between the proton and deuteron are found to be larger for lighter fragments. The experimental data are compared with the SPACS semi-empirical parameterization and the PHITS calculations including both the intra-nuclear cascade and evaporation processes.
Fission cross-section normalization problems

International Nuclear Information System (INIS)

Wagemans, C.; Ghent Rijksuniversiteit; Deruytter, A.J.

1983-01-01

The present measurements yield σsub(f)-data in the neutron energy from 20 MeV to 30 keV directly normalized in the thermal region. In the keV-region these data are consistent with the absolute σsub(f)-measurements of Szabo and Marquette. For the secondary normalization integral I 2 values have been obtained in agreement with those of Gwin et al. and Czirr et al. which were also directly normalized in the thermal region. For the I 1 integral, however, puzzling low values have been obtained. This was also the case for σsub(f)-bar in neutron energy intervals containing strong resonances. Three additional measurements are planned to further investigate these observations: (i) maintaining the actual approx.2π-geometry but using a 10 B-foil for the neutron flux detection (ii) using a low detection geometry with a 10 B- as well as a 6 Li-flux monitor. Only after these measurements definite conclusions on the I 1 and I 2 integrals can be formulated and final σsub(f)-bar-values can be released. The present study also gives some evidence for a correlation between the integral I 2 and the neutron flux monitor used. The influence of a normalization via I 1 or I 2 on the final cross-section has been shown. The magnitude of possible normalization errors is illustrated. Finally, since 235 U is expected to be an ''easy'' nucleus (low α-activity high σsub(f)-values), there are some indications that the important discrepancies still present in 235 U(n,f) cross-section measurements might partially be due to errors in the neutron flux determination
Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

International Nuclear Information System (INIS)

2009-01-01

This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations
Calculation of the Reaction Cross Section for Several Actinides

International Nuclear Information System (INIS)

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

2005-01-01

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

International Nuclear Information System (INIS)

Khan, H.A.; Khan, N.A.; Peterson, R.J.

1991-01-01

Fission cross sections for a range of targets have been measured by solid-state track detectors following 80 and 100 MeV π + bombardment. Fission probabilities have been inferred by comparison to computed reaction cross sections. Fission probabilities for heavy targets agree with those for other probes of comparable energy and with statistical calculations. Probabilities for lighter targets are much above those previously observed or computed. Ternary fission cross sections and multiplicities of light fragments have also been determined
Measurements of neutron cross sections of radioactive waste nuclides

Energy Technology Data Exchange (ETDEWEB)

Katoh, Toshio [Gifu College of Medical Technology, Seki, Gifu (Japan); Harada, Hideo; Nakamura, Shoji; Tanase, Masakazu; Hatsukawa, Yuichi

1998-01-01

Accurate nuclear reaction cross sections of radioactive fission products and transuranic elements are required for research on nuclear transmutation methods in nuclear waste management. Important fission products in the nuclear waste management are {sup 137}Cs, {sup 135}Cs, {sup 90}Sr, {sup 99}Tc and {sup 129}I because of their large fission yields and long half-lives. The present authors have measured the neutron capture cross sections and resonance integrals of {sup 137}Cs, {sup 90}Sr and {sup 99}Tc. The purpose of this study is to measure the neutron capture cross sections and resonance integrals of nuclides, {sup 129}I and {sup 135}Cs accurately. Preliminary experiments were performed by using Rikkyo University Reactor and JRR-3 reactor at Japan Atomic Energy Research Institute (JAERI). Then, it was decided to measure the cross section and resonance integral of {sup 135}Cs by using the JRR-3 Reactor because this measurement required a high flux reactor. On the other hand, those of {sup 129}I were measured at the Rikkyo Reactor because the product nuclides, {sup 130}I and {sup 130m}I, have short half-lives and this reactor is suitable for the study of short lived nuclide. In this report, the measurements of the cross section and resonance integral of {sup 135}Cs are described. To obtain reliable values of the cross section and resonance integral of {sup 135}Cs(n, {gamma}){sup 136}Cs reaction, a quadrupole mass spectrometer was used for the mass analysis of nuclide in the sample. A progress report on the cross section of {sup 134}Cs, a neighbour of {sup 135}Cs, is included in this report. A report on {sup 129}I will be presented in the Report on the Joint-Use of Rikkyo University Reactor. (author)
The fission cross sections of 230Th, 232Th, 233U, 234U, 236U, 238U, 237Np, 239Pu and 242Pu relative 235U at 14.74 MeV neutron energy

International Nuclear Information System (INIS)

Meadows, J.W.

1986-12-01

The measurement of the fission cross section ratios of nine isotopes relative to 235 U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for 235 U are: 230 Th - 0.290 +- 1.9%; 232 Th - 0.191 +- 1.9%; 233 U - 1.132 +- 0.7%; 234 U - 0.998 +- 1.0%; 236 U - 0.791 +- 1.1%; 238 U - 0.587 +- 1.1%; 237 Np - 1.060 +- 1.4%; 239 Pu - 1.152 +- 1.1%; 242 Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs
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
Average cross section measurements in U-235 fission neutron spectrum for some threshold reactions

International Nuclear Information System (INIS)

Maidana, N.L.

1993-01-01

The average cross section in the 235 U fission spectrum has been measured by the activation technique, for the following thresholds reactions: 115 In(n,n') 115m In, 232 Th(n,f) P.F., 46 , 47 , 48 Ti(n,p) 46,47 , 48 Sc, 55 Mn(n,2 n) 54 Mn, 51 V(n,α) 48 Sc, 90 Zr(n,2 n) 89 Zr, 93 Nb(n,2 n) 92m Nb, 58 Ni(n,2 n) 57 Ni, 24 Mg(n,p) 24 Na, 56 Fe(n,p) 56 Mn, 59 Co(n,α) 56 Mn and 63 Cu(n,α) 60 Co. The activation foils were irradiated close (∼ 4 mm) to the core of the IEA-R1 research reactor in the IPEN-CNEN/SP. The reactor was operated at 2 MW yielding a fast neutron flux around 5 x 10 12 n.cm -2 . s -1 . The neutron flux density was monitored by activation reactions with well known averaged cross sections and with effective thresholds above 1 MeV. The foil activities were measured in a calibrated HPGe spectrometer. The neutron spectrum has been calculated using the SAIPS unfolding system applied to the activation data. A detailed error analysis was performed using the covariance matrix methodology. The results were compared with those from other authors. (author)
Relativistic Coulomb Fission

Science.gov (United States)

Norbury, John W.

1992-01-01

Nuclear fission reactions induced by the electromagnetic field of relativistic nuclei are studied for energies relevant to present and future relativistic heavy ion accelerators. Cross sections are calculated for U-238 and Pu-239 fission induced by C-12, Si-28, Au-197, and U-238 projectiles. It is found that some of the cross sections can exceed 10 b.
Measurement of cross-sections of fission reactions induced by neutrons on actinides from the thorium cycle at n-TOF facility; Mesures de sections efficaces de fission induite par neutrons sur des actinides du cycle du thorium a n-TOF

Energy Technology Data Exchange (ETDEWEB)

Ferrant, L

2005-09-01

In the frame of innovating energy source system studies, thorium fuel cycle reactors are considered. Neutron induced fission cross section on such cycle involved actinides play a role in scenario studies. To feed them, data bases are built with experimental results and nuclear models. For some nuclei, they are not complete or in disagreement. In order to complete these data bases, we have built an original set up, consisting in an alternation of PPACs (Parallel Plate Avalanche Chamber) and ultra - thin targets, which we installed on n-TOF facility. We describe detectors, set up, and the particular care brought to target making and characterization. Fission products in coincidence are detected with precise time measurement and localization with delay line read out method. We contributed, within the n-TOF collaboration, to the CERN brand new intense spallation neutron source characterization, based on time of flight measurement, and we describe its characteristics and performances. We were able to measure such actinide fission cross sections as {sup 232}Th, {sup 234}U, {sup 233}U, {sup 237}Np, {sup 209}Bi, and {sup nat}Pb relative to {sup 235}U et {sup 238}U standards, using an innovative acquisition system. We took advantage of the lame accessible energy field, from 0.7 eV to 1 GeV, combined with the excellent energy resolution in this field. Data treatment and analysis advancement are described to enlighten performance and limits of the obtained results. (author)
Fission level densities

International Nuclear Information System (INIS)

Maslov, V.M.

1998-01-01

Fission level densities (or fissioning nucleus level densities at fission saddle deformations) are required for statistical model calculations of actinide fission cross sections. Back-shifted Fermi-Gas Model, Constant Temperature Model and Generalized Superfluid Model (GSM) are widely used for the description of level densities at stable deformations. These models provide approximately identical level density description at excitations close to the neutron binding energy. It is at low excitation energies that they are discrepant, while this energy region is crucial for fission cross section calculations. A drawback of back-shifted Fermi gas model and traditional constant temperature model approaches is that it is difficult to include in a consistent way pair correlations, collective effects and shell effects. Pair, shell and collective properties of nucleus do not reduce just to the renormalization of level density parameter a, but influence the energy dependence of level densities. These effects turn out to be important because they seem to depend upon deformation of either equilibrium or saddle-point. These effects are easily introduced within GSM approach. Fission barriers are another key ingredients involved in the fission cross section calculations. Fission level density and barrier parameters are strongly interdependent. This is the reason for including fission barrier parameters along with the fission level densities in the Starter File. The recommended file is maslov.dat - fission barrier parameters. Recent version of actinide fission barrier data obtained in Obninsk (obninsk.dat) should only be considered as a guide for selection of initial parameters. These data are included in the Starter File, together with the fission barrier parameters recommended by CNDC (beijing.dat), for completeness. (author)
Fission of 209 Bi by 60-270 MeV tagged photons: cross section measurement and analysis of photo fissility

International Nuclear Information System (INIS)

Terranova, M.L.; Tavares, O.A.P.

1996-07-01

Tagged photons produced by the ROKK-2 facility have been used to measure the photofission cross section of 209 Bi in the energy range 60-270 MeV. Photofission events were detected by using a nuclear fragment detector designed for fission experiments, based on multiwire spark counters. Fissility values have been deduced and compared with available data obtained in other laboratories by using monochromatic photons. These data, together with early measurements obtained near photofission threshold, have been analysed in the framework of a two-step model which considers the primary photo interaction occurring via the quasi-deuteron and/or photo mesonic processes, followed by a mechanism of evaporation-fission competition for the excited residual nucleus. The model was found to reproduce the main experimental features of 209 Bi photo fissility up to 300 MeV. (author). 52 refs., 7 figs., 2 tabs

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

International Nuclear Information System (INIS)

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

1988-01-01

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

International Nuclear Information System (INIS)

Kumawat, H.; Saxena, A.; Carminati, F.; )

2016-12-01

Monte Carlo neutron transport codes are one of the best tools to simulate complex systems like fission and fusion reactors, Accelerator Driven Sub-critical systems, radio-activity management of spent fuel and waste, optimization and characterization of neutron detectors, optimization of Boron Neutron Capture Therapy, imaging etc. The neutron cross-section and secondary particle emission properties are the main input parameters of such codes. The fission, capture and elastic scattering cross-sections have complex resonating structures. Evaluated Nuclear Data File (ENDF) contains these cross-sections and secondary parameters. We report the development of reconstruction procedure to generate point cross-sections and probabilities from ENDF data file. The cross-sections are compared with the values obtained from PREPRO and in some cases NJOY codes. The results are in good agreement. (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)
Model for fission-product calculations

International Nuclear Information System (INIS)

Smith, A.B.

1984-01-01

Many fission-product cross sections remain unmeasurable thus considerable reliance must be placed upon calculational interpolation and extrapolation from the few available measured cross sections. The vehicle, particularly for the lighter fission products, is the conventional optical-statistical model. The applied goals generally are: capture cross sections to 7 to 10% accuracies and inelastic-scattering cross sections to 25 to 50%. Comparisons of recent evaluations and experimental results indicate that these goals too often are far from being met, particularly in the area of inelastic scattering, and some of the evaluated fission-product cross sections are simply physically unreasonable. It is difficult to avoid the conclusion that the models employed in many of the evaluations are inappropriate and/or inappropriately used. In order to alleviate the above unfortunate situations, a regional optical-statistical (OM) model was sought with the goal of quantitative prediction of the cross sections of the lighter-mass (Z = 30-51) fission products. The first step toward that goal was the establishment of a reliable experimental data base consisting of energy-averaged neutron total and differential-scattering cross sections. The second step was the deduction of a regional model from the experimental data. It was assumed that a spherical OM is appropriate: a reasonable and practical assumption. The resulting OM then was verified against the measured data base. Finally, the physical character of the regional model is examined
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
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)
Isotopic production cross-sections and recoil velocities of spallation-fission fragments in the reaction 238U(1A GeV)+e

CERN Document Server

Pereira, J; Wlazlo, W; Benlliure, J; Casarejos, E; Armbruster, P; Bernas, M; Enqvist, T; Legrain, R; Leray, S; Rejmund, F; Mustapha, B; Schmidt, K.-H; Stéphan, C; Taïeb, J; Tassan-Got, L; Volant, C; Boudard, A; Czajkowski, S; 10.1103/PhysRevC.75.014602

2007-01-01

Fission fragments of 1A GeV 238U nuclei interacting with a deuterium target have been investigatedwith the Fragment Separator (FRS) at GSI (Darmstadt) by measuring their isotopicproduction cross-sections and recoil velocities. The results, along with those obtained recently forspallation-evaporation fragments, provide a comprehensive analysis of the spallation nuclear productionsin the reaction 238U(1A GeV)+d. Details about experiment performance, data reductionand results will be presented.
ENEA-Bologna production and testing of JEF-2.2 multi-group cross section libraries for nuclear fission applications

International Nuclear Information System (INIS)

Pescarini, M.; Orsi, R.; Martinelli, T.; Sinitsa, V.; Blokhin, A.I.

2005-01-01

The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section σ 0 . Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)
The fast fission effect in a cylindrical fuel element

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I; Pershagen, B

1959-06-15

A new formula for the fast fission factor is derived, which takes proper account to fast capture. The fission neutron spectrum is divided into two groups with constant fission cross section in one group and zero fission cross section in the other. The average total, elastic, inelastic and capture cross sections in the two groups are calculated. Different assumptions regarding anisotropic and inelastic scattering are investigated. The effects of backscattering from the moderator and fast fission in neighbouring fuel elements are pointed out. Formulas for the fast fission ratio and for the fast conversion ratio are derived. The calculated fast fission ratios are compared with experimental values. Curves are given for the fast fission factor in uranium metal and uranium oxide.
INTER, ENDF/B Thermal Cross-Sections, Resonance Integrals, G-Factors Calculation

International Nuclear Information System (INIS)

Dunford, Charles L.

2007-01-01

1 - Description of program or function: INTER calculates thermal cross sections, g-factors, resonance integrals, fission spectrum averaged cross sections and 14.0 MeV (or other energy) cross sections for major reactions in an ENDF-6 or ENDF-5 format data file. Version 7.01 (Jan 2005): set success flag after return from beginning. 2 - Method of solution: INTER performs integrations by using the trapezoidal rule
Consistent evaluation of neutron cross sections for the 242-244Cm isotopes

International Nuclear Information System (INIS)

Ignatyuk, A.V.; Maslov, V.M.

1989-01-01

The knowledge of neutron cross-sections for Curium isotopes is necessary for solving the problems of the external fuel cycle. Experimental information on the cross-sections is very meager and does not satisfy requirements and existing evaluations in different libraries differ substantially for fission and (n,2n) reaction cross-sections. This situation requires a critical review of the entire set of evaluations of the neutron cross-sections for Curium. 17 refs, 3 figs
Status of measured neutron cross sections of transactinium isotopes in the fast region

International Nuclear Information System (INIS)

Igarasi, S.

1976-01-01

This paper reviews present status of measured neutron cross sections of transactinium isotopes from a viewpoint of requested data in application field of the nuclear data. The measured cross sections from 1 keV to 15 MeV are examined. Comparison between different data sets is mainly performed on the fission cross sections
Measurement of the fission cross section of uranium-235 between 4 eV and 20 keV; Mesure de la section efficace de fission de l'uranium-235 entre 4 eV et 20 keV

Energy Technology Data Exchange (ETDEWEB)

Michaudon, A; Genin, R; Joly, R; Vendryes, G

1959-01-01

The neutron fission cross section of uranium-235 has been measured between 4 ev and 20 kev by the time of flight method with the Saclay electron linear accelerator as a pulsed neutron source. After a brief description of the experimental apparatus and the conditions of work during the experiment, the curve {sigma}{sub F} {radical}E in the energy range studied is shown. This curve is then analyzed by the ''area'' method and a set of {sigma}{sub 0} {gamma}{sub F} values is obtained. With {sigma}{sub 0} {gamma} values measured in other laboratories, it is possible to compute fission widths for several resonances and to study their distribution. This distribution is then compared to Porter-Thomas distributions with different values of the number of exit channels. (authors) [French] La section efficace de fission de l'uranium--235 a ete mesuree entre 4 eV et 20 KeV par la methode du temps de vol en utilisant l'accelerateur lineaire a electrons de Saclay comme source pulses de neutrons. Apres une rapide description de l'appareillage experimental et des conditions de fonctionnement au cours de l'experience, on presente la courbe {sigma}{sub F} {radical}E obtenue dans la game d'energie etudiee. Cette courbe est ensuite analysee par la methode de surface des resonances et un lot de valeurs de {sigma}{sub 0} {gamma}{sub F} est obtenue. Conjuguee avec les valeurs de {sigma}{sub 0} {gamma} obtenues dans d'autres laboratoires, cette analyse permet de calculer les largeurs de fission pour plusieurs resonances et d'etudier leur distribution. Cette distribution est ensuite comparee aux distributions de Porter et Thomas correspondant a differentes valeurs du nombre de voies de sortie. (auteurs)
The evaluation of the 237Np fission cross section in the 20 KeV - 20 MeV energy range

International Nuclear Information System (INIS)

Dushin, V.N.; Kalinin, V.A.; Shpakov, V.I.

1997-01-01

The results of the development of nuclear data evaluation based on the generalized least squares method is presented. The method to interpolate experimental data measured at arbitrary energy points, and their transfer to a fixed energy grid is described. The results of the 237 Np fission cross section measurements performed until 1988 were critically analyzed. A 781 x 781 covariant matrix was derived from the correlation analysis of the experimental results. The results of the evaluation, and the associated correlation matrix was obtained using the generalized least square method. (author). 34 refs, 4 figs, 2 tabs
Development of fine-group (315n/42γ) cross section library ENDL3.0/FG for fusion-fission hybrid systems

International Nuclear Information System (INIS)

Zeng Qin; Zou Jun; Xu Dezhen; Jiang Jieqiong; Wang Minghuang; Wu Yican; Qiu Yuefeng; Chen Zhong; Chen Yan

2011-01-01

To improve the accuracy of the neutron analyses for subcritical systems with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-Ⅶ. 0 has been produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. (authors)
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.)
High resolution measurement of the 237Np(n,f) cross section from 100keV to 2MeV

International Nuclear Information System (INIS)

Plattard, S.; Pranal, Y.; Blons, J.

1975-01-01

237 Np fission is one of the reactions utilized in the threshold detector method of neutron spectrometry for the determination of fast neutron spectra in nuclear reactors. Therefore, an accurate knowledge of the energy-dependent fission cross section is essential for a precise generation of the spectral indices of this method. A high resolution measurement of the fission cross section is presented [fr
CHEMICAL ENGINEERING DIVISION BURNUP, CROSS SECTIONS, AND DOSIMETRY SEMIANNUAL REPORT, JANUARY--JUNE 1972.

Energy Technology Data Exchange (ETDEWEB)

Larsen, R P; Dudey, N D; Crouthamel, C E; Tevebaugh, A D; Levenson, M; Vogel, R C

1972-09-01

Research and development efforts of the burnup, cross sections and dosimetry programs in the Chemical Engineering Division of Argonne National Laboratory are reported for the period January to June 1972. Work is reported in the following areas: (1) development of an X-ray spectrometric method for the determination of the rare-earth fission products and application of this method to the determinations of burnup in nuclear fuels; (2) determination of fast ·fission yields of bum up monitors and other fission products; (3) a search for a spon~aneously fissioning isomer of {sup 241}Pu; (4) measurements of the tritium and alpha particle yields in fast-neutron fission of {sup 235}U and {sup 239}Pu; (5) evaluations of available data on the differential cross sections for the {sup 56}Fe(n,p){sup 56}Mn and {sup 32}S(n,p){sup 32}P reactions; and (6) measurements of both fission rates by solid-state track recorders and reaction rates by foil activation, in the Coupled Fast Reactivity Measurement Facility.
Uranium, thorium and bismuth photofission cross sections at high energies

International Nuclear Information System (INIS)

Tavares, O.A.P.

1973-01-01

The U 238 , Th 232 and Bi 209 photofission using nuclear emulsion technique for fission fragments detection is presented. The photofission cross sections were measured using Bremsstrahlung photon which were produced irradiating thin tungsten radiators with electrons accelerated at the energy range from 1,0 to 5,5 GeV in the ''Deutsches Elektronen Synchrotron'' (Hamburg), and aluminium radiator with electrons accelarated at 16,0 GeV in Stanford Linear Accelerator Center. A special revelation technique for nuclear emulsion pellicles loaded with uranium and thorium, allowed the discrimination between alpha particles tracks and fission fragments tracks. The results show a decrease in the cross sections, which is in good agreement, within experimental errors, with the conclusions of other authors. The estimations from the two-step mechanism for high energy nuclear reactions (intranuclear cascade followed by fission-evaporation competition) show that, the primary interaction according to the photomesonic model and the quasi-deuteron photon interaction are sufficient to explain the general behavior exhibited by photofission cross sections for investigated nuclei. The calculations show a resonant structure around 300 MeV, with a width at half maximum of 200 MeV, and another not so pronounced, near to 700 MeV. (Author) [pt

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1984-09-01

This is the tenth issue of a report series on Fission Product Data, which informs us about all the activities in this field, which are planned, ongoing, or have recently been completed. The types of activities included are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission), neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products, lumped fission product data (decay heat, absorption, etc.). There is also a section with recent references relative to fission product nuclear data
Measurement of the fission cross section induced by fast neutrons of the {sup 232}Th/{sup 233}U nuclei within the innovating fuel cycles framework; Mesure de la section efficace de fission induite par neutrons rapides des noyaux {sup 232}Th/{sup 233}U dans le cadre des cycles de combustible innovants

Energy Technology Data Exchange (ETDEWEB)

Grosjean, C

2005-03-15

The thorium-U{sup 233} fuel cycle might provided safer and cleaner nuclear energy than the present Uranium/Pu fuelled reactors. Over the last 10 years, a vast campaign of measurements has been initiated to bring the precision of neutron data for the key nuclei (Th{sup 232}, Pa{sup 233} and U{sup 233}) at the level of those for the U-Pu cycle. This is the framework of these measurements, the energy dependent neutron induced fission cross section of Th{sup 232} and U{sup 233} has been measured from 1 to 7 MeV with a target accuracy lesser than 5 per cent. These measurements imply the accurate determination of the fission rate, the number of the target nuclei as well as the incident neutron flux impinging on the target, the latter has been obtained using the elastic scattering (n,p). The cross section of which is very well known in a large neutron energy domain ({approx} 0,5 % from 1 eV to 50 MeV) compared to the U{sup 235}(n,f) reaction. This technique has been applied for the first time to the Th{sup 232}(n,f) and U{sup 233}(n,f) cases. A Hauser-Feshbach statistical model has been developed. It consists of describing the different decay channels of the compound nucleus U{sup 234} from 0,01 to 10 MeV neutron energy. The parameters of this model were adjusted in order to reproduce the measured fission cross section of U{sup 233}. From these parameters, the cross sections from the following reactions could be extracted: inelastic scattering U{sup 233}(n,n'), radiative capture U{sup 233}(n,{gamma}) and U{sup 233}(n,2n). These cross sections are still difficult to measure by direct neutron reactions. The calculated values have allowed us to fill the lack of experimental data for the major fissile nucleus of the thorium cycle. (author)
Neutron total cross section measurements on 249Cf

International Nuclear Information System (INIS)

Carlton, R.F.; Harvey, J.A.; Hill, N.W.; Pandey, M.S.; Benjamin, R.W.

1979-01-01

Neutron total cross section measurements were performed on a sample of 249 Cf (5.65 mg total weight) with the ORELA as a source of pulsed neutrons. The sample, the inverse thickness of which was 1542 barns/atom, consisted of 85.3% 249 Cf and 14.4% 249 Bk, and was cooled to liquid nitrogen temperature. Analyses were also made of data from a thin sample (l/n = 17430) of 65% 249 Cf in the region of the large fission resonance at 0.7 eV. Fifty-five resonances in 249 Cf were observed and analyzed over the energy range 0.1 eV to 90 eV by use of an R-matrix multilevel formalism. The resonance parameters obtained were used to determine the level spacing and the s-wave neutron and fission strength functions. Thermal total cross section measurements were also performed. 5 figures, 3 tables
MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

Energy Technology Data Exchange (ETDEWEB)

Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(E_i), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after
Study of the variation with the energy of the fission cross-sections of {sup 233}U, {sup 235}U, {sup 239}Pu for the fast neutrons; Etude de la variation avec l'energie des sections efficaces de fission de {sup 233}U, {sup 235}U, {sup 239}Pu pour les neutrons rapides

Energy Technology Data Exchange (ETDEWEB)

Szteinsznaider, D; Naggiar, V; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

This measurements have been done while taking the value of the fission cross-sections of {sup 238}U as reference. The neutrons are produced by the reaction {sup 7}Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, cross-section constant between 150 and 2000 keV, for {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, cross-section constant between 700 and 1000 keV, for {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, for neutrons of 850 keV. (authors) [French] Ces mesures ont ete effectuees en prenant la valeur de la section efficace de fission de {sup 238}U comme reference. Les neutrons sont produits par la reaction {sup 7}Li(p,n) au generateur Van de Graaff de Saclay. Le domaine explore s'etend de quelques dizaines de kev a 2000 kev. Nous trouvons: pour {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, section efficace constante entre 150 et 2000 kev. pour {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, section efficace constante entre 700 et 1000 kev. pour {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, pour des neutrons de 850 kev. (auteurs)
Determination of the fission barrier height in fission of heavy radioactive beams induced by the (d,p)-transfer

CERN Multimedia

A theoretical framework is described, allowing to determine the fission barrier height using the observed cross sections of fission induced by the (d,p)-transfer with accuracy, which is not achievable in another type of low-energy fission of neutron-deficient nuclei, the $\\beta$-delayed fission. The primary goal is to directly determine the fission barrier height of proton-rich fissile nuclei, preferably using the radio-active beams of isotopes of odd elements, and thus confirm or exclude the low values of fission barrier heights, typically extracted using statistical calculations in the compound nucleus reactions at higher excitation energies. Calculated fission cross sections in transfer reactions of the radioactive beams show sufficient sensitivity to fission barrier height. In the probable case that fission rates will be high enough, mass asymmetry of fission fragments can be determined. Results will be relevant for nuclear astrophysics and for production of super-heavy nuclei. Transfer induced fission of...
Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

International Nuclear Information System (INIS)

Belo, Thiago F.; Fiel, Joao Claudio B.

2015-01-01

Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)
Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

Energy Technology Data Exchange (ETDEWEB)

Belo, Thiago F.; Fiel, Joao Claudio B., E-mail: thiagofbelo@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)

2015-07-01

Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)
Determination of the 54Fe(n, 2n)53gFe and 54Fe(n, 2n)53mFe cross sections averaged over a 235U fission neutron spectrum

International Nuclear Information System (INIS)

Ribeiro Guevara, S.; Arribere, M.; Kestelman, A.J.

2002-01-01

The reaction cross sections averaged over a 235 U fission neutron spectrum have been measured for the 54 Fe(n, 2n) 53g Fe and 54 Fe(n, 2n) 53m Fe threshold reactions. The values found are, respectively: (1.14 ± 0.13) μb, and (0.52 ± 0.16) μb. The measured cross sections are referred to the (111± 3) mb standard cross section of the 58 Ni(n, p) 58m+g Co reaction. The (81.7 ± 2.2) mb standard cross section value for the 54 Fe(n, p) 54 Mn reaction, was also used as a monitor to check the results obtained with the Ni standard, leading to an excellent agreement. (author)
Optical and statistical model calculation of the americium 242m capture cross section

International Nuclear Information System (INIS)

Tellier, Henry.

1981-04-01

The capture cross sections of Am 242m can be deduced from resonances analysis at low energy and computed with theoretical models at high energy. In this work, a coherent set of cross sections which reproduced the experimental values of the fission cross sections is computed. These calculations were performed for an energy of the incoming neutron between 1 keV and 1 MeV
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 the U-234(n,f) cross section with PPAC detectors at the nTOF facility

International Nuclear Information System (INIS)

Dobarro, C.P.

2005-06-01

The aim of this work was twofold: to measure the 234 U neutron-induced fission cross section in an extended energy range with an unprecedented resolution, and, in the process, to validate the experimental method we used at the new n-TOF-CERN facility. The experiment was designed in order to take advantage of the unique characteristics of the n-TOF facility: the long flight path offers a high energy resolution and the high-intensity, instantaneous neutron flux greatly reduces the background from the sample activities, making it possible to measure highly radioactive samples. The fission detection setup is based on an innovative technique that benefits from the use of very thin targets and detectors. Up to nine targets of high purity fission samples are sandwiched by Parallel Plate Avalanche Counters (PPAC). When a fission event happens, the two complementary fission fragments are detected by the PPACs adjacent to the fissioning target in a narrow time coincidence. Because several targets are simultaneously placed in-beam, relative measurements with respect to reference nuclei can be obtained. In this work, an original data-reduction method has been developed to deal with the particular characteristics of both the n-TOF data acquisition system, which is based on very accurate Flash-ADC digitizers, and the fission detection setup. The data reduction includes the coincidence windows and the signal amplitude requirements that we obtained from preliminary data analysis. The applied coincidence method is very powerful for dealing with the background rejection such as contamination by α activity, which is quite high for 234 U, and the signals produced by highly energetic reactions in the detectors. The data-reduction method also implements the fission event reconstruction using the position information obtained from the stripped cathodes and the delay line readout, which makes it possible to determine the fission fragment angular distributions, and the time-of-flight to
Polynomial parameterized representation of macroscopic cross section for PWR reactor

International Nuclear Information System (INIS)

Fiel, Joao Claudio B.

2015-01-01

The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 U 92 enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K inf , generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)
Remarks on the fission barriers of super-heavy nuclei

Energy Technology Data Exchange (ETDEWEB)

Hofmann, S. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Goethe-Universitaet Frankfurt, Institut fuer Physik, Frankfurt (Germany); Heinz, S.; Mann, R.; Maurer, J.; Muenzenberg, G.; Barth, W.; Dahl, L.; Kindler, B.; Kojouharov, I.; Lang, R.; Lommel, B.; Runke, J.; Scheidenberger, C.; Tinschert, K. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Antalic, S. [Comenius University, Department of Nuclear Physics and Biophysics, Bratislava (Slovakia); Eberhardt, K.; Thoerle-Pospiech, P.; Trautmann, N. [Johannes Gutenberg-Universitaet Mainz, Mainz (Germany); Grzywacz, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Tennessee, Knoxville, TN (United States); Hamilton, J.H. [Vanderbilt University, Department of Physics and Astronomy, Nashville, TN (United States); Henderson, R.A.; Kenneally, J.M.; Moody, K.J.; Shaughnessy, D.A.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Miernik, K. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Warsaw, Warsaw (Poland); Miller, D. [University of Tennessee, Knoxville, TN (United States); Morita, K. [RIKEN Nishina Center for Accelerator-Based Science, Wako, Saitama (Japan); Nishio, K. [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Popeko, A.G.; Yeremin, A.V. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Roberto, J.B.; Rykaczewski, K.P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Uusitalo, J. [University of Jyvaeskylae, Department of Physics, Jyvaeskylae (Finland)

2016-04-15

Shell-correction energies of super-heavy nuclei are approximated by using Q{sub α} values of measured decay chains. Five decay chains were analyzed, which start at the isotopes {sup 285}Fl, {sup 294}118, {sup 291}Lv, {sup 292}Lv and {sup 293}Lv. The data are compared with predictions of macroscopic-microscopic models. Fission barriers are estimated that can be used to eliminate uncertainties in partial fission half-lives and in calculations of evaporation-residue cross-sections. In that calculations, fission probability of the compound nucleus is a major factor contributing to the total cross-section. The data also provide constraints on the cross-sections of capture and quasi-fission in the entrance channel of the fusion reaction. Arguments are presented that fusion reactions for synthesis of isotopes of elements 118 and 120 may have higher cross-sections than assumed so far. (orig.)
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
Fission studies of gold induced by (1665 MeV) π- using a CR-39 detector

International Nuclear Information System (INIS)

Muhammad Ikram Shahzad; Yasin, Zafar; Sher, Gul

2012-01-01

The fission cross section and fission probability of 197 Au, induced by (1665 MeV) π'-, have been studied using CR-39 track detectors. A 4π-geometry was used to count track statistics. A beam of negative pions of 1665 MeV was produced at AGS of Brookhaven National Laboratory, USA, and allowed to fall normally on the stack. Two detectors from the stack were scanned for fission fragment tracks after etching in 6N NaOH at 70 ℃. The statistics of fission fragment tracks in both detectors were obtained. It was found that there was a marked asymmetry of registered tracks with respect to the forward and backward hemispheres. This asymmetry could be partly accounted for on the basis of momentum transfer to the struck nucleus. On the basis of counting statistics fission cross section was measured, and fission probability was determined by dividing the fission cross section with the reaction cross section. The fission cross-section and fission probability were compared with the computed values using the cascade-exciton model code CEM95. (authors)
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)
Plutonium-239 fission cross-section between 1 and 100 keV - International Evaluation Co-operation Volume 5

International Nuclear Information System (INIS)

Fort, E.; Salvatores, M.; Derrien, H.; Lagrange, Ch.; Kawai, M.; Nakajima, J.; Takano, H.; Weston, L.W.; Young, P.G.; Wagemans, C.

1994-01-01

A Working Party on International Evaluation Co-operation was established under the sponsorship of the OECD/NEA Nuclear Science Committee (NSC) to promote the exchange of information on nuclear data evaluations, validation, and related topics. Its aim is also to provide a framework for co-operative activities between members of the major nuclear data evaluation projects. This includes the possible exchange of scientists in order to encourage co-operation. Requirements for experimental data resulting from this activity are compiled. The Working Party determines common criteria for evaluated nuclear data files with a view to assessing and improving the quality and completeness of evaluated data. The Parties to the project are: ENDF (United States), JEFF/EFF (NEA Data Bank Member countries), and JENDL (Japan). Co-operation with evaluation projects of non-OECD countries are organised through the Nuclear Data Section of the International Atomic Energy Agency (IAEA). The following report was issued by a Subgroup investigating the fission cross-section of Plutonium-239 in the energy range 1 to 100 keV. This cross section is of particular importance for fast reactor applications, such as k eff , sodium void reactivity coefficient and control rod worth. An analysis of recent experimental data by L. Weston et al. give significantly lower cross-section values that the simultaneous evaluation performed by W. Poenitz for the ENDF/B-VI library. The objective of the subgroup was to resolve this discrepancy. One experimental program and one evaluation one have been agreed upon: The experimental program which essentially aims at normalisation checking has been performed in Geel and Oak Ridge. It supports an upward re-normalisation by ∼3.1%. The evaluation program has not been completed and even, as a consequence of the experimental results, loses a part of its justification. But some acquired results are important and can be used for future 239 Pu evaluations. The JEFF-2
Evaluation for ENDF/B-IV of the neutron cross sections for 235U from 82 eV to 25 keV

International Nuclear Information System (INIS)

Peelle, R.W.

1976-05-01

Capture and fission cross sections for 235 U in the ''unresolved resonance'' energy region were evaluated to permit determination of local-average resonance parameters for the ENDF/B-IV cross section file. Microscopic data were examined for infinitely dilute average fission and capture cross sections and also for intermediate structure unlikely to be reproduced by statistical fluctuations of resonance widths and spacings within known laws. Evaluated cross sections, averaged over lethargy intervals greater than 0.1, were obtained as an average over selected data sets after appropriate renormalization. Estimated uncertainties are given for these evaluated average cross sections. The ''intermediate'' structure fluctuations common to a few independent data sets were approximated by straight lines joining successive cross sections at 120 selected energy points; the cross sections at the vertices were adjusted to reproduce the evaluated average cross sections over the broad energy regions. Data sources and methods are reviewed, output values are tabulated, and some modified procedures are suggested for future evaluations. Evaluated fission and capture integrals for the resolved resonance region are also tabulated. These are not in agreement with integrals based on the resonance parameters of ENDF/B versions III and IV. 8 tables, 5 figures
Investigation on macroscopic cross section model for BWR pin-by-pin core analysis - 118

International Nuclear Information System (INIS)

Fujita, T.; Tada, K.; Yamamoto, A.; Yamane, Y.; Kosaka, S.; Hirano, G.

2010-01-01

A cross section model used in the pin-by-pin core analysis for BWR is investigated. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of state and history variables that have influences on the cross section and are tabulated prior to the core calculations. Variation of a cross section in a core simulator is classified into two different types, i.e., the instantaneous effect and the history effect. The instantaneous effect is incorporated by the variation of cross section which is caused by the instantaneous change of state variables. For this effect, the exposure, the void fraction, the fuel temperature, the moderator temperature and the control rod are used as indexes. The history effect is the cumulative effect of state variables. We treat this effect with a unified approach using the spectral history. To confirm accuracy of the cross section model, the pin-by-pin fission rate distribution and the k-infinity of fuel assembly which are obtained with the tabulated and the reference cross sections are compared. For the instantaneous effect, the present cross section model well reproduces the reference results for all off-nominal conditions. For the history effect, however, considerable differences both on the pin-by-pin fission rate distribution and the k-infinity are observed at high exposure points. (authors)

Optimization of multi-group cross sections for fast reactor analysis

International Nuclear Information System (INIS)

Chin, M. R.; Manalo, K. L.; Edgar, C. A.; Paul, J. N.; Molinar, M. P.; Redd, E. M.; Yi, C.; Sjoden, G. E.

2013-01-01

The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO 2 -UO 2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)
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)
Parameterized representation of macroscopic cross section for PWR reactor

International Nuclear Information System (INIS)

Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

2015-01-01

Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement
Review of the microscopic cross sections for the americium isotopes in the resolved resonance region

International Nuclear Information System (INIS)

Browne, J.C.

1978-01-01

The differential cross section measurements for 241 Am, /sup 242m/Am and 243 Am are reviewed in the energy range from 0.5 eV to 10 keV. Parameters extracted from resonance analysis, such as the neutron strength function, the average level spacing, the average capture and fission widths, are compared for the various measurements. The average capture and fission cross sections from 100 eV to 10 keV are directly compared. The status of the data set is discussed with suggestions for further measurements. 24 references
Estimation of (n,f) Cross-Sections by Measuring Reaction Probability Ratios

Energy Technology Data Exchange (ETDEWEB)

Plettner, C; Ai, H; Beausang, C W; Bernstein, L A; Ahle, L; Amro, H; Babilon, M; Burke, J T; Caggiano, J A; Casten, R F; Church, J A; Cooper, J R; Crider, B; Gurdal, G; Heinz, A; McCutchan, E A; Moody, K; Punyon, J A; Qian, J; Ressler, J J; Schiller, A; Williams, E; Younes, W

2005-04-21

Neutron-induced reaction cross-sections on unstable nuclei are inherently difficult to measure due to target activity and the low intensity of neutron beams. In an alternative approach, named the 'surrogate' technique, one measures the decay probability of the same compound nucleus produced using a stable beam on a stable target to estimate the neutron-induced reaction cross-section. As an extension of the surrogate method, in this paper they introduce a new technique of measuring the fission probabilities of two different compound nuclei as a ratio, which has the advantage of removing most of the systematic uncertainties. This method was benchmarked in this report by measuring the probability of deuteron-induced fission events in coincidence with protons, and forming the ratio P({sup 236}U(d,pf))/P({sup 238}U(d,pf)), which serves as a surrogate for the known cross-section ratio of {sup 236}U(n,f)/{sup 238}U(n,f). IN addition, the P({sup 238}U(d,d{prime}f))/P({sup 236}U(d,d{prime}f)) ratio as a surrogate for the {sup 237}U(n,f)/{sup 235}U(n,f) cross-section ratio was measured for the first time in an unprecedented range of excitation energies.
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
Measurement of the $^{242}$Pu(n,f) reaction cross-section at the CERN n_TOF facility

CERN Document Server

AUTHOR|(CDS)2080481; Kokkoris, Michael; Vlachoudis, Vasilis

The accurate knowledge of relevant nuclear data, such as the neutron-induced fission cross sections of various plutonium isotopes and other minor actinides, is crucial for the design of advanced nuclear systems as well as the development of comprehensive theoretical models of the fission process. The $^{242}$Pu(n,f) cross section was measured at the CERN n_TOF facility taking advantage of the wide energy range and the high instantaneous flux of the neutron beam. In this work, results for the $^{242}$Pu(n,f) measurement are presented along with a detailed description of the experimental setup, Monte-Carlo simulations and the analysis procedure, and a theoretical cross section calculation performed with the EMPIRE code.
Neutron-induced capture cross sections via the surrogate reaction method

International Nuclear Information System (INIS)

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

2011-01-01

The surrogate reaction method is an indirect way of determining cross sections for nuclear reactions that proceed through a compound nucleus. This technique enables neutron-induced cross sections to be extracted for nuclear reactions on short-lived unstable nuclei that otherwise can not be measured. This technique has been successfully applied to determine the neutron-induced fission cross sections of several short-lived nuclei. In this work, we investigate whether this powerful technique can also be used to determine of neutron-induced capture cross sections. For this purpose we use the surrogate reaction 174 Yb( 3 He, pγ) 176 Lu to infer the well known 175 Lu(n, γ) cross section and compare the results with the directly measured neutron-induced data. This surrogate experiment has been performed in March 2010. The experimental technique used and the first preliminary results will be presented. (authors)
Effective cross sections of U-235 and Au in a TRIGA-type reactor core

International Nuclear Information System (INIS)

Harasawa, S.; Auu, G.A.

1992-01-01

The dependence of effective cross sections of gold and uranium for neutron spectrum in Rikkyo University Reactor (TRIGA Mark- II, RUR) fuel cell was studied using computer calculations. The dependence of thermal neutron spectrum with temperature was also investigated. The effective cross section of gold in water of the fuel cell at 32degC was 90.3 barn and the fission cross section of U-235, 483 barn. These two values are similar to the cross sections for neutron energy of 0.034 eV. (author)
Use of an oscillation technique to measure effective cross-sections of fissionable samples in critical assemblies; Mesure des sections efficaces effectives d'echantillons fissiles par une methode d'oscillation dans les-assemblages critiques

Energy Technology Data Exchange (ETDEWEB)

Tretiakoff, O; Vidal, R; Carre, J C; Robin, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The authors describe the technique used to measure the effective absorption and neutron-yield cross-sections of a fissionable sample. These two values are determined by analysing the signals due to the variation in reactivity (over-all signal) and the local perturbation in the flux (local signal) produced by the oscillating sample. These signals are standardized by means of a set of samples containing quantities of fissionable material ({sup 235}U) and an absorber, boron, which are well known. The measurements are made for different neutron spectra characterized by lattice parameters which constitute the central zone within which the sample moves. This technique is used to study the effective cross-sections of uranium-plutonium alloys for different heavy-water and graphite lattices in the MINERVE and MARIUS critical assemblies. The same experiments are carried out on fuel samples of different irradiations in order to determine the evolution of effective cross-sections as a function of the spectrum and the irradiations. (authors) [French] On decrit la methode utilisee pour mesurer les sections efficaces effectives d'absorption et de production de neutrons d'un echantillon fissile. Ces deux grandeurs sont determinees en analysant les signaux dus a la variation de reactivite (signal global) et a la perturbation locale de flux (signal local) produits par l'echantillon oscillant. Ces signaux sont etalonnes a l'aide d'un jeu d'echantillons dont les teneurs en materiau fissile ({sup 235}U) et en absorbeur (bore) sont bien connues. Les mesures sont realisees pour differents spectres de neutrons caracterises par les parametres du reseau constituant la zone centrale a l'interieur de laquelle se deplace l'echantillon. A l'aide de cette methode on etudie les sections efficaces effectives d'alliage uranium-plutonium pour differents reseaux a eau lourde et a graphite dans les assemblages crtiques MINERVE et MARIUS. Les memes experiences sont effectuees sur des echantillons de
Impact of New Gadolinium Cross Sections on Reaction Rate Distributions in 10 * 10 BWR Assemblies

Energy Technology Data Exchange (ETDEWEB)

Perret, G.; Murphy, M.F.; Jatuff, F.; Chawla, R. [Paul Scherrer Inst, CH-5232 Villigen, (Switzerland); Sublet, J.Ch.; Bouland, O. [DEN, Commissariat Energie Atom, F-13108 St Paul Les Durance, (France); Chawla, R. [Ecole Polytech Fed Lausanne, CH-1015 Lausanne, (Switzerland)

2009-07-01

Radial distributions of the total fission rate and the {sup 238}U-capture-to-total-fission (C{sub 8}/F{sub tot}) ratio were measured in SVEA-96+ and SVEA-96 Optima2 assemblies during the LWR-PROTEUS program. Fission rates predicted using MCNPX with JEFF-3.1 cross sections underestimated the measured values in the gadolinium-poisoned pins of the SVEA-96 Optima2 assembly; similarly, C{sub 8}/F{sub tot} ratios were overestimated in some gadolinium-poisoned pins of the SVEA-96+ assembly. A considerable effort was invested at the Paul Scherrer Institut to explain the discrepancies in gadolinium pins, without success. Recently, gadolinium cross sections were measured at the Rensselaer Polytechnic Institute by Leinweber et al. and differed significantly from current library values. ENDF/B-VII.0 gadolinium cross sections have currently been modified to include the new measurements, and these data have been processed with NJOY to yield files usable by MCNPX. Fission rates in the gadolinium-poisoned fuel pins of the SVEA-96 Optima2 pins were increased by 1.4 to 2.0% using the newly produced cross sections, yielding to a better agreement with the experimental values. Predicted C{sub 8}/F{sub tot} ratios were decreased on average by 1.7% in both clustered and un-clustered groups of gadolinium-poisoned fuel pins of the SVEA-96+ assembly correcting the over predictions previously reported in the clustered gadolinium pins. Earlier reported discrepancies observed in PROTEUS integral experiments, between measured and calculated reaction rates in the gadolinium-poisoned pins, might thus be due to inaccurate gadolinium cross sections. The PROTEUS results support the new thermal and epithermal gadolinium data measured by Leinweber et al. (authors)
The fission cross sections of /sup 230/Th, /sup 232/Th, /sup 233/U, /sup 234/U, /sup 236/U, /sup 238/U, /sup 237/Np, /sup 239/Pu and /sup 242/Pu relative /sup 235/U at 14. 74 MeV neutron energy

Energy Technology Data Exchange (ETDEWEB)

Meadows, J.W.

1986-12-01

The measurement of the fission cross section ratios of nine isotopes relative to /sup 235/U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for /sup 235/U are: /sup 230/Th - 0.290 +- 1.9%; /sup 232/Th - 0.191 +- 1.9%; /sup 233/U - 1.132 +- 0.7%; /sup 234/U - 0.998 +- 1.0%; /sup 236/U - 0.791 +- 1.1%; /sup 238/U - 0.587 +- 1.1%; /sup 237/Np - 1.060 +- 1.4%; /sup 239/Pu - 1.152 +- 1.1%; /sup 242/Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs.
Fission in Empire-II version 2.19 beta1, Lodi

International Nuclear Information System (INIS)

Sin, M.

2003-01-01

This is a description of the fission model implemented presently in EMPIRE-II. This package offers two ways to calculate the fission probability selected by parameters in the optional input. Fission barriers, fission transmission coefficients, fission cross sections and fission files are calculated
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.)
Evaluated (n,p) cross sections of 46Ti, 47Ti and 48Ti

International Nuclear Information System (INIS)

Philis, C.; Bersillon, O.; Smith, D.; Smith, A.

1977-01-01

Microscopic evaluated neutron cross sections for the reactions 46 Ti(n;p) 46 Sc, 47 Ti(n;p) 47 Sc and 48 Ti(n;p) 48 Sc are obtained from threshold (or zero energy) to 20 MeV. The results are presented in graphical and numerical (ENDF format) form. The microscopic evaluated cross sections are compared with measured fission-spectrum-averaged values
Elliptical cross section fuel rod study II

International Nuclear Information System (INIS)

Taboada, H.; Marajofsky, A.

1996-01-01

In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab
Testing of ENDF/B cross section data in the Californium-252 neutron benchmark field

International Nuclear Information System (INIS)

Mannhart, W.

1979-01-01

The fission neutron field of 252 Cf presently represents one of the most well-known neutron benchmark fields. For 13 neutron reactions which are of importance in reactor metrology, measurements of spectrum-averaged cross sections, [sigma], performed in this neutron field were compared with calculated average cross sections. This comparison allows one to draw conclusions as to the quality of different sigma(E) data taken from ENDF/B-IV, from ENDF/B-V, and from recent experiments and used in the calculation of average cross sections. The comparison includes an uncertainty analysis regarding the different uncertainty contributions of [sigma], of sigma(E), and of the spectral distribution of 252 Cf fission neutrons. Additionally, in a few examples, sensitivity studies were carried out. The sensitivity of the spectrum-averaged cross sections to individual characteristics of the sigma(E) data, such as normalization factors or shifts in the energy scale, was investigated. Similarly, the sensitivity of [sigma] to the spectral distribution of 252 Cf was determined. 4 figures, 2 tables
(237)Np(n,f) Cross Section: New Data and Present Status

CERN Document Server

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

2011-01-01

In this document, we present the final result obtained at the n_TOF experiment; for the neutron-induced fission cross section of the (237)Np, from the fission threshold up to 1 GeV. The method applied to get tins result is briefly discussed. n_TOF data are compared to the last experimental measurements using other TOF facilities or the surrogate method, reported experiments performed with monoenergetic sources and the FISCAL systematic, including a discussion about the existing discrepancies.
Nuclear fission and fission-product spectroscopy: 3. International workshop on nuclear fission and fission-product spectroscopy

International Nuclear Information System (INIS)

Goutte, Heloise; Fioni, Gabriele; Faust, Herbert; Goutte, Dominique

2005-01-01

The present book contains the proceedings of the third workshop in a series of workshops previously held in Seyssins in 1994 and 1998. The meeting was jointly organized by different divisions of CEA and two major international laboratories. In the opening address, Prof. B. Bigot, the French High Commissioner for Atomic Energy, outlined France's energy policy for the next few decades. He emphasized the continuing progress of nuclear fission in both technical and economic terms, allowing it to contribute to the energy needs of the planet even more in the future than it does today. Such progress implies a very strong link between fundamental and applied research based on experimental and theoretical approaches. The workshop gathered the different nuclear communities studying the fission process, including topics as the following: - nuclear fission experiments, - spectroscopy of neutron rich nuclei, - fission data evaluation, - theoretical aspects of nuclear fission, - and innovative nuclear systems and new facilities. The scientific program was suggested by an International Advisory Committee. About 100 scientists from 13 different countries attended the conference in the friendly working atmosphere of the Castle of Cadarache in the heart of the Provence. The proceedings of the workshop were divided into 11 sections addressing the following subject matters: 1. Cross sections and resonances (5 papers); 2. Fission at higher energies - I (5 papers); 3. Fission: mass and charge yields (4 papers); 4. Light particles and cluster emission (4 papers); 5. Spectroscopy of neutron rich nuclei (5 papers); 6. Resonances, barriers, and fission times (5 papers); 7. Fragment excitation and neutron emission (4 papers); 8. Mass and energy distributions (4 papers); 9. Needs for nuclear data and new facilities - I (4 papers); 10. Angular momenta and fission at higher Energies - II (3 papers); 11. New facilities - II (2 papers). A poster session of 8 presentations completed the workshop
Least squares analysis of fission neutron standard fields

International Nuclear Information System (INIS)

Griffin, P.J.; Williams, J.G.

1997-01-01

A least squares analysis of fission neutron standard fields has been performed using the latest dosimetry cross sections. Discrepant nuclear data are identified and adjusted spectra for 252 Cf spontaneous fission and 235 U thermal fission fields are presented

Bonderenko self-shielded cross sections and multiband parameters derived from the LLL Evaluated-Nuclear-Data Library (ENDL)

International Nuclear Information System (INIS)

Cullen, D.E.

1978-01-01

Bonderenko self-shielded cross sections and multiband parameters from the Lawrence Livermore Laboratory Evaluated-Nuclear-Data Library (ENDL) as of July 4, 1978 are presented. These data include total, elastic, capture, and fission cross sections in the TART 175 group structure. Multiband parameters are listed. Bonderenko self-shielded cross section and the multiband parameters are presented on microfiche
Theory of nuclear fission: a review

International Nuclear Information System (INIS)

Mosel, U.

1976-01-01

General properties of nuclear fission are reviewed and related to our present knowledge of fission theory. For this purpose the basic reasons for the shape of the fission barriers are discussed and their consequences compared with experimental results on barrier shapes and structures. Special emphasis is put on the asymmetry of the fission barriers and mass-distributions and its relation to the shells of the nascent fragment shells. Finally the problem of calculating fission cross sections is discussed
Status of (n,2n) cross section measurements at Bruyeres-le-Chatel

International Nuclear Information System (INIS)

Frehaut, J.; Bertin, A.; Bois, R.; Jary, J.

1980-05-01

Cross sections for the (n,2n) reactions have been measured between threshold and 15 MeV for about 50 elements and separated isotopes using the large gadolinium-loaded liquid scintillator method and the 7 MV tandem Van de Graaff accelerator as a pulsed neutron source. The (n,2n) cross sections have been normalized to the fission cross section of 238 U; they are obtained with a relative accuracy of 4% to 10%. The systematic trends of the data obtained on series of separated isotopes are discussed, and some comparaisons with statistical model calculations are presented
Fission excitation function for 19F + 194,196,198Pt at near and above barrier energies

Directory of Open Access Journals (Sweden)

Singh Varinderjit

2015-01-01

Full Text Available Fission excitation functions for 19F + 194,196,198Pt reactions populating 213,215,217Fr compound nuclei are reported. Out of these three compound nuclei, 213Fr is a shell closed (N=126 compound nucleus and the other two are away from the shell closure. From a comparison of the experimental fission cross-sections with the statistical model predictions, it is observed that the fission cross-sections are underestimated by the statistical model predictions using shell corrected finite range rotating liquid drop model (FRLDM fission barriers. Further the FRLDM fission barriers are reduced to fit the fission cross-sections over the entire measured energy range.
Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, G.

1975-01-01

This is the first issue of a report series on Fission Product Nuclear Data (FPND), published every six months by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). Its purpose is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat. The present issue includes contributions which were received by NDS before 1 November 1975
Analysis of the 239Pu neutron cross sections from 300 to 2000 eV

International Nuclear Information System (INIS)

Derrien, H.; de Saussure, G.

1990-01-01

A recent high-resolution measurement of the neutron fission cross section of 239 Pu has allowed the extension from 1 to 2 keV of a previously reported resonance analysis of the neutron cross sections, and an improvement of the previous analysis in the range 0.3 to 1 keV. This report analyzes this region. 8 refs., 1 fig., 2 tabs
Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1981-06-01

This is the seventh issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The sixth issue of this series has been published in June 1980 as INDC(NDS)-113/G+P. The present issue includes contributions which were received by NDS between 1 August 1980 and 25 May 1981
Maxwellian-averaged cross sections calculated from JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

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

2000-02-01

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

International Nuclear Information System (INIS)

Gardner, M.A.

1981-09-01

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

Energy Technology Data Exchange (ETDEWEB)

Gardner, M.A.

1981-09-01

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

Directory of Open Access Journals (Sweden)

Morel P.

2011-10-01

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

Directory of Open Access Journals (Sweden)

Tassan-Got L.

2012-02-01

Full Text Available The measurement of neutron-induced cross sections of short-lived nuclei is extremely difficult due to the radioactivity of the samples. The surrogate reaction method is an indirect way of determining cross sections for nuclear reactions that proceed through a compound nucleus. This method presents the advantage that the target material can be stable or less radioactive than the material required for a neutron-induced measurement. We have successfully used the surrogate reaction method to extract neutron-induced fission cross sections of various short-lived actinides. In this work, we investigate whether this technique can be used to determine neutron-induced capture cross sections in the rare-earth region.
R-matrix analysis of 235U neutron transmission and cross sections in the energy range 0 to 2.25 keV

International Nuclear Information System (INIS)

Leal, L.C.; Derrien, H.; Larson, N.M.; Wright, R.Q.

1997-11-01

This document describes a new R-matrix analysis of 235 U cross section data in the energy range from 0 to 2,250 eV. The analysis was performed with the computer code SAMMY, that has recently been updated to permit, for the first time, inclusion of both differential and integral data within the analysis process. Fourteen differential data sets and six integral quantities were used in this evaluation: two measurements of fission plus capture, one of fission plus absorption, six of fission alone, two of transmission, and one of eta, plus standard values of thermal cross sections for fission, capture, and scattering, and of K1 and the Westcott g-factors for both fission and absorption. An excellent representation was obtained for the high-resolution transmission, fission, and capture cross-section data as well as for the integral quantities. The result is a single set of resonance parameters spanning the entire range up to 2,250 eV, a decided improvement over the present ENDF/VI evaluation, in which eleven discrete resonance parameter sets are required to cover that same energy range. This new evaluation is expected to greatly improve predictability of the criticality safety margins for nuclear systems in which 235 U is present
A Pebble Bed Reactor cross section methodology

International Nuclear Information System (INIS)

Hudson, Nathanael H.; Ougouag, Abderrafi M.; Rahnema, Farzad; Gougar, Hans

2009-01-01

A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.
Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, G.

1976-05-01

The purpose of this series is to inform scientists working on Fission Product Nuclear Data, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. This report consists of reproductions of essentially unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat
Calculation and Evaluation of Fission Yields and Capture Cross Sections Leading to the Production of Therapeutic Radionuclide by Means of Nuclear Reactors

International Nuclear Information System (INIS)

Sublet, J.C.

2009-01-01

Much progress has been made in nuclear medicine that involves the use of radionuclides for both diagnosis and therapy. Because of this qualitative and quantitative growth, the adoption of a set of established radionuclides for various applications, the methods of nuclide production need to be addressed and consideration given to other, emerging radionuclides that are judged to be developing in importance. The methods involved are characterized by the transmutation of isotopes by neutron-induced reactions and decays. Therefore, newly evaluated cross sections, fission yields and decay characteristics of relevance to the reactor production of those therapeutic radionuclides have been reviewed. Considerations of the decay schemes of all the nuclides involved are also included. (author)
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)
Ternary Fission of U235 by Resonance Neutrons

International Nuclear Information System (INIS)

Kvitek, I.; Popov, Ju.P.; Rjabov, Ju.V.

1965-01-01

Recently a number of papers have appeared indicating considerable variations in the ratio of the ternary-fission cross-section to the binary-fission cross-section of U 235 on transition from one neutron resonance to another. However, such variations have not been discovered in U 233 and Pu 239 . The paper reports investigations of the ternary fission of U 235 by neutrons with an energy of 0.1 to 30 eV. Unlike other investigators of the ternary fission of U 235 , we identified the ternary-fission event by the coincidence of one of the fission fragments with a light long-range particle. This made it passible to separate ternary fissions from the possible contribution of the (n, α)reaction. The measurements were performed at the fast pulsed reactor of the Joint Institute for Nuclear Research by the time-of-flight method. A flight length of 100 m was used, giving a resolution of 0.6 μs/m. Gas scintillation counters filled with xenon at a pressure of 2 atm were used to record the fission fragments and the light long-range particle. A layer of enriched U 235 ∼2 mg/cm 2 thick and ∼300 cm 2 in area was applied to an aluminium foil 20-fim thick. The scintillations from the fission fragments were recorded in the gas volume on one side of the foil and those from the light long-range particles in that on the other. In order to assess the background (e.g . coincidences of the pulse from a fragment with that from a fission gamma quantum or a proton from the (n, p) reaction in the aluminium foil), a measurement was carried out in which the volume recording the long-range particle was shielded with a supplementary aluminium filter 1-mm thick. The results obtained indicate the absence of the considerable variations in the ratio between the ternary-and binary- fission cross-sections for U 235 that have been noted by other authors. Measurements showed no irregularity in the ratio of the cross-sections in the energy range 0.1 to 0.2 eV. The paper discusses the possible effect of
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.
Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1983-08-01

This is the ninth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The eighth issue of this series has been published in July 1982 as INDC(NDS)-130. The present issue includes contributions which were received by NDS between 1 August 1982 and 25 June 1983

Progress in fission product nuclear data. No. 14

International Nuclear Information System (INIS)

Lammer, M.

1994-06-01

This is the 14th issue of a report series on Fission Product Nuclear Data published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of fission product yields, neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data from neutron induced and spontaneous fission, lumped fission product data. The first part of the report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. The third part contains requirements for further measurements
Progress in fission product nuclear data. No. 13

International Nuclear Information System (INIS)

Lammer, M.

1990-11-01

This is the 13th issue of a report series published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission), neutron reaction cross-sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products and bumped fission product data (decay heat, absorption, etc.). The first part of the report consists of unaltered original data which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. Part 3 contains requirements for further measurements
Neutron displacement damage cross sections for SiC

International Nuclear Information System (INIS)

Huang Hanchen; Ghoniem, N.

1993-01-01

Calculations of neutron displacement damage cross sections for SiC are presented. We use Biersack and Haggmark's empirical formula in constructing the electronic stopping power, which combines Lindhard's model at low PKA energies and Bethe-Bloch's model at high PKA energies. The electronic stopping power for polyatomic materials is computed on the basis of Bragg's Additivity Rule. A continuous form of the inverse power law potential is used for nuclear scattering. Coupled integro-differential equations for the number of displaced atoms j, caused by PKA i, are then derived. The procedure outlined above gives partial displacement cross sections, displacement cross sections for each specie of the lattice, and for each PKA type. The corresponding damage rates for several fusion and fission neutron spectra are calculated. The stoichiometry of the irradiated material is investigated by finding the ratio of displacements among various atomic species. The role of each specie in displacing atoms is also investigated by calculating the fraction of displacements caused by each PKA type. The study shows that neutron displacement damage rates of SiC in typical magnetic fusion reactor first walls will be ∝10-15 dpa MW -1 m 2 ; in typical lead-protected inertial confinement fusion reactor first walls they will be ∝15-20 dpa MW -1 m 2 . For fission spectra, we find that the neutron displacement damage rate of SiC is ∝74 dpa per 10 27 n/m 2 in FFTF, ∝39 dpa per 10 27 n/m 2 in HFIR, and 25 dpa per 10 27 n/m 2 in NRU. Approximately 80% of displacement atoms are shown to be of the carbon-type. (orig.)
Basic physics of the fission process. Chapter 2

International Nuclear Information System (INIS)

Michaudon, A.

1981-01-01

A general description of the fission process is given with special emphasis on those aspects which are necessary for the understanding of the measurements and calculations of neutron-induced fission cross-sections. Having considered the various phases of the process, some typical properties of the low-energy fission of actinide nuclei are presented and the more specific features of neutron induced fission are examined. (U.K.)
Measurement of 237Np fission rate ratio relative to 235U fission rate in cores with various thermal neutron spectrum at the Kyoto University Critical Assembly

International Nuclear Information System (INIS)

Unesaki, Hironobu; Shiroya, Seiji; Iwasaki, Tomohiko; Fujiwara, Daisuke; Kitada, Takanori; Kuroda, Mitsuo; Kohashi, Akio; Kato, Takeshi; Ikeuchi, Yoshitaka

2000-01-01

Integral measurements of 237 Np fission rate ratio relative to 235 U fission rate have been performed at Kyoto University Citrical Assembly. The fission rates have been measured using the back-to back type double fission chamber at five thermal cores with different H/ 235 U ratio so that the neutron spectra of the cores were systematically varied. The measured fission rate ratio per atom was 0.00439 to 0.0298, with a typical uncertainty of 2 to 3%. The measured data were compared with the calculated results using SRAC/TWOTRAN and MVP based on JENDL-3.2, which gave the averaged C/E values of 0.93 and 0.95, respectively. Obtained results of C/E using 237 Np cross sections from JENDL-3/2, ENDF/B-VI.5 and JEF2.2 show that the latter two gave smaller results than JENDL-3.2 by about 4%, which clearly reflects the discrepancy in the evaluated cross section among the libraries. This difference arises from both fast fission and resonance region. Although further improvement is recommended, 237 Np fission cross section in JENDL-3.2 is considered to be superior to those in the other libraries and can be adopted for use in design calculations for minor actinide transmutation system using thermal reactors with prediction precision of 237 Np fission rate with in 10%. (author)
A study of the differential cross section in subbarrier photofission of 238U

International Nuclear Information System (INIS)

Lindgren, L.J.; Sandell, A.

1977-03-01

A measurement of the angular distribution and yield of fission fragments from photofission of 238 U has been performed between 5.2 MeV and 6.4 MeV. As γ-source the bremsstrahlung from a microtron has been used. For the detection of the fission fragments solid state track detectors were used. The yield data were evaluated to approximate cross sections. The data were analyzed within the framework of the double hump barrier model. (Auth.)
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)
Assessment of the ''thermal normalization technique'' for measurement of neutron cross sections vs energy

International Nuclear Information System (INIS)

Peelle, R.W.; de Sassure, G.

1977-01-01

Refined knowledge of the thermal neutron cross sections of the fissile nuclides and of the (n,α) reaction standards, together with the reasonably well known energy dependence of the latter, have permitted resonance-region and low-keV fissile nuclide cross sections to be based on these standards together with count-rate ratios observed as a function of energy using a pulsed ''white'' source. As one evaluates cross sections for energies above 20 keV, optimum results require combination of cross section shape measurements with all available absolute measurements. The assumptions of the ''thermal normalization method'' are reviewed, and an opinion is given of the status of some of the standards required for its use. The complications which may limit the accuracy of results using the method are listed and examples are given. For the 235 U(n,f) cross section, the option is discussed of defining resonance-region fission integrals as standards. The area of the approximately 9 eV resonances in this nuclide may be known to one percent accuracy, but at present the fission integral from 0.1 to 1.0 keV is known to no better than about two percent. This uncertainty is based on the scatter among independent results, and has not been reduced by the most recent measurements. This uncertainty now limits the accuracy attainable for the 235 U(n,f) cross section below about 50 keV. Suggestions are given to indicate how future detailed work might overcome past sources of error
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 integral cross section ratios in two dosimetry benchmark neutron fields

Energy Technology Data Exchange (ETDEWEB)

Fabry, A [CEN-SCK, Mol (Belgium); Czock, K H [International Atomic Energy Agency, Laboratory Seibersdorf, Vienna (Austria)

1974-12-01

In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the {sup 103}Rh(n,n'){sup 103m}Rh and {sup 58}Ni(n,p){sup 58}Co integral cross sections have been accurately measured relatively to the {sup 115}In(n,n'){sup 115m} In cross section in the {sup 235}U thermal fission neutron spectrum and in the MOL-{sigma}{sigma} intermediate-energy standard neutron field. In this last neutron field, the data are related also to the {sup 235}U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific {sup 103}Rh(n,n'){sup 103m}Rh differential-energy cross section among the existing, conflicting data. (author)
Improvements on burnup chain model and group cross section library in the SRAC system

International Nuclear Information System (INIS)

Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.

1992-01-01

Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)
Methodology and application of the WIMS-D4M fission product data

International Nuclear Information System (INIS)

Mo, S.C.

1995-01-01

The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the 135 Xe and 149 Sm chains, and a lumped fission product to account for the absorption by fission products not explicitly treated. Burn-up calculations were performed for the ANS MEU core using WIMS and EPRI-CELL cross sections. The calculated eigenvalues and material loadings are in good agreements
Analytical evaluation of fission product sensitivities

International Nuclear Information System (INIS)

Sola, A.

1977-01-01

Evaluating the concentration of a fission product produced in a reactor requires the knowledge of a fairly large number of variables. Sensitivity studies were made to ascertain the important variables. Analytical formulae were developed sufficiently simple to allow numerical computations. Some simplified formulas are also given and they are applied to the following isotopes: 80 Se, 82 Se, 81 Br, 82 Br, 82 Kr, 83 Kr, 84 Kr, 85 Kr, 86 Kr. Their sensitivities to capture cross sections, fission yields, ratios of activation cross sections, half-lives (during and after irradiation), branching ratios, as well as to the neutron flux and to the time are considered
What happens to the fission process above the 2nd- and 3rd-chance fission thresholds

International Nuclear Information System (INIS)

Stewart, L.; Howerton, R.J.

1976-01-01

Although the multiple fission process is important at high neutron energies, most of the evaluations available today do not include these individual fission cross sections or their associated fission spectra. The representations used in the Los Alamos and Livermore libraries are described and calculations compared with 14-MeV integral experiments available on 235 U, 238 U, and 239 Pu. Further work is needed to clearly delineate the specific problems in order to propose unique solutions
Actinide cross section data and inertial confinement fusion for long term waste disposal

International Nuclear Information System (INIS)

Meldner, H.

1979-01-01

Actinide cross section data at thermonuclear neutron energies are needed for the calculation of ICF pellet center burnup of fission reactor waste, viz. 14 MeV neutron fission of the very long-lived actinides that pose storage problems. A major advantage of pellet center burnup is safety: only milligrams of highly toxic and active material need to be present in the fusion chamber, whereas blanket burnup requires the continued presence of tons of actinides in a small volume. The actinide data tables required for Monte Carlo calculations of the burnup of 241 Am and 243 Am are discussed in connection with typical burnup reactor fusion and fission spectra. 2 figures
Analysis of the 235U neutron cross sections in the resolved resonance range

International Nuclear Information System (INIS)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1989-01-01

Using recent high-resolution measurements of the neutron transmission of 235 U and the spin-separated fission cross-section data of Moore et al., a multilevel analysis of the 235 U neutron cross sections was performed up to 300 eV. The Dyson Metha Δ 3 statistics were used to help locate small levels above 100 eV where resonances are not clearly resolved even in the best resolution measurements available. The statistical properties of the resonance parameters are discussed
Application of dynamic pseudo fission products and actinides for accurate burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

1996-09-01

The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)
Theoretical analysis of knock-out release of fission products from nuclear fuels

International Nuclear Information System (INIS)

Yamagishi, S.

1975-01-01

The knock-out release of fission products is studied theoretically. The general equations of knock-out release are derived, assuming that a fission fragment passing through the surface of nuclear fuels knocks out a local region of the surface with an effective thickness and an effective cross-sectional area. Using these equations, the knock-out release of fission gases is calculated for various cases. The conditions under which the knock-out coefficients (the average number of uranium atoms knocked out by one fission fragment) is obtainable are clarified by experiments on the knock-out release of fission gases. A method of determining the effective thickness and the effective cross-sectional area of a knock-out region is proposed. (Auth.)
Evaluation of thermal neutron cross-sections and resonance integrals of protactinium, americium, curium, and berkelium isotopes

International Nuclear Information System (INIS)

Belanova, T.S.

1994-12-01

Data on the thermal neutron fission and capture cross-sections as well as their corresponding resonance integrals are reviewed and analysed. The data are classified according to the form of neutron spectra under investigation. The weighted mean values of the cross-sections and resonance integrals for every type of neutron spectra were adopted as evaluated data. (author). 87 refs, 2 tabs
Contribution to the study of nuclear fission

International Nuclear Information System (INIS)

Serot, O.

2009-09-01

The author proposes an overview of his research activity during the past fifteen years and more particularly that dealing with nuclear fission. The first part reports works on nucleus physics at the scission via the investigation of ternary fission (experimental procedure, influence of fission modes, influence of resonance spin, influence of excitation energy of the fissioning nucleus, emission probabilities, energy spectra of ternary alphas and tritons, emission mechanism). The second part reports measurements and assessments of neutron-induced fission cross sections. The third part reports the investigation of some properties of fission products (efficiencies, branching ratios of the main delayed neutron precursors)

Energy-differential cross section measurement for the 51V(n,α)48Sc reaction

International Nuclear Information System (INIS)

Kanno, I.; Meadows, J.W.; Smith, D.L.

1984-07-01

The activation method was used to measure cross sections for the 51 V(n,α) 48 Sc reaction in the threshold region, from 5.515 MeV up to 9.567 MeV. Twenty approximately-monoenergetic cross section values were obtained in this experiment. These data points span the energy region at roughly equal intervals. The experimental resolutions were in the range 0.153 to 0.233 MeV (FWHM). The present differential data cover approx. 50% of the total integral response of this reaction for the standard 235 U thermal-neutron-induced-fission neutron spectrum, and approx. 44% of the corresponding response for the standard 252 Cf spontaneous-fission neutron spectrum. Over the range 7.6 to 9.5 MeV the present experimental cross sections are noticeably larger (e.g., by approx. 50% at approx. 8.6 MeV) than the corresponding values from the ENDF/B-V evaluation. From approx. 6.7 to 7.5 MeV, the present values are somewhat below those of ENDF/B-V. At still lower energies the agreement is reasonably good considering the uncertainties introduced by energy scale definition very near the effective threshold where the cross section varies rapidly with neutron energy. Calculated integral cross sections based in part on the present work agree reasonably well within errors with reported integral results, provided that the reported data are renormalized to conform with recently-accepted values for appropriate standard reactions. 70 references
Fission rate distribution at the 84-pin radial section of a SVEA-96 Optima2 BWR assembly

Energy Technology Data Exchange (ETDEWEB)

Perret, Gregory; Murphy, Michael F.; Jatuff, Fabian [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

2008-07-01

Westinghouse boiling water reactor SVEA-96 Optima2 assemblies were studied during the LWRPROTEUS program at the PROTEUS facility in the Paul Scherrer Institute. Measured radial fission rate distributions at the 84-pin elevation are compared with MCNPX predictions using both ENDF/B-VI (Release 2) and JEFF-3.1 data libraries. Predicted fission rates agree within +-4.5% using both libraries. Fission rates were over-predicted in UO{sub 2} pins close to the missing 1/3 pins and under-predicted in UO{sub 2} pins close to the missing 2/3 pins. Recurrent under-estimations were observed in the UO{sub 2}-Gd{sub 2}O{sub 3} pins, for both libraries, which might be explained by over-estimated thermal cross-sections of {sup 157}Gd, as suggested in a recent work of G. Leinweber et al. (2006). (authors)
Effects of fissioning nuclei distributions on fragment mass distributions for high energy fission

Directory of Open Access Journals (Sweden)

Rossi P C R

2012-02-01

Full Text Available We study the effects of fissioning nuclei mass- and energy-distributions on the formation of fragments for fission induced by high energy probes. A Monte Carlo code called CRISP was used for obtaining mass distributions and spectra of the fissioning nuclei for reactions induced by 660 MeV protons on 241Am and on 239Np, by 500 MeV protons on 208Pb, and by Bremsstrahlung photons with end-point energies at 50 MeV and 3500 MeV on 238U. The results show that even at high excitation energies, asymmetric fission may still contribute significantly to the fission cross section of actinide nuclei, while it is the dominante mode in the case of lead. However, more precise data for high energy fission on actinide are necessary in order to allow definite conclusions.
Neutron capture cross section of $^{93}$Zr

CERN Document Server

We propose to measure the neutron capture cross section of the radioactive isotope $^{93}$Zr. This project aims at the substantial improvement of existing results for applications in nuclear astrophysics and emerging nuclear technologies. In particular, the superior quality of the data that can be obtained at n_TOF will allow on one side a better characterization of s-process nucleosynthesis and on the other side a more accurate material balance in systems for transmutation of nuclear waste, given that this radioactive isotope is widely present in fission products.
Impact of newly-measured gadolinium cross sections on BWR fuel rod reaction rate distributions

Energy Technology Data Exchange (ETDEWEB)

Jatuff, F.; Perret, G.; Murphy, M.; Grimm, P.; Seiler, R. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federal de Lausanne, CH-1015 Lausanne (Switzerland)

2008-07-01

Recent measurements of capture and total cross sections performed at the Rensselaer Polytechnic Institute in the USA confirmed many of the gadolinium thermal and resonant neutron cross section parameters within uncertainties, but they also showed up important discrepancies well out of uncertainties, such as an approx11% overestimation of the {sup 157}Gd thermal capture cross section in ENDF/B-VI and -VII with respect to the newly measured data. In this work, the impact of the newly measured gadolinium cross sections on BWR reactor physics parameters has been preliminarily evaluated. The comparisons of rod-by-rod fission rate and modified conversion ratio predictions with selected cold critical experiments at the PROTEUS reactor in Switzerland show the potential to resolve long-term unexplained discrepancies. (authors)
Analysis of the 235U neutron cross sections in the resolved resonance range

International Nuclear Information System (INIS)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1989-01-01

Using recent high-resolution measurements of the neutron transmission of 235 U and the spin-separated fission cross-section data of Moore et al., a multilevel analysis of the 235 U neutron cross sections was performed up to 300 eV. The Dyson Metha Δ 3 statistics were used to help locate small levels above 100 eV where resonances are not clearly resolved even in the best resolution measurements available. The statistical properties of the resonance parameters are discussed. 13 refs., 8 figs., 1 tab
Asymmetrically deformed states of thorium isotopes during fission process

International Nuclear Information System (INIS)

Blons, J.

1982-05-01

Some theoretical considerations are recalled on fission barriers calculated from macroscopic, microscopic or macroscopic-microscopic and ''thorium anomaly'' problem is set. Experimental techniques used to measure fission cross sections in (n,f) reactions near the threshold are described. Fission dectector is described; stray resonance problems and retrodiffused neutrons are discussed. Results obtained in experimental study of 230 Th(n,f) and 232 Th(n,f) reactions are presented. They are compared with results obtained in other laboratories. The analysis model which allows to describe a (n,f) reaction is exposed. The compound nucleus formation cross section and transmission coefficients in neutron and gamma output channel are presented according to neutron energy for each value of angular moment and parity. Cross-section analysis and angular distribution obtained respectively in 230 Th(n,f) and 232 Th(n,f) reactions is exposed. Result interpretation show new aspects of nuclei rotational spectra and new nuclear forms [fr
[Absolute fission cross sections in the 14 MeV energy region]. Progress report, July 1982-June 1983

International Nuclear Information System (INIS)

1983-01-01

Progress is reported on the following studies: thermal neutron absorption cross section of sulfur and the 252 Cf nu bar dilemma, the sigma (H)/sigma (Mn) cross section ratio, the sigma (H)/sigma (B) cross section ratio, 14 MeV neutron cross section measurements, beryllium-based pulsed neutron detector, and testing charged particle transport and Monte Carlo codes
Prompt fission neutron spectra and average prompt neutron multiplicities

International Nuclear Information System (INIS)

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

1983-01-01

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

International Nuclear Information System (INIS)

Lammer, G.

1978-07-01

This is the fourth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.)
Fission fragment angular distributions in proton-induced fission of 209Bi (p,f) and 197Au (p,f)

International Nuclear Information System (INIS)

Soheily, S.; Noshad, H.; Lamehi-Rashti, M.

2002-01-01

The fission fragment angular distributions have been measured for proton-induced fission of 209 B i and 197 A u nuclei using surface barrier detectors at several energies between 25 MeV and 30 MeV. The experimental anisotropies are found to be in agreement with the predictions of the Standard Saddle-Point Statistical Model. The fission cross sections of 209 B i and 197 A u nuclei were also measured and compared with the previous works
Some remarks on the neutron elastic- and enelastic-scattering cross sections of palladium

International Nuclear Information System (INIS)

Chiba, S.; Guenther, P.T.; Smith, A.B.

1989-05-01

The cross sections for the elastic-scattering of 5.9, 7.1 and 8.0 MeV neutrons from elemental palladium were measured at forty scattering angles distributed between ∼15/degree/ and 160/degree/. The inelastic-scattering cross sections for the excitation of palladium levels at energies of 260 keV to 560 keV were measured with high resolution at the same energies, and at a scattering angle of 80/degree/. The experimental results were combined with lower-energy values previously obtained by this group to provide a comprehensive data base extending from near the inelastic-scattering threshold to 8 MeV. That data base was interpreted in terms of a coupled-channel model, including the inelastic excitation of one- and two-phonon vibrational levels of the even isotopes of palladium. It was concluded that the palladium inelastic-scattering cross section, at the low energies of interest in assessment of fast-fission-reactor performance, are large (∼50% greater than given in widely used evaluated fission-product data files). They primarily involve compound-nucleus processes, with only a small direct-reaction component attributable to the excitation of the one-phonon, 2 + , vibrational levels of the even isotopes of palladium. 24 refs., 6 figs
Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1982-07-01

This is the eighth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. Therefore, the IAEA cannot be held responsible for the information contained nor for any consequences resulting from the use of this information. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The seventh issue of this series has been published in July 1981 as INDC(NDS)-116. The present issue includes contributions which were received by NDS between 1 August 1981 and 15 June 1982
Fission of {sup 209}Bi and {sup 197}Au nuclei induced by 30 MeV protons

Energy Technology Data Exchange (ETDEWEB)

Noshad, Houshyar; Soheyli, Saeed [Amir-Kabir University of Technology, Physics and Nuclear Science Department, Tehran (Iran); Lamehi-Rachti, Mohammad [Atomic Energy Organization of Iran (AEOI), Nuclear Research Center, Van de Graaff Laboratory, Tehran (Iran)

2001-10-01

Thin targets of {sup 209}Bi and {sup 197}Au were bombarded with 30 MeV protons at the Cyclotron Department of Nuclear Research Center for Agriculture and Medicine (NRCAM). Correlated measurements of kinetic energies of fission fragment pairs, and their time-of-flights were made using pair spectrometry. The fission cross sections, fragment mass distributions, and total kinetic energy distributions of the fragments were measured in our experiment. The accurate values of cross sections for fission of {sup 209}Bi and {sup 197}Au nuclei with 30 MeV protons were obtained to be 1,100{+-}100 and 62{+-}5.6 {mu}b, respectively. The cross section of {sup 209}Bi fission with its associated error, through using this method, has not been reported previously. The interpretation in terms of liquid-drop model of fissioning nucleus {sup 210}Po at the excitation energy of 35 MeV was confirmed by the dispersion of the distribution in fragment mass for bismuth fission. (author)
Effect of fission dynamics on the spectra and multiplicities of prompt fission neutrons

International Nuclear Information System (INIS)

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

1985-01-01

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

International Nuclear Information System (INIS)

Kopach, Yu.N.; Popov, A.B.; Furman, V.I.; Alfimenkov, V.P.; Lason', L.; Pikel'ner, L.B.; ); Gonin, N.N.; Kozlovskij, L.K.; Tambovtsev, D.I.; Gagarskij, A.M.; Petrov, G.A.; Sokolov, V.E.

2001-01-01

A new approach to the description of the fission, similar to the well-known reaction theory and based on the helicity representation for the exit fission channels, is briefly summarized. This approach allows one to connect the multimodal fission representation with A. Bohr's concept of the fission transition states and to obtain formulae for the partial and differential fission cross sections. The formulae are used for analysis of the angular anisotropy of fragments in the neutron resonance induced fission of aligned 235 U nuclei and of the P-even angular forward-backward and right-left correlations of fragments oe the P-odd correlations caused by the interference of s- and p-wave neutron resonances [ru
Determination of reaction cross sections with the aid of α decay in the 12C, 14C + 209Bl reactions

International Nuclear Information System (INIS)

Hick, H.

1980-01-01

For the reactions 14 C + 209 Bi and 12 C + 209 Bi excitation functions at energies in the range between 57 MeV and 76 MeV are measured. Radiative capture and particle evaporation cross sections were determined by means of α-spectroscopy, and fission cross sections were determined by the measurement of the γ-radiation after the β-decay of the fission products. For the radiative capture for the reaction 14 C + 209 Bi upper limits for the cross section from 21 nbarn to 178 nbarn in the energy interval 61-74 MeV were determined. The fission cross sections were 80 +- 30 mbarn at 490 +- 200 mbarn at 76 MeV. For the reaction 12 C + 209 Bi three new α-lines were found. They were due to the slope at their excitation functions assigned to the decay of isomeric states of following nuclei: 219 Ac Esub(α) = 9419 +- 4 keV Tsub(1/2) = 830 +- 100/μsec, 218 Ac Esub(α) = 9271 +- 4 keV Tsub(1/2) = 810 +- 70/μsec, 217 Ac Eα = 9730 +- 5 keV Tsub(1/2) = 970 +- 190/μsec. For the reactions respectively 12 C + 209 Bi calculations using the statistical model code Grogi of J. Gilat are performed. The calculated branchings of the evaporation channels were compared with the experiment. (orig./HSI) [de
Fission of heavy hypernuclei

International Nuclear Information System (INIS)

Nifenecker, H.

1993-01-01

The results on delayed and prompt fission of heavy hypernuclei obtained by the LEAR PS177 collaboration are recalled and discussed. It is shown that the hypernuclei life-times can be explained in term of a weak strangeness violating lambda-nucleon interaction with a cross section close to 6.0 10 -15 barns. The lambda attachment function is shown to be sensitive to the scission configuration, just before fission, and to the neck dynamics. This function provides a new way to study the nuclear scission process. (author)
SFINX: Soviet-French integral experiment on measuring the capture and fission at Masurca and BFS

International Nuclear Information System (INIS)

Doulin, V.A.; Mikhailov, J.M.; Mozhaev, V.K.

1990-01-01

The SFINX experiment was aimed at the comparison of experimental procedures used at the MASURCA and the BFS critical assemblies for measuring the ratio of the 238 U and 239 Pu average fission cross-sections to the 235 U average cross-section (F8/F5, F9/F5) and of the 238 U average capture cross-section to the 239 Pu average fission cross section (C8/F9). As part of the calibration of the measurements F8/F5 and F9/F5 were also measured in a thermal column. To obtain C8/F9, absolute measurements of capture rates in 238 U and fission rates in 239 Pu and 235 U were carried out. The measurements were made in September 1987 at the MASURCA facility (FRANCE) in the BALZAC 1 critical assembly and in the thermal column of the HARMONIE facility. In April 1989 these measurements were complemented by joint measurements of the 239 Pu absolute fission rate at the BFS 55-1 critical assembly (USSR)

Nuclear data processing for cross-sections generation for fusion-fission, ADS, and IV generation reactors utilization

International Nuclear Information System (INIS)

Velasquez, Carlos E.; Fernandes, Lorena C.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

2017-01-01

One of the mains topics about nuclear reactors is the microscopic cross section for incident neutrons. Therefore, in this work, it is evaluated the microscopic and macroscopic cross section for a nuclide and a material. One of the nuclides microscopic cross-section studied is the 56 Fe which is the highest compound from the material macroscopic cross section studied SS316. On the other hand, it was studied the microscopic cross section of the 242 Pu which is one of the nuclides that composes the nuclear fuel. The nuclear fuel chosen is a spent fuel reprocessed by UREX+ technique and spiked with thorium with 20% of fissile material. Therefore it was studied the macroscopic cross section from this nuclear fuel. Both of them were compared by using three different ways to reprocess the nuclides, one for LWR, another for ADS and the last one for Fusion reactors. The library used was JEFF-3.2 recommend for the reactors studied. The comparison was made at 1200 K for the nuclear fuel and 700K for the SS316.The results present differences due to the energy discretization, the number of groups chosen for each reactor and some nuclear reactions taken into consideration according to the neutron spectrum for each reactor. The nuclides were processed by NJOY99.364 and plotted with MCNP-Vised. (author)
Nuclear data processing for cross-sections generation for fusion-fission, ADS, and IV generation reactors utilization

Energy Technology Data Exchange (ETDEWEB)

Velasquez, Carlos E.; Fernandes, Lorena C.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

One of the mains topics about nuclear reactors is the microscopic cross section for incident neutrons. Therefore, in this work, it is evaluated the microscopic and macroscopic cross section for a nuclide and a material. One of the nuclides microscopic cross-section studied is the {sup 56}Fe which is the highest compound from the material macroscopic cross section studied SS316. On the other hand, it was studied the microscopic cross section of the {sup 242}Pu which is one of the nuclides that composes the nuclear fuel. The nuclear fuel chosen is a spent fuel reprocessed by UREX+ technique and spiked with thorium with 20% of fissile material. Therefore it was studied the macroscopic cross section from this nuclear fuel. Both of them were compared by using three different ways to reprocess the nuclides, one for LWR, another for ADS and the last one for Fusion reactors. The library used was JEFF-3.2 recommend for the reactors studied. The comparison was made at 1200 K for the nuclear fuel and 700K for the SS316.The results present differences due to the energy discretization, the number of groups chosen for each reactor and some nuclear reactions taken into consideration according to the neutron spectrum for each reactor. The nuclides were processed by NJOY99.364 and plotted with MCNP-Vised. (author)
Micromegas detector for $^{33}$S(n,$\\alpha$) cross section measurement at n_TOF

CERN Multimedia

The present proposal is a consequence of the successful tests performed in 2011 related to the Letter of Intent CERN-INTC-2010-023/I-092. The main goal of this proposal is a first (n,$\\alpha$) cross section measurement with the Micromegas detector presently running at n_TOF for monitoring purposes and fission cross section measurements. The $^{33}$S(n,$\\alpha$) cross section is of interest in astrophysics mainly due to the origin of $^{36}$S which is still an open question. $^{33}$S is also of interest in medical physics since it has been proposed as a possible/alternative cooperating target to boron neutron capture therapy. Important discrepancies between previous measurements of $^{33}$S(n,$\\alpha$) cross section and especially between the resonance parameters are found in the literature. We propose to measure the (n,$\\alpha$) cross section of the stable isotope $^{33}$S in the energy range up to 300 keV covering the astrophysical range of interest. The possibility of increasing this energy range will be st...
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
Fission yields and cross section uncertainty propagation in Boltzmann/Bateman coupled problems: Global and local parameters analysis with a focus on MTR

International Nuclear Information System (INIS)

Frosio, Thomas; Bonaccorsi, Thomas; Blaise, Patrick

2016-01-01

Highlights: • Nuclear data uncertainty propagation for neutronic quantities in coupled problems. • Uncertainties are detailed for local isotopic concentrations and local power maps. • Correlations are built between space areas of the core and for different burnups. - Abstract: In a previous paper, a method was investigated to calculate sensitivity coefficients in coupled Boltzmann/Bateman problem for nuclear data (ND) uncertainties propagation on the reactivity. Different methodologies were discussed and applied on an actual example of multigroup cross section uncertainty problem for a 2D Material Testing Reactor (MTR) benchmark. It was shown that differences between methods arose from correlations between input parameters, as far as the method enables to take them into account. Those methods, unlike Monte Carlo (MC) sampling for uncertainty propagation and quantification (UQ), allow obtaining sensitivity coefficients, as well as correlations values between nuclear data, during the depletion calculation for the parameters of interest. This work is here extended to local parameters such as power factors and isotopic concentrations. It also includes fission yield (FY) uncertainty propagation, on both reactivity and power factors. Furthermore, it introduces a new methodology enabling to decorrelate direct and transmutation terms for local quantities: a Monte-Carlo method using built samples from a multidimensional Gaussian law is used to extend the previous studies, and propagate fission yield uncertainties from the CEA’s COMAC covariance file. It is shown that, for power factors, the most impacting ND are the scattering reactions, principally coming from 27 Al and (bounded hydrogen in) H 2 O. The overall effect is a reduction of the propagated uncertainties throughout the cycle thanks to negatively correlated terms. For fission yield (FY), the results show that neither reactivity nor local power factors are strongly affected by uncertainties. However, they
Applications of the nuclear theory to the computation of neutron cross sections for actinide isotopes

International Nuclear Information System (INIS)

Konshin, V.A.

1981-01-01

Neutron cross section calculational methods for actinides in the unresolved resonance energy range (1-150 kev) are discussed, with a special emphasis on calculation of width fluctuation factors for the generalized distribution, as well as for a sub-threshold fission. It is shown that the energy dependence of sub(J), the (n,n') -process competition and the structure in neutron cross section has to be taken into account in the energy range considered. Analysis of different approaches in the statistical theory for heavy nuclei neutron cross-section calculation is given, and it is shown to be important to allow for the (n,γf)-reaction in neutron cross section calculations for fissile nuclei. The use of the non-spherical potential, the Lorentzian spectral factor and the Fermi-gas model involving the collective modes enables to obtain the self-consistent data for all neutron cross sections, including σnγ. (author)
Some aspects of the nuclear fission process

International Nuclear Information System (INIS)

Netter, F.

1961-01-01

In the following report one can find first a short general view on the present situation of our knowledge concerning the nuclear fission process, namely on the nucleus going through the saddle-point. Then there are some aspects connected with the excitation energy of the fissioning nucleus. The measurements made at Saclay on the fast neutron fission cross-section of U 233 , U 235 , Pu 239 , U 238 are described at the beginning of this work. It appears that for U 233 there is some characteristic shape modulation of the cross-section curve, in relation with the collective excited state of the deformed nucleus at the saddle-point. Good evidence of this is also given by the study of the relative fission rate with emission of long-range particles; it appears also that this ternary fission rate does not change substantially for neutron between thermal energy and 2 MeV, but that is very lower for the compound nucleus U 239 than for even-even compound nuclei. At the end there are some experiments on the strong 4,5 MeV gamma-ray originated by slow neutron absorption in U 235 . Time-of-flight device is used to establish that this 4,5 MeV gamma-ray seems mostly connected with radiative capture. (author) [fr
Fission neutron spectrum averaged cross sections of some threshold reactions on cadmium: production feasibility of no-carrier-added 103Pd in a nuclear reactor

International Nuclear Information System (INIS)

Abbasi, I.A.; Subhani, M.S.; Zaidi, J.H.; Arif, M.

2006-01-01

Systematic studies on fission neutron spectrum averaged cross sections of some threshold reactions like (n, p) and (n, α) on cadmium were carried out using the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy. Special attention was paid to the formation of 103 Pd via the 106 Cd(n,a α) 103 Pd reaction since it is an important therapeutic radionuclide. At a fast flux neutron density of 7.5 x 10 13 cm 2 s -1 and an irradiation time of 120 h, using 100% enriched 106 Cd target 340 MBq of no-carrier-added 103 Pd per batch could be produced. The method is thus suitable for medium-scale production of this radionuclide. (orig.)
Contributions of Gerard de Saussure to fission physics: Subthreshold and near-subthreshold measurements at the ORELA

International Nuclear Information System (INIS)

Perez, R.B.; Difilippo, F.C.

1992-01-01

In the late 1960s, Strutinsky's theoretical work on the structure of the fission barrier, together with some new neutron cross-section measurements performed at Saclay and Gel, was called to Gerard de Saussure's attention, and he immediately recognized the importance of those developments. Indeed, de Saussure was quick to note that for neutron energies in the range of the fission barrier potential energy, fission cross-section measurements would yield direct information about the physical properties of the fission barrier. Under his leadership, a set of extensive and precise measurements was performed at the Oak Ridge Electron Linear Accelerator (ORELA) by an international group of researchers, lasting over two decades. Those measurements were unique in many respects: Fission widths were determined for many previously unreported resonances in the subthreshold region, and a detailed study was performed on the physical properties of the fission barrier at high nuclear deformations. In this paper, we report a specific part of the de Saussure's extensive involvement in the area of subthreshold fission physics: the 1979 measurement of the 238 U(n,f) cross ection in the subthreshold and near-subthreshold regions. This cross section was measured in the energy region between 5 eV and 3.5 MeV using a large fission chamber loaded with high-purity 238 U, which had 235 U
What can we learn about heavy ion fusion by studying fission angular distributions

International Nuclear Information System (INIS)

Back, B.B.

1984-01-01

Determinations of complete fusion reactions leading to fissionable systems are associated with problems of separating fragments from quasi-fission reactions from those arising from fission of the completely fused system. Inferring complete fusion cross sections from the minute cross sections for the evaporation residue channel is hampered by the insufficient knowledge of the branching ratio for neutron emission and fission in the decay sequence of the completely fused system. From a quantitative analysis of the fragment angular distributions it is, however, possible under certain assumptions to deduce the relative contribution of complete fusion and quasi-fission. It is found that the complete fusion process is hindered for heavy projectiles. The excess radial energy over the interaction barrier needed to induce fusion with heavy projectiles is determined in several cases and systematic trends are presented
Electromagnetic fission of 238U at 600 and 1000 MeV per nucleon

International Nuclear Information System (INIS)

Rubehn, T.; Mueller, W.F.J.; Bassini, R.; Begemann-Blaich, M.; Blaich, T.; Gross, C.; Imme, G.; Iori, I.; Kunde, G.J.; Kunze, W.D.; Lindenstruth, V.; Lynen, U.; Moehlemkamp, T.; Moretto, L.G.; Ocker, B.; Pochodzalla, J.; Raciti, G.; Reito, S.; Sann, H.; Schuettauf, A.; Serfling, V.; Trautmann, W.; Trzcinski, A.; Verde, G.; Woerner, A.; Zude, E.; Zwieglinski, B.

1995-05-01

Electromagnetic fission of 238 U projectiles at E/A=600 and 1000 MeV was studied with the ALADIN spectrometer at the heavy-ion synchrotron SIS. Seven different targets (Be, C, Al, Cu, In, Au and U) were used. By considering only those fission events where the two charges added up to 92, most of the nuclear interactions were excluded. The nuclear contributions to the measured fission cross sections were determined by extrapolating from beryllium to the heavier targets with the concept of factorization. The obtained cross sections for electromagnetic fission are well reproduced by extended Weizsaecker-Williams calculations which include E1 and E2 excitations. The asymmetry of the fission fragments' charge distribution gives evidence for the excitation of the double giant-dipole resonance in uranium. (orig.)
Measurement of cross sections of threshold detectors with spectrum average technique

International Nuclear Information System (INIS)

Agus, Y.; Celenk, I.; Oezmen, A.

2004-01-01

Cross sections of the reactions 103 Rh(n, n') 103m Rh, 115 In(n, n') 115m In, 232 Th(n, f), 47 Ti(n, p) 47 Sc, 64 Zn(n, p) 64 Cu, 58 Ni(n, p) 58 Co, 54 Fe(n, p) 54 Mn, 46 Ti(n, p) 46 Sc, 27 Al(n, p) 27 Mg, 56 Fe(n, p) 56 Mn, 24 Mg(n, p) 24 Na, 59 Co(n, α) 56 Mn, 27 Al(n, α) 24 Na and 48 Ti(n, p) 48 Sc were measured with average neutron energies above effective threshold by using the activation method through usage of spectrum average technique in an irradiation system where there are three equivalent Am/Be sources, each of which has 592 GBq activity. The cross sections were determined with reference to the fast neutron fission cross section of 238 U. The measured values and published values are generally in agreement. (orig.)
Proceedings of the specialists' meeting on physics and engineering of fission and spallation, 1989

International Nuclear Information System (INIS)

Nakagome, Yoshihiro

1990-07-01

The third meeting was held on August 1, and the fourth meeting was held on December 12, 1989. The reports of the international conferences on 50 years research on nuclear fission in Germany and USA, and the reports on the nuclear data of fission-produced nuclei for evaluating reactor decay heat, the atomic mass formula considering proton-neutron interaction and unstable nuclei, research on short life fission fragments by on-line isotope separation process, the reactor physics on waste annihilation disposal and fuel breeding with an accelerator, the double differential cross section of back neutrons in nuclear spallation reaction, measurement of fission cross section and fission neutron spectra with fast neutrons, U-235 fission spectra by unfolding activation foil data and production mechanisms of intermediate mass fragments from hot nuclei-emission of complex and fission fragments for 84 Kr+ 27 Al at 10.6 MeV/u were made. (K.I.)
(n,α reactions cross section research at IPPE

Directory of Open Access Journals (Sweden)

Giorginis G.

2012-02-01

Full Text Available An experimental set-up based on an ionization chamber with a Frisch grid and wave form digitizer was used for (n,α cross section measurements. Use of digital signal processing allowed us to select a gaseous cell inside the sensitive area of the ionization chamber and determine the target atoms in it with high accuracy. This kind of approach provided us with a powerful method to suppress background arising from the detector structure and parasitic reactions on the working gas components. This method is especially interesting to study neutron reactions with elements for which solid target preparation is difficult (noble gases for example. In the present experiments we used a set of working gases which contained admixtures of nitrogen, oxygen, neon, argon and boron. Fission of 238U was used as neutron flux monitor. The cross section of the (n,α reaction for 16O, 14N, 20Ne, 36Ar, 40Ar and the yield ratio α0/α1 of 10B(n,α0 to 10B(n,α1 reactions was measured for neutron energies between 1.5 and 7 MeV. Additionally a measurement of the 50Cr(n,α cross section using a solid chromium target is also reported.
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
Measurement of the ^235mU Production Cross Section Using a Critical Assembly*

Science.gov (United States)

Macri, Robert; Authier, Nicolas; Becker, John; Belier, Gilbert; Bond, Evelyn; Bredeweg, Todd; Glover, S.; Meot, Vincent; Rundberg, Robert; Vieira, David; Wilhelmy, Jerry

2006-10-01

Measurements of the creation and destruction cross sections for actinide nuclei constitute an important experimental effort in support of Stockpile Stewardship. In this talk I will give a progress report on the effort to measure the production cross section of the ^235mU isomer integrated over a fission neutron spectrum. This ongoing experiment is fielded at CEA in Valduc, France, taking advantage of the CALIBAN critical assembly. This effort is performed in collaboration with LANL, LLNL, Bruyeres le Chatel, and Valduc staff. This experiment utilizes a technique to measure internal conversion electrons from the ^235mU isomer with the French BIII detector (Bruyeres le Chatel), and involves a substantial chemistry effort (LANL) to prepare targets for irradiation and counting, as well as to remove fission fragments after irradiation. Experimental techniques will be discussed and preliminary data presented. *Work performed under the auspices of the U.S. Department of Energy by Los Alamos National Laboratory (W-7405-ENG-36) and Lawrence Livermore National Laboratory (W-7405-ENG-48), and CEA-DAM under CEA-DAM NNSA-DOE agreement.
Reich-Moore and Adler-Adler representations of the 235U cross sections in the resolved resonance region

International Nuclear Information System (INIS)

Saussure, G. de; Leal, L.C.; Perez, R.B.

1990-01-01

In the first part of this paper, a reevaluation of the low-energy neutron cross sections of 235 U is described. This reevaluation was motivated by the discrepancy between the measured and computed temperature coefficients of reactivity and is based on recent measurements of the fission cross section and of η in the thermal and subthermal neutron energy regions. In the second part of the paper, we discuss the conversion of the Reich-Moore resonance parameters, describing the neutron cross sections of 235 U in the resolved resonance region, into equivalent Adler-Adler resonance parameters and into equivalent momentum space multipole resonance parameters
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.)
AFCI-2.0 Neutron Cross Section Covariance Library

Energy Technology Data Exchange (ETDEWEB)

Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-03-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural
AFCI-2.0 Neutron Cross Section Covariance Library

International Nuclear Information System (INIS)

Herman, M.; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-01-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R and D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78

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

International Nuclear Information System (INIS)

Mann, F.M.

1979-07-01

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

International Nuclear Information System (INIS)

Anderl, R.A.; Harker, Y.D.; Millsap, D.A.; Rogers, J.W.; Ryskamp, J.M.

1982-01-01

The Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory (INEL) is a Cross Section Evaluation Working Group (CSEWG) benchmark for data testing of dosimetry, fission-product and actinide cross sections important to fast-reactor technology. In this paper we present the results of our work in updating the CFRMF spectrum characterization and in applying CFRMF integral data to testing ENDF/B-V dosimeter cross sections. Updated characterization of the central neutron spectrum includes the results of neutronics calculations with ENDF/B-V nuclear data, the generation of a fine-group spectrum representation for integral data-testing applications, and a sensitivity and uncertainty analysis which provides a flux-spectrum covariance matrix related to uncertainties and correlations in the nuclear data used in a neutronics calculation. Our application of CFRMF integral data to cross section testing has included both conventional integral testing analyses and least-squares-adjustment analyses with the FERRET code. The conventional integral data-testing analysis, based on C/E ratios, indicates discrepancies outside the estimated integral test uncertainty for the 6 Li(n,He), 10 B(n,He), 47 Ti(n,p), 58 Fe(n,γ), 197 Au(n,γ) and 232 Th(n,γ) cross sections. The integral test uncertainty included contributions from the measured integral data and from the spectrum and cross sections used to obtain the calculated integral data. Within the uncertainty and correlation specifications for the input spectrum and dosimeter cross sections, the least-squares-adjustment analysis indicated a high degree of consistency between the measured integral data and the ENDF/B-V dosimeter cross sections for all reactions except 10 B
Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2) multi-group (199 n + 42 γ) cross section library in AMPX format for nuclear fission applications

Science.gov (United States)

Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

2017-09-01

The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ) cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data). VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data) and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.
Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2 multi-group (199 n + 42 γ cross section library in AMPX format for nuclear fission applications

Directory of Open Access Journals (Sweden)

Pescarini Massimo

2017-01-01

Full Text Available The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data. VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.
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
Evaluation of neutron cross-sections for 242Cm to obtain a complete file

International Nuclear Information System (INIS)

Bakhanovich, L.A.; Klepetskij, A.B.; Maslov, V.M.; Porodzinskij, Yu.V.; Sukhovitskij, E.Sh.

1994-01-01

Experimental fission, capture, inelastic scattering, (n2n), (n3n) and other cross-sections are scarce or unavailable. As a consequence, theoretical models and parameters systematics have been used extensively in the calculation of these data. Data obtained in this work are compared with previous evaluations. Severe discrepancies were found. (author). 10 refs, 2 figs, 2 tabs
A revised calculational model for fission

Energy Technology Data Exchange (ETDEWEB)

Atchison, F

1998-09-01

A semi-empirical parametrization has been developed to calculate the fission contribution to evaporative de-excitation of nuclei with a very wide range of charge, mass and excitation-energy and also the nuclear states of the scission products. The calculational model reproduces measured values (cross-sections, mass distributions, etc.) for a wide range of fissioning systems: Nuclei from Ta to Cf, interactions involving nucleons up to medium energy and light ions. (author)
Some aspects of the nuclear fission process; Quelques aspects du processus de fission nucleaire

Energy Technology Data Exchange (ETDEWEB)

Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

In the following report one can find first a short general view on the present situation of our knowledge concerning the nuclear fission process, namely on the nucleus going through the saddle-point. Then there are some aspects connected with the excitation energy of the fissioning nucleus. The measurements made at Saclay on the fast neutron fission cross-section of U{sup 233}, U{sup 235}, Pu{sup 239}, U{sup 238} are described at the beginning of this work. It appears that for U{sup 233} there is some characteristic shape modulation of the cross-section curve, in relation with the collective excited state of the deformed nucleus at the saddle-point. Good evidence of this is also given by the study of the relative fission rate with emission of long-range particles; it appears also that this ternary fission rate does not change substantially for neutron between thermal energy and 2 MeV, but that is very lower for the compound nucleus U{sup 239} than for even-even compound nuclei. At the end there are some experiments on the strong 4,5 MeV gamma-ray originated by slow neutron absorption in U{sup 235}. Time-of-flight device is used to establish that this 4,5 MeV gamma-ray seems mostly connected with radiative capture. (author) [French] Le present travail debute par un apercu de l'etat actuel de nos connaissances sur le processus de fission nucleaire, notamment sur le passage par le point-seuil. Puis sont evoques des aspects lies au niveau d'energie d'excitation auquel est porte le noyau qui subit la fission. Les mesures de sections efficaces de fission induite dans {sup 233}U, {sup 235}U, {sup 239}Pu et {sup 238}U par des neutrons rapides effectuees a Saclay sont decrites en premier lieu; elles font apparaitre pour {sup 233}U une ondulation caracteristique du role des etats collectifs d'excitation du noyau deforme au point-seuil. Des experiences sur la fission avec emission de particules de long parcours confirment cet aspect tout en demontrant que la frequence
Program RECENT (version 79-1): reconstruction of energy-dependent neutron cross sections from resonance parameters in the ENDF/B format

International Nuclear Information System (INIS)

Cullen, D.E.

1979-01-01

Program RECENT reconstructs energy-dependent neutron total, elastic, capture, and fission cross sections from a combination of resonance parameters and tabulated background cross sections in the ENDF/B format. Entire evaluations, not just cross sections, are written to the result file, which is in ENDF/B format. The output includes the original resonance parameters in a form that can be used in Doppler broadening and self-shielding calculations. A listing of the source deck is available on request. 5 figures, 5 tables
Resonance structure in the fission of ( sup 235 U+n)

Energy Technology Data Exchange (ETDEWEB)

Moore, M.S. (Los Alamos National Lab. (LANL), NM (USA). Physics Div.); Leal, L.C.; De Saussure, G.; Perez, R.B.; Larson, N.M. (Oak Ridge National Lab., TN (USA))

1989-10-09

A new multilevel reduced R-matrix analysis of the neutron-induced resonance cross sections of {sup 235}U has been carried out. We used as a constraint in the analysis the angular anisotropy measurements of Pattenden and Postma, obtaining a Bohr-channel (or J, K channel) representation of the resonances in a two-fission vector space for each spin state. Hambsch et al., have reported definitive measurements of the mass- and kinetic-energy distributions of fission fragments of ({sup 235}U+n) in the resonance region and analyzed their results according to the fission-channel representation of Brosa et al., extracting relative contributions of the two asymmetric and one symmetric Brosa fission channels. We have explored the connection between Bohr-channel and asymmetric Brosa-channel representations. The results suggest that a simple rotation of coordinates in channel space may be the only transformation required; the multilevel fit to the total and partial cross sections is invariant to such a transformation. (orig.).
A macroscopic cross-section model for BWR pin-by-pin core analysis

International Nuclear Information System (INIS)

Fujita, Tatsuya; Endo, Tomohiro; Yamamoto, Akio

2014-01-01

A macroscopic cross-section model used in boiling water reactor (BWR) pin-by-pin core analysis is studied. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of core state and depletion history variables and are tabulated prior to core calculations. Variations of cross sections in a core simulator are caused by two different phenomena (i.e. instantaneous and history effects). We treat them through the core state variables and the exposure-averaged core state variables, respectively. Furthermore, the cross-term effect among the core state and the depletion history variables is considered. In order to confirm the calculation accuracy and discuss the treatment of the cross-term effect, the k-infinity and the pin-by-pin fission rate distributions in a single fuel assembly geometry are compared. Some cross-term effects could be negligible since the impacts of them are sufficiently small. However, the cross-term effects among the control rod history (or the void history) and other variables have large impacts; thus, the consideration of them is crucial. The present macroscopic cross-section model, which considers such dominant cross-term effects, well reproduces the reference results and can be a candidate in practical applications for BWR pin-by-pin core analysis on the normal operations. (author)
Observation of fission residues in the 16O + 181Ta system at Elab ≈ 6 MeV/A

Directory of Open Access Journals (Sweden)

Singh B. P.

2011-10-01

Full Text Available Present paper reports on the production cross-section of 24 fission like events (30 ≤ Z ≤ 60 formed via complete fusion-fission and/or incomplete fusion-fission processes in 16O+181Ta system at energies ≈ 6 MeV/A. Experiments have been performed using the recoil-catcher technique followed by off-line γ-spectroscopy. The measured cross-section of fission-like events is satisfactorily described by a statistical model code. Further, an attempt has been made to study the mass and isotopic yield distributions of fission fragments. The variance of the presently measured isotopic yield distributions has been found to be in agreement with the literature values for some other fissioning systems.
XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections

International Nuclear Information System (INIS)

Ganesan, S.; Jagannathan, V.; Thiyagarajan, T.K.

2005-01-01

1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections
Uranium content of petroleum by Fission track technique

International Nuclear Information System (INIS)

Paschaa, A.S.; Mafra, O.Y.; Oliveira, C.A.N.; Pinto, L.R.

1982-01-01

This paper examines the feasibility of the fission track registration technique to investigate the natural uranium concentration in petroleum. The application is briefly described and the results obtained indicate the presence of uranium concentrations in samples of Brazilian petroleum which are over the detection limit of the fission track technique. The irradiations were performed by using fluxes with predominance of thermal neutrons, which have a fission cross-section for U 235 equal to 579 barns. Since the neutron fluxes were not comp sed exclusively of thermal neutrons, fissions from fast neutrons would also be taken into account for U 238 and Th 232
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.
Direct and preequilibrium effects in the fission-product mass range

International Nuclear Information System (INIS)

Gruppelaar, H.; Hogenbirk, A.

1992-07-01

Until recently inelastic scattering did not gain the proper attention in fission-product cross section evaluations. In many existing evaluations global spherical optical models have been used, neglecting direct and pre-equilibrium effects. There are also few experimental data relevant to inelastic scattering in fission products. This paper is focussed on the anomalously high inelastic scattering cross sections observed in even-mass nuclei near mass A=100 at low energies. Both more data and more refined theoretical analyses are required. A number of suggestions for relevant coupled-channel calculations is made. (author). 29 refs., 5 figs., 1 tab
235U and 238U (n,xn gamma) cross-sections

International Nuclear Information System (INIS)

Bacquias, A.; Dessagne, Ph.; Kerveno, M.; Rudolf, G.; Thiry, J.C.; Borcea, C.; Negret, A.L.; Drohe, J.C.; Nankov, N.; Nyman, M.; Plompen, A.; Rouki, C.; Stanoiu, M.

2014-01-01

The (n,n') and (n,2n) are important processes in the energy domain of fission neutrons, but the cross-sections suffer from large uncertainties, not compatible with the objectives fixed for future and advanced nuclear reactors. This paper presents our experimental effort to improve 235 U and 238 U (n,xnγ) cross-section data. The experiments were performed at the GELINA facility (Belgium), which provides a pulsed (800 Hz) neutron beam covering a wide energy spectrum (from a few eV to about 20 MeV). The GRAPhEME set-up is designed for prompt gamma spectroscopy and time-of-flight measurement. The analysis methods are presented. Already published results on 235 U are shown, as well as results on 238 U. The interpretation and discussion rely on the comparison with TALYS and EMPIRE predictions. (authors)
Setup for fission and evaporation cross-section measurements in reactions induced by secondary beams

International Nuclear Information System (INIS)

Hassan, A.A.; Luk'yanov, S.M.; Kalpakchieva, R.; Skobelev, N.K.; Penionzhkevich, Yu.Eh.; Dlouhy, Z.; Radnev, S.; Poroshin, N.V.

2002-01-01

A setup for studying reactions induced by secondary radioactive beams has been constructed. It allows simultaneous measurement of α-particle and fission fragment energy spectra. By measuring the α-particles, identification of evaporation residues is achieved. A set of three targets can be used so as to ensure sufficient statistics. Two silicon detectors, located at 90 degrees to the secondary beam direction, face each target, thus covering 30% of the solid angle. This experimental setup is to be used to obtain excitation functions of fusion-fission reactions and of reactions leading to evaporation residue production
Spectral history correction of microscopic cross sections for the PBR using the slowing down balance

International Nuclear Information System (INIS)

Hudson, N.; Rahnema, F.; Ougouag, A. M.; Gougar, H. D.

2006-01-01

A method has been formulated to account for depletion effects on microscopic cross sections within a Pebble Bed Reactor (PBR) spectral zone without resorting to calls to the spectrum (cross section generation) code or relying upon table interpolation between data at different values of burnup. In this method, infinite medium microscopic cross sections, fine group fission spectra, and modulation factors are pre-computed at selected isotopic states. This fine group information is used with the local spectral zone nuclide densities to generate new cross sections for each spectral zone. The local spectrum used to generate these microscopic cross sections is estimated through the solution to the cell-homogenized, infinite medium slowing down balance equation during the flux calculation. This technique is known as Spectral History Correction (SHC), and it is formulated to specifically account for burnup within a spectral zone. It was found that the SHC technique accurately calculates local broad group microscopic cross sections with local burnup information. Good agreement is obtained with cross sections generated directly by the cross section generator. Encouraging results include improvement in the converged fuel cycle eigenvalue, the power peaking factor, and the flux. It was also found that the method compared favorably to the benchmark problem in terms of the computational speed. (authors)
Evaporation channel as a tool to study fission dynamics

Science.gov (United States)

Di Nitto, A.; Vardaci, E.; La Rana, G.; Nadtochy, P. N.; Prete, G.

2018-03-01

The dynamics of the fission process is expected to affect the evaporation residue cross section because of the fission hindrance due to the nuclear viscosity. Systems of intermediate fissility constitute a suitable environment for testing such hypothesis since they are characterized by evaporation residue cross sections comparable or larger than the fission ones. Observables related to emitted charged particles, due to their relatively high emission probability, can be used to put stringent constraints on models describing the excited nucleus decay and to recognize the effects of fission dynamics. In this work model simulations are compared with the experimental data collected via the 32S +100 Mo reaction at Elab = 200 MeV. Consequently we pointed out, exploring an extended set of evaporation channel observables, the limits of the statistical model and the large improvement obtained with a dynamical model. Moreover we stress the importance of using an apparatus covering a large fraction of 4π to extract observables. Finally, we discuss the opportunity to measure more sensitive observables by a new detection device in operation at LNL.

Capture cross sections for very heavy systems

International Nuclear Information System (INIS)

Rowley, N.; Grar, N.; Ntshangase, S.S.

2006-01-01

In intermediate-mass systems, collective excitations of the target and projectile can greatly enhance the sub-barrier capture cross section σ cap by giving rise to a distribution of Coulomb barriers. For such systems, capture essentially leads directly to fusion (formation of a compound nucleus (CN)), which then decays through the emission of light particles (neutrons, protons, and alpha particles). Thus the evaporation-residue (ER) cross section is essentially equal to σ cap . For heavier systems the experimental situation is significantly more complicated due to the presence of quasifission (QF) (rapid separation into two fragments before the CN is formed) and by fusion-fission (FF) of the CN itself. Thus three cross sections need to be measured in order to evaluate σ cap . Although the ER essentially recoil along the beam direction. QF and FF fragments are scattered to all angles and require the measurement of angular distribution in order to obtain the excitation function and barrier distribution for capture. Two other approaches to this problem exist. If QF is not important, one can still measure just the ER cross section and try to reconstruct the corresponding σ cap through use of an evaporation-model code that takes account of the FF degree of freedom. Some earlier results on σ cap obtained in this way will be re-analyzed with detail coupled-channels calculations, and the extra-push phenomenon discussed. One may also try to obtain σ cap by exploiting unitarity, that is, by measuring instead the flux of particles corresponding to quasielastic (QE) scattering from the Coulomb barrier. Some new QE results obtained for the 86 Kr + 208 Pb system at iThemba LABS in South Africa will be presented [ru
Reich-Moore and Adler-Adler representations of the 235U cross sections in the resolved resonance region

International Nuclear Information System (INIS)

de Saussure, G.; Leal, L.C.; Perez, R.B.

1990-01-01

In the first part of this paper, a reevaluation of the low-energy neutron cross sections of 235 U is described. This reevaluation was motivated by the discrepancy between the measured and computed temperature coefficients of reactivity and is based on recent measurements of the fission cross section and of η in the thermal and subthermal neutron energy regions. In the second part of the paper, we discuss the conversion of the Reich-Moore resonance parameters, describing the neutron cross sections of 235 U in the resolved resonance region, into equivalent Adler-Adler resonance parameters and into equivalent momentum space multipole resonance parameters. 25 refs., 4 figs., 5 tabs
On the fission probability for 235U, 239Pu and 241Pu

International Nuclear Information System (INIS)

Benzi, V.; Maino, G.; Menapace, E.

1978-01-01

An evaluation of the GAMMAsub(n)/GAMMAsub(f) ratio for the 236 U, 240 Pu and 242 Pu compound nuclei is carried out. First chance and second chance fission cross sections are estimated from the ''evaporation'' model; particularly, a largely increasing trend was found for the first chance fission cross section above the (n,n'f) process threshold. The GAMMAsub(n)/GAMMAsub(f) ratios for the analyzed nuclei show a bump-like structure, that seems to be in agreement with the theoretical predictions reported in literature
(d,p)-transfer induced fission of heavy radioactive beams

CERN Document Server

Veselsky, Martin

2012-01-01

(d,p)-transfer induced fission is proposed as a tool to study low energy fission of exotic heavy nuclei. Primary goal is to directly determine the fission barrier height of proton-rich fissile nuclei, preferably using the radio-active beams of isotopes of odd elements, and thus confirm or exclude the low values of fission barrier heights, typically extracted using statistical calculations in the compound nucleus reactions at higher excitation energies. Calculated fission cross sections in transfer reactions of the radioactive beams show sufficient sensitivity to fission barrier height. In the probable case that fission rates will be high enough, mass asymmetry of fission fragments can be determined. Results will be relevant for nuclear astrophysics and for production of super-heavy nuclei. Transfer induced fission offers a possibility for systematic study the low energy fission of heavy exotic nuclei at the ISOLDE.
R-matrix analyses of the 235U and 239Pu neutron cross sections

International Nuclear Information System (INIS)

Derrien, H.; de Saussure, G.; Larson, N.M.; Leal, L.C.; Perez, R.B.

1988-01-01

The resonance parameter analysis code SAMMY was used to perform consistent resonance analyses of several 235 U and 239 Pu fission and capture cross section and transmission measurements up to 110 eV for 235 U and up to 1 keV for 239 Pu. The method of analysis, the measurement selection and the results are briefly outlined in this paper
Formats and processing of evaluated nuclear data into multigroup cross-sections

International Nuclear Information System (INIS)

Motta, M.

1984-01-01

The first part of these lectures concerns the data in nuclear files and their manipulation. The structure of the data files as divided into the resonance region (subdivided into the resolved and the unresolved regions) and the continuum region is presented. The reactions concerned are the elastic scattering; the radiative capture and the fission methods for averaging the cross sections are given. Then, the group averaging formulas and the self-shielding factors are presented in some detail. The second part concerns a presentation of nuclear data files handling and conversion. The main libraries are listed and several maintenance computer codes presented. The way the conversion among different files is handled is also presented. The listings of several BASIC programs for different cross section calculations are given. These codes are self-guided
Investigation of the Impact of ENDF/B-VI Cross Sections on the H.B. Robinson-2 Pressure-Vessel Flux Prediction

International Nuclear Information System (INIS)

Remec, I

1999-01-01

This report discusses the impact of the change from the SAILOR cross-section library, based on the ENDF/B-IV data, to the BUGLE-96 cross-section library, based on the ENDF/B-VI data, on the neutron flux prediction in the H. B. Robinson-2 pressure vessel, in the surveillance capsule, and in the cavity. The fast flux (E > 1 MeV) from the transport calculations with the BUGLE-96 library is approximately6% higher in the surveillance capsule and at the PV inner wall, and approximately25% higher in the reactor cavity than the flux from the transport calculations with the SAILOR library. These changes result from the combined effect of the changes in the cross sections, which cause significant increases in the calculated fluxes, and much smaller decreases in the fast fluxes due to the changes in the fission spectra. The increase in the calculated fast flux due to the changes in the cross sections only is approximately9% in the capsule and at the pressure vessel (PV) wall, and approximately30% in the cavity. The changes in the fission spectra lead to decreases in the order of approximately3-4% in calculated fast fluxes
Fission properties of actinide nuclei from proton-induced fission at 26.5 and 62.9 MeV incident proton energies

International Nuclear Information System (INIS)

Demetriou, P.; Keutgen, Th.; Prieels, R.; El Masri, Y.

2010-01-01

Fission properties of proton-induced fission on 232 Th, 237 Np, 238 U, 239 Pu, and 241 Am targets, measured at the Louvain-la-Neuve cyclotron facility at proton energies of 26.5 and 62.9 MeV, are compared with the predictions of the state-of-the-art nuclear reaction code talys. The code couples the multimodal random neck-rupture model with the pre-equilibrium exciton and statistical models to predict fission fragment mass yields, pre- and post-scission neutron multiplicities, and total fission cross sections in a consistent approach. The sensitivity of the calculations to the input parameters of the code and possible improvements are discussed in detail.
Non-compound nucleus fission in actinide and pre-actinide regions

Indian Academy of Sciences (India)

Data on the evaporation residue cross-sections, in addition to those on mass and angular distributions, are necessary for better understanding of the contribution from non-compound nucleus fission in the pre-actinide region. Measurement of mass-resolved angular distribution of fission products in 20Ne+232Th reaction ...
New Burnup Calculation System for Fusion-Fission Hybrid System

International Nuclear Information System (INIS)

Isao Murata; Shoichi Shido; Masayuki Matsunaka; Keitaro Kondo; Hiroyuki Miyamaru

2006-01-01

Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise
Fusion barrier distributions and fission anisotropies

International Nuclear Information System (INIS)

Hinde, D.J.; Morton, C.R.; Dasgupta, M.; Leigh, J.R.; Lestone, J.P.; Lemmon, R.C.; Mein, J.C.; Newton, J.O.; Timmers, H.; Rowley, N.; Kruppa, A.T.

1995-01-01

Fusion excitation functions for 16,17 O+ 144 Sm have been measured to high precision. The extracted fusion barrier distributions show a double-peaked structure interpreted in terms of coupling to inelastic collective excitations of the target. The effect of the positive Q-value neutron stripping channel is evident in the reaction with 17 O. Fission and evaporation residue cross-sections and excitation functions have been measured for the reaction of 16 O+ 208 Pb and the fusion barrier distribution and fission anisotropies determined. It is found that the moments of the fusion l-distribution determined from the fusion and fission measurements are in good agreement. ((orig.))
Measurement and analysis of thorium fission rate in a polyethylene shell with a D-T neutron source

Energy Technology Data Exchange (ETDEWEB)

Zheng, Lei [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Yang, Yiwei, E-mail: winfield1920@126.com [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Liu, Zhujun [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Department of Nuclear Engineering and Technology, Sichuan University, Chengdu 610065,China (China); Liu, Rong, E-mail: liurongzy@163.com [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Jiang, Li; Wang, Mei [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China)

2016-12-15

Highlights: • Associated angular dependencies of the source neutron energy and intensity was given. • Two sets of fission yields from evaluated libraries were considered and applied. • Calculated results employing ENDF/B-VII.0 agreed with the experimental ones best. • Small discrepancies exist between thorium fission cross section evaluated libraries. - Abstract: In order to validate the {sup 232}Th fission cross section, an integral experiment was carried out using the activation method in a polyethylene shell with a D-T neutron source. Thorium samples were arranged in the 0° direction to the incident D{sup +} beam. The {sup 232}Th fission rate was determined by measuring the 151.195 keV characteristic γ ray emitted from the fission fragment {sup 85m}Kr, and the experimental uncertainties were about 5.3%. MCNP calculation results employing ENDF/B-VII.0, JENDL-3.3, JENDL-4.0 libraries are in good agreement with that of experiments within uncertainties except that employing ENDF/B-VII.1 (∼6.5%). The experiment results can be used to re-evaluate the {sup 232}Th fission cross section.
MIRANDA - a module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system

International Nuclear Information System (INIS)

Robinson, G.S.

1985-12-01

MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data, resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B L flux solution, and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual
Intermediate energy nuclear fission

International Nuclear Information System (INIS)

Hylten, G.

1982-01-01

Nuclear fission has been investigated with the double-kinetic-energy method using silicon surface barrier detectors. Fragment energy correlation measurements have been made for U, Th and Bi with bremsstrahlung of 600 MeV maximum energy. Distributions of kinetic energy as a function of fragment mass are presented. The results are compared with earlier photofission data and in the case of bismuth, with calculations based on the liquid drop model. The binary fission process in U, Yb, Tb, Ce, La, Sb, Ag and Y induced by 600 MeV protons has been investigated yielding fission cross sections, fragment kinetic energies, angular correlations and mass distributions. Fission-spallation competition calculations are used to deduce values of macroscopic fission barrier heights and nuclear level density parameter values at deformations corresponding to the saddle point shapes. We find macroscopic fission barriers lower than those predicted by macroscopic theories. No indication is found of the Businaro Gallone limit expected to occur somewhere in the mass range A = 100 to A = 140. For Ce and La asymmetric mass distributions similar to those in the actinide region are found. A method is described for the analysis of angular correlations between complementary fission products. The description is mainly concerned with fission induced by medium-energy protons but is applicable also to other projectiles and energies. It is shown that the momentum and excitation energy distributions of cascade residuals leading to fission can be extracted. (Author)
FIZCON, ENDF/B Cross-Sections Redundancy Check

International Nuclear Information System (INIS)

Dunford, Charles L.

2007-01-01

1 - Description of program or function: FIZCON is a program for checking that an evaluated data file has valid data and conforms to recommended procedures. Version 7.01 (April 2005): set success flag after return from beginning; fixed valid level check for an isomer; fixed subsection energy range test in ckf9; changed lower limit on potential scattering test; fixed error in j-value test when l=0 and i=0; added one more significant figure to union grid check and sum up output messages; partial fission cross sections mt=19,20,21 and 38 did not require secondary energy distributions in file 5; corrected product test for elastic scattering; moved potential scattering test to psyche. Version 7.02 (May 2005): Fixed resonance parameter sum test. 2 - Method of solution: FIZCON can recognise the difference between ENDF-6 and ENDF-5 formats and performs its tests accordingly. Some of the tests performed include: data arrays are in increasing energy order; resonance parameter widths add up to the total; Q-values are reasonable and consistent; no required sections are missing and all cover the proper energy range; secondary distributions are normalized to 1.0; energy conservation in decay spectra. Optional tests can be performed to check the redundant cross sections, and algorithms can be used to check for possible incorrect entry of data values (Deviant Point test)
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
BARC 75 - A 75 group neutron-photon coupled cross-section library with P5- anisotropic scattering matrices

International Nuclear Information System (INIS)

Garg, S.B.

1990-01-01

A 75 group neutron-photon coupled cross-section library has been developed for 42 reactor nuclides utilizing the basic cross-section files - ENDF/B-IV for neutrons and DLC-7F for photons. 50 neutron energy groups and gamma energy groups are included in this library which should be well suited to carry out safety, shielding and core physics studies of nuclear reactors based on fission or fusion processes. This library is also adequate for oil logging and mineral exploration investigations. (author). 11 refs., 3 tabs
Investigating the fission process at high excitation energies through proton induced reactions on 181Ta

International Nuclear Information System (INIS)

Ayyad, Y.; Benlliure, J.; Casajeros, E.; Alvarez Pol, H.; Paradela, C.; Perez-Loureido, D.; Tarrio, D.; Bacquias, A.; Boudard, A.; Kezzar, K.; Leray, S.; Enqvist, T.; Foehr, V.; Kelic, A.; Pleskac, R.

2010-01-01

In this work we have investigated the total fission cross section of 181 Ta + 1 H at FRS (Fragment Separator - GSI) at 1, 0.8, 0.5 and 0.3 GeV with a specific setup, providing high accuracy measurements of the cross section values. the comparison of our data with previous results reveals a good agreement at high energies. However the situation remains unclear at lower energies. In general, our results covering a wide range of energy, are smoother. We have also compared the results obtained in this experiment, with several calculations performed with the intra-nuclear cascade model (INCL v4.1) coupled to de-excitation code (ABLAv3p), according to two different models describing fission process at high-excitation energies: statistical model of Bohr and Wheeler and the dynamical description of the fission process. We have showed that a simple statistical description largely over-predict the measured cross-section. Only a dynamical description of the fission, involving the role of the viscosity of the nuclear matter, provides a realistic result.
Identification of fission-like events in the 16O + 181Ta system: Mass and isotopic yield distribution

International Nuclear Information System (INIS)

Sharma, Vijay R.; Yadav, Abhishek; Singh, Devendra P.; Unnati,; Singh, B. P.; Prasad, R.; Singh, Pushpendra P.; Sharma, Manoj K.; Kumar, R.; Golda, K. S.; Sinha, A. K.

2011-01-01

In this paper, nuclear reaction cross sections for 24 fission-like fragments (30≤Z≤60) have been measured for the 6.5 MeV/A 16 O + 181 Ta system. The recoil-catcher activation technique was employed followed by off-line γ spectroscopy. The isotopic yield distributions for yttrium and indium isotopes have been obtained from the experimental data. The variance of the presently measured isotopic yield distributions have been found to be in agreement with the literature values. However, the variance of the mass distribution of fission residues has found to be narrower as compared to other relatively heavier systems. A self-consistent approach to determining the isobaric charge dispersion parameters has been adopted. The measured fission cross sections at 97 and 100 MeV are satisfactorily described by a statistical model code. An attempt has been made to explain the production cross sections of intermediate mass residues in the fission of heavy residues populated via complete and/or incomplete fusion processes.
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.

Fusion-fission of superheavy nuclei at low excitation energies

International Nuclear Information System (INIS)

Itkis, M.G.; Oganesyan, Yu.Ts.; Kozulin, E.M.

2000-01-01

The process of fusion-fission of superheavy nuclei with Z = 102 -122 formed in the reactions with 22 Ne, 26 Mg, 48 Ca, 58 Fe and 86 Kr ions at energies near and below the Coulomb barrier has been studied. The experiments were carried out at the U-400 accelerator of the Flerov Laboratory of Nuclear Reactions (JINR) using a time-of-flight spectrometer of fission fragments CORSET and a neutron multi-detector DEMON. As a result of the experiments, mass and energy distributions of fission fragments, fission and quasi-fission cross sections, multiplicities of neutrons and gamma-rays and their dependence on the mechanism of formation and decay of compound superheavy systems have been studied
From cutting-edge pointwise cross-section to groupwise reaction rate: A primer

Science.gov (United States)

Sublet, Jean-Christophe; Fleming, Michael; Gilbert, Mark R.

2017-09-01

The nuclear research and development community has a history of using both integral and differential experiments to support accurate lattice-reactor, nuclear reactor criticality and shielding simulations, as well as verification and validation efforts of cross sections and emitted particle spectra. An important aspect to this type of analysis is the proper consideration of the contribution of the neutron spectrum in its entirety, with correct propagation of uncertainties and standard deviations derived from Monte Carlo simulations, to the local and total uncertainty in the simulated reactions rates (RRs), which usually only apply to one application at a time. This paper identifies deficiencies in the traditional treatment, and discusses correct handling of the RR uncertainty quantification and propagation, including details of the cross section components in the RR uncertainty estimates, which are verified for relevant applications. The methodology that rigorously captures the spectral shift and cross section contributions to the uncertainty in the RR are discussed with quantified examples that demonstrate the importance of the proper treatment of the spectrum profile and cross section contributions to the uncertainty in the RR and subsequent response functions. The recently developed inventory code FISPACT-II, when connected to the processed nuclear data libraries TENDL-2015, ENDF/B-VII.1, JENDL-4.0u or JEFF-3.2, forms an enhanced multi-physics platform providing a wide variety of advanced simulation methods for modelling activation, transmutation, burnup protocols and simulating radiation damage sources terms. The system has extended cutting-edge nuclear data forms, uncertainty quantification and propagation methods, which have been the subject of recent integral and differential, fission, fusion and accelerators validation efforts. The simulation system is used to accurately and predictively probe, understand and underpin a modern and sustainable understanding
From cutting-edge pointwise cross-section to groupwise reaction rate: A primer

Directory of Open Access Journals (Sweden)

Sublet Jean-Christophe

2017-01-01

Full Text Available The nuclear research and development community has a history of using both integral and differential experiments to support accurate lattice-reactor, nuclear reactor criticality and shielding simulations, as well as verification and validation efforts of cross sections and emitted particle spectra. An important aspect to this type of analysis is the proper consideration of the contribution of the neutron spectrum in its entirety, with correct propagation of uncertainties and standard deviations derived from Monte Carlo simulations, to the local and total uncertainty in the simulated reactions rates (RRs, which usually only apply to one application at a time. This paper identifies deficiencies in the traditional treatment, and discusses correct handling of the RR uncertainty quantification and propagation, including details of the cross section components in the RR uncertainty estimates, which are verified for relevant applications. The methodology that rigorously captures the spectral shift and cross section contributions to the uncertainty in the RR are discussed with quantified examples that demonstrate the importance of the proper treatment of the spectrum profile and cross section contributions to the uncertainty in the RR and subsequent response functions. The recently developed inventory code FISPACT-II, when connected to the processed nuclear data libraries TENDL-2015, ENDF/B-VII.1, JENDL-4.0u or JEFF-3.2, forms an enhanced multi-physics platform providing a wide variety of advanced simulation methods for modelling activation, transmutation, burnup protocols and simulating radiation damage sources terms. The system has extended cutting-edge nuclear data forms, uncertainty quantification and propagation methods, which have been the subject of recent integral and differential, fission, fusion and accelerators validation efforts. The simulation system is used to accurately and predictively probe, understand and underpin a modern and
Needs and accuracy requirements for fission product nuclear data in the physics design of power reactor cores

International Nuclear Information System (INIS)

Rowlands, J.L.

1978-01-01

The fission product nuclear data accuracy requirements for fast and thermal reactor core performance predictions were reviewed by Tyror at the Bologna FPND Meeting. The status of the data was assessed at the Meeting and it was concluded that the requirements of thermal reactors were largely met, and the yield data requirements of fast reactors, but not the cross section requirements, were met. However, the World Request List for Nuclear Data (WRENDA) contains a number of requests for fission product capture cross sections in the energy range of interest for thermal reactors. Recent reports indicate that the fast reactor reactivity requirements might have been met by integral measurements made in zero power critical assemblies. However, there are requests for the differential cross sections of the individual isotopes to be determined in addition to the integral data requirements. The fast reactor requirements are reviewed, taking into account some more recent studies of the effects of fission products. The sodium void reactivity effect depends on the fission product cross sections in a different way to the fission product reactivity effect in a normal core. This requirement might call for different types of measurement. There is currently an interest in high burnup fuel cycles and alternative fuel cycles. These might require more accurate fission product data, data for individual isotopes and data for capture products. Recent calculations of the time dependence of fission product reactivity effects show that this is dependent upon the data set used and there are significant uncertainties. Some recent thermal reactor studies on approximations in the treatment of decay chains and the importance of xenon and samarium poisoning are also summarized. (author)
Thorium-uranium fission radiography

Science.gov (United States)

Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

1976-01-01

Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.
AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

International Nuclear Information System (INIS)

1990-01-01

1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup
Fast-neutron capture cross sections for the most important fission-product nuclei

International Nuclear Information System (INIS)

Gruppelaar, H.

1982-01-01

The main activity of the fission-product (FP) Working Group was the discussion of the current status of neutron capture knowledge of the most important FP nuclides, including the formulation of recommendations toward improved understanding. The results of the discussion are summarized. General conclusions and recommendations are given. The status of integral data is summarized by R. Anderl; and nuclear models and calculations are reviewed by D. Gardner and G. Reffo
Proposal to represent neutron absorption by fission products by a single pseudo-fragment

International Nuclear Information System (INIS)

Tsibulya, A.M.; Kochetkov, A.L.; Kravchenko, I.V.; Nikolaev, M.N.

1991-01-01

The concentration of fission products during reactor operation is analyzed. The dependence of a composite fission product capture cross-section as a function of time and on the nature of the A of the fissile nuclide are investigated, and the neutron radiative capture in fission products of a thermal reactor is evaluated. It is concluded that neutron absorption by fission products can be described by pseudo-fragments. (author). 18 refs, 2 figs, 3 tabs
Measurement of the ${240}$Pu(n,f) reaction cross-section

CERN Multimedia

Following proposal CERN-INTC-2010-042 / INTC-P-280 (“Measurement of the fission cross-section of $^{240}$Pu and $^{242}$Pu at CERN’s n_TOF Facility”), the parallel measurement of the $^{240}$Pu(n,f) and $^{242}$Pu(n,f) reaction cross-sections was carried out at n_TOF EAR-1. While the $^{242}$Pu measurement was successful, unexpected sample-induced damage to the detectors caused by the high α-activity of the 240Pu samples resulted in a deterioration of the detector performance over the data taking period of several months, which compromised the measurement. This obstacle can be eliminated by performing the measurement in EAR-2, where the higher neutron flux will allow collecting data in a much shorter time, thus preventing the degradation of the detectors. In addition to this obvious advantage, the measurement would also benefit from the stronger suppression of the sample-induced α-background, due to the shorter times-of-flight involved.
Extensive set of low-fidelity cross sections covariances in fast neutron region

International Nuclear Information System (INIS)

Pigni, M.T.; Herman, M.; Oblozinsky, P.

2008-01-01

We produced a large set of neutron cross section covariances in the energy range of 5 keV - 20 MeV. The covariance matrices were calculated for 307 isotopes divided into three major regions: structural materials, fission products, and heavy nuclei. These results have been developed to provide initial, but consistent estimates of covariance data for nuclear criticality safety applications. The methodology for the determination of such covariance matrices is presented. It combines the nuclear reaction model code EMPIRE which calculates sensitivity of cross sections to nuclear reaction model parameters, and the Bayesian code KALMAN that propagates uncertainties of the model parameters to cross sections. Taking into account large number of materials, only marginal reference to experimental data was made. The covariances were derived from the perturbation of several key model parameters selected by the sensitivity analysis. These parameters refer to the optical model potential, the level densities and the strength of the pre-equilibrium emission. This work represents the first try ever to generate nuclear data covariances on such a large scale. (authors)
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)
Cross-sections of spallation residues produced in 1A GeV 208Pb on proton reactions

International Nuclear Information System (INIS)

Wlazlo, W.; Uniwersytet Jagiellonski, Cracow; Enqvist, T.; Armbruster, P.

2000-02-01

Spallation residues produced in 1 GeV per nucleon 208 Pb on proton reactions have been studied using the fragment separator facility at GSI. Isotopic production cross-sections of elements from 61 Pm to 82 Pb have been measured down to 0.1 mb with a high accuracy. The recoil kinetic energies of the produced fragments were also determined. The obtained cross-sections agree with most of the few existing gamma-spectroscopy data. Data are compared with different intranuclear-cascade and evaporation-fission models. Drastic deviations were found for a standard code used in technical applications. (orig.)
Effect of γ-ray emission on transuranium element production cross sections in heavy ion reactions

International Nuclear Information System (INIS)

Il'inov, A.S.; Oganesyan, Yu.Ts.; Cherepanov, E.A.

1980-01-01

The effect of competition of the γ ray emission with neutron evaporation and of compound nuclei fission induced by heavy ion reactions on the production cross sections for transuranium elements is considered. It is shown that taking account of γ ray emission leads to the broadening of the excitation functions of the (HI, xny) reactions such as 18 O+ 238 U, 40 Ar+ 206 Pb, 40 Ar+ 207 Pb and 40 Ar+ 208 Pb reactions and to the displacement of their maximum toward the higher energies as well as to an increase of the absolute cross sections which is especially strong close to the fusion barrier. Cross sections for the radiative capture of heavy ions by a heavy target nucleus in 40 Ar+ 206 Pb, 40 Ar+ 208 Pb, 48 Ca+ 204 Pb and 48 Ca+ 208 Pb reactions are estimated
Statistical model calculations with a double-humped fission barrier GIVAB computer code

International Nuclear Information System (INIS)

Delagrange, H.; Gilat, J.

1979-01-01

Neutron and gamma emission probabilities and fission probabilities are computed, taking into account the special feature of the actinide fission barriers with two maxima. Spectra and cross sections are directly deduced from these probabilities. Populations of both wells are followed step by step. For each initial E and J, decay rates are computed and normalized in order to obtain the de-excitation probabilities imposed by the two-humped fission barrier
Fission characteristics of Ra formed in heavy-ion induced reactions

Indian Academy of Sciences (India)

A Kramers-modified statistical model is used to calculate the cross-section of the evap- oration residue, fission ... where ρCN and ρsad are the level density of the compound nucleus at the ground and saddle points ... where P(K) is the probability that the system is in a given K. P(K) = T ..... time to be emitted before fission.
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
On unambiguous parametrization of neutron cross-sections in the low-energy region

International Nuclear Information System (INIS)

Novoselov, G.M.; Kolomiets, V.M.

1982-08-01

One of the most important aims of analysis in the resonance region is the evaluation of neutron resonance parameters on the basis of a given formalism of the theory of nuclear reactions. However, the task of finding resonance parameters from experimental data on the energy dependence of cross-sections is subject to a number of difficulties. These difficulties are not only of a theoretical character associated with the selection of one version or another of the theory taking into account the effects necessary (interference between resonances, Doppler effect etc.), but also involve problems of principle. Whether the set of parameters found is the only possible one within the context of a single formalism used remains open. The specific features of processing the experimental data are such that even with good resolution a number of overlapping resonances (occurring as a result of the fluctuation in inter-level distances or the Doppler effect) may be classified as an isolated resonance. Moreover, even given a very weak inter-level interference and Doppler effect, unambiguous parametrization of the cross-sections is not always possible. In the present paper these questions (the choice of the approximation needed for describing experimentally observed cross-sections, allowance for inter-level interference and the Doppler effect and the possibility of ambiguous reproduction of the resonance structure of cross-sections) are examined with reference to the parametrization of the total cross-sections for non-fissionable nuclei in the low-neutron-energy region
RESEND, Infinitely Dilute Point Cross-Sections Calculation from ENDF/B Resonance Parameter. ADLER, ENDF/B Adler-Adler Resonance Parameter to Point Cross-Sections with Doppler Broadening

International Nuclear Information System (INIS)

Bhat, M.R.; Ozer, O.

1982-01-01

1 - Description of problem or function: RESEND generates infinitely- dilute, un-broadened, point cross sections in the ENDF format by combining ENDF File 3 background cross sections with points calculated from ENDF File 2 resonance parameter data. ADLER calculates total, capture, and fission cross sections from the corresponding Adler-Adler parameters in the ENDF/B File 2 Version II data and also Doppler-broadens cross sections. 2 - Method of solution: RESEND calculations are done in two steps by two separate sections of the program. The first section does the resonance calculation and stores the results on a scratch file. The second section combines the data from the scratch file with background cross sections and prints the results. ADLER uses the Adler-Adler formalism. 3 - Restrictions on the complexity of the problem: RESEND expects its input to be a standard mode BCD ENDF file (Version II/III). Since the output is also a standard mode BCD ENDF file, the program is limited by the six significant figure accuracy inherent in the ENDF formats. (If the cross section has been calculated at two points so close in energy that only their least significant figures differ, that interval is assumed to have converged, even if other convergence criteria may not be satisfied.) In the unresolved range the cross sections have been averaged over a Porter-Thomas distribution. In some regions the calculated resonance cross sections may be negative. In such cases the standard convergence criterion would cause an unnecessarily large number of points to be produced in the region where the cross section becomes zero. For this reason an additional input convergence criterion (AVERR) may be used. If the absolute value of the cross section at both ends of an interval is determined to be less than AVERR then the interval is assumed to have converged. There are no limitations on the total number of points generated. The present ENDF (Version II/III) formats restrict the total number of
Investigation of excitation functions using new evaluated empirical and semi-empirical systematic for 14-15 MeV (n, t) reaction cross sections

International Nuclear Information System (INIS)

Tel, E.; Aydin, E. G.; Aydin, A.; Kaplan, A.

2007-01-01

The hybrid reactor is a combination of the fusion and fission processes. In the fusion-fission hybrid reactor, tritium self-sufficiency must be maintained for a commercial power plant. For self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. Working out the systematics of (n,t) reaction cross sections is of great importance for the definition of the excitation function character for the given reaction taking place on various nuclei at energies up to 20 MeV. In this study, we have investigated the asymmetry term effect for the (n,t) reaction cross sections at 14-15 neutron incident energy. It has been discussed the odd even effect and the pairing effect considering binding energy systematic of the nuclear shell model for the new experimental data and new cross section formulas (n,t) reactions developed by Tel et al. We have determined a different parameter groups by the classification of nuclei into even-even, even-odd and odd-even for (n,t) reactions cross sections. The obtained empirical formulas by fitting two parameter for (n,t) reactions were given. All calculated results have been compared with the experimental data. By using the new cross sections formulas (n,t) reactions the obtained results have been discussed and compared with the available experimental data
Multilevel fitting of 235U resonance data sensitive to Bohr-and Brosa-fission channels

International Nuclear Information System (INIS)

Moore, M.S.

1995-01-01

The recent determination of the K, J dependence of the neutron induced fission cross section of 235 U by the Dubna group has led to a renewed interest in the mechanism of fission from saddle to scission. The K quantum numbers designate the so-called Bohr fission channels, which describe the fission properties at the saddle point. Certain other fission properties, e.g., the fragment mass and kinetic-energy distribution, are related to the properties of the scission point. The neutron energy dependence of the fragment kinetic energies has been measured by Hambsch et al., who analyzed their data according to a channel description of Brosa et al. How these two channel descriptions, the saddle-point Bohr channels and the scission-point Brosa channels, relate to one another is an open question, and is the subject matter of the present paper. We use the correlation coefficient between various data sets, in which variations are reported from resonance to resonance, as a measure of both-the statistical reliability of the data and of the degree to which different scission variables relate to different Bohr channels. We have carried out an adjustment of the ENDF/B-VI multilevel evaluation of the fission cross section of 235 U, one that provides a reasonably good fit to the energy dependence of the fission, capture, and total cross sections below 100 eV, and to the Bohr-channel structure deduced from an earlier measurement by Pattenden and Postma. We have also further explored the possibility of describing the data of Hambsch et al. in the Brosa-channel framework with the same set of fission-width vectors, only in a different reference system. While this approach shows promise, it is clear that better data are also needed for the neutron energy variation of the scission-point variables

ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

International Nuclear Information System (INIS)

2002-01-01

and fission cross sections were obtained from collapsed 89-group ENDF/B-V and -VI data from the WIMS-AECL lattice code. Other reaction cross sections were obtained from the SCALE 27-group ENDF/B-IV data. In all, cross sections for more than 200 important actinides and fission products were updated with burnup-dependent data. The source of the nuclear decay data and cross sections not updated were the base ORIGEN-S libraries distributed with the SCALE 4.2 code package. The burnup-dependent cross sections were generated using eight burnup intervals that extend up to a burnup of approximately 12000 MWd/MgU. The cross sections for each interval are stored in positions on the ORIGEN-S library and are accessed by referencing the position number. The burnup values associated with each position are listed below. They are the same for both the 37-element and 28-element libraries. Library Position → Burnup (MWd/MgU): 1 → 240; 2 → 720; 3 → 1440; 4 → 2880; 5 → 4800; 6 → 6720; 7 → 8640; 8 → 10560
Fission and fragmentation of silver and bromine nuclei by 1-6 GeV energy photons

International Nuclear Information System (INIS)

Pinheiro Filho, J. de D.

1983-01-01

Fission and fragmentation of silver and bromine nuclei induced by bremsstrahlung photons in the maximum energy range of 1-6 GeV are studied. A special technique of nuclear emulsion for the highly ionizing nuclear fragment detection is used in the discrimination between nuclear fission and fragmentation events. Films of Ilford-KO nuclear emulsion (approximatelly 10 20 atoms/cm 2 of Ag, Br) which had been exposed to bremsstrahlung beams in 'Deutsches Elektronen Synchrotron' (DESY, Hamburg) with total doses of approximatelly 10 11 equivalent photons are used. Through a detailed analysis of range, angular and angle between fragment distributions, and empirical relations which permit to estimate nuclear fragment energy, range and velocity, the discrimination between fission and fragmentation events is made. Results related to fragment range distribution, angular distribution, distribution of angle between fragments, distribution of ratio between ranges, velocity distributions, forward/backward ratio, fission and fragmentation cross sections, nuclear fissionability and ternary fission frequency are presented and discussed. The results show that the mean photofragmentation cross section in the internal 1-6 GeV (0,09+-0,02mb) is significant when compared to the photofission (0,29+-0,05mb). It is also shown that the mean photofission cross section between 1 and 6 GeV is great by a factor of approximatelly 10 when compared to the foreseen by the cascade-evaporation nuclear model for monoenergetic photons of 0,6 GeV. (L.C.) [pt
New signatures on dissipation from fission induced by relativistic heavy-ion collisions

Energy Technology Data Exchange (ETDEWEB)

Jurado, B.; Schmitt, C.; Schmidt, K.H.; Enqvist, T.; Kelic, A.; Rejmund, F.; Benlliure, J. [Universidad de Santiago de Compostela (Spain); Junghans, A.R. [Forschungszentrum Rossendorf e.V. (FZR), Dresden (Germany)

2004-03-01

Fissile nuclei with small shape distortion relative to the ground-state deformation and with low angular momentum were produced in peripheral heavy-ion collisions. Under the conditions of small shape distortions and low angular momentum, the theoretical description of the fission process can be considerably simplified, and the relevant information on dissipation can be better extracted than in conventional experiments based on fusion-fission reactions. In addition, this experimental approach induces very high excitation energies, a condition necessary to observe transient effects. The experimental data were taken at GSI using a set-up especially conceived for fission studies in inverse kinematics. This set-up allowed determining three observables whose sensitivity to dissipation was investigated for the first time: the total fission cross sections of {sup 238}U at 1 A GeV as a function of the target mass, and, for the reaction of {sup 238}U at 1 A GeV on a (CH{sub 2}){sub n} target, the partial fission cross sections and the partial charge distributions of the fission fragments. The comparison of the new experimental data with a reaction code adapted to the conditions of the reactions investigated leads to clear conclusions on the strength of dissipation at small deformation where the existing results are rather contradictory. (orig.)
New signatures on dissipation from fission induced by relativistic heavy-ion collisions

International Nuclear Information System (INIS)

Jurado, B.; Schmitt, C.; Schmidt, K.H.; Enqvist, T.; Kelic, A.; Rejmund, F.; Benlliure, J.; Junghans, A.R.

2004-03-01

Fissile nuclei with small shape distortion relative to the ground-state deformation and with low angular momentum were produced in peripheral heavy-ion collisions. Under the conditions of small shape distortions and low angular momentum, the theoretical description of the fission process can be considerably simplified, and the relevant information on dissipation can be better extracted than in conventional experiments based on fusion-fission reactions. In addition, this experimental approach induces very high excitation energies, a condition necessary to observe transient effects. The experimental data were taken at GSI using a set-up especially conceived for fission studies in inverse kinematics. This set-up allowed determining three observables whose sensitivity to dissipation was investigated for the first time: the total fission cross sections of 238 U at 1 A GeV as a function of the target mass, and, for the reaction of 238 U at 1 A GeV on a (CH 2 ) n target, the partial fission cross sections and the partial charge distributions of the fission fragments. The comparison of the new experimental data with a reaction code adapted to the conditions of the reactions investigated leads to clear conclusions on the strength of dissipation at small deformation where the existing results are rather contradictory. (orig.)
Recent progress in fission at saddle point and scission point

International Nuclear Information System (INIS)

Blons, J.; Paya, D.; Signarbieux, C.

High resolution measurements of 230 Th and 232 Th fission cross sections for neutrons exhibit a fine structure. Such a structure is interpreted as a superposition of two rotational bands in the third, asymmetric, well of the fission barrier. The fragment mass distribution in the thermal fission of 235 U and 233 U does not show any even-odd effect, even at the highest kinetic energies. This is the mark of a strong viscosity in the descent from saddle point to scission point [fr
Angular distributions in the neutron-induced fission of actinides

CERN Multimedia

In 2003 the n_TOF Collaboration performed the fission cross section measurement of several actinides ($^{232}$Th, $^{233}$U, $^{234}$U, $^{237}$Np) at the n_TOF facility using an experImental setup made of Parallel Plate Avalanche Counters (PPAC). The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. We have been therefore able to cover the very broad neutron energy range 1eV-1GeV, taking full benefit of the unique characteristics of the n_TOF facility. Figure 1 shows an example obtained in the case of $^{237}$Np where the n_ TOF measurement showed that the cross section was underestimated by a large factor in the resonance region.
Evaluation of ETOG-3Q/ETOG-3, FLANGE-II, XLACS, NJOY and linear/recent/groupie codes for calculations of resonance and reference cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1991-01-01

The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analyzed. (author)
GROGi-F. Modified version of GROGi 2 nuclear evaporation computer code including fission decay channel

International Nuclear Information System (INIS)

Delagrange, H.

1977-01-01

This report is the user manual of the GR0GI-F code, modified version of the GR0GI-2 code. It calculates the cross sections for heavy ion induced fission. Fission probabilities are calculated via the Bohr-Wheeler formalism
Measurements and calculations of integral capture cross-sections of structural materials in fast reactor spectra

International Nuclear Information System (INIS)

Seth, S.; Brunson, G.; Gmuer, K.; Jermann, M.; McCombie, C.; Richmond, R.; Schmocker, U.

1979-01-01

This paper relates the checking of integral data of steel and iron in fast reactor lattices. The fully-rodded GCFR benchmark lattice of the zero-energy reactor PROTEUS was successively modified by replacing the PuO 2 -UO 2 fuel rods by steel-18/8 or steel-37 (iron) rods. The neutron spectra of the modified lattices in fact have median energies close to that of a typical LMFBR. The replacement of fuel by the structural material of interest was such that in each case the value of k(infinity) was reduced to near-unity. This allowed the measurement of the lattice-k(infinity) by the null-reactivity technique. In addition, the principal reaction rates (namely U238 capture and fission, relative to Pu239 fission) and the neutron spectrum were measured. These directly measured integral data which are particularly sensitive to the steel cross-sections can be used for the checking and systematic adjustment of data sets. The results may also be analysed so as to derive specific values for the integral capture cross-sections of steel and iron. Neutron balance equations were set-up for each of the lattices using the measured k(infinity) and reaction rates
Transmutation of fission products through accelerator

International Nuclear Information System (INIS)

Nakamura, H.; Tani, S.; Takahashi, T.; Yamamura, O.

1995-01-01

The transmutation of fission products through particle accelerators has been studied under the OMEGA program. The photonuclear reaction has also been investigated to be applied to transmuting long-lived fission products, such as Cesium and Strontium, which have difficulties on reaction with neutrons due to its so small cross section. It is applicable for the transmutation if the energy balance can be improved with a monochromatic gamma rays in the range of the Giant Dipole Resonance generated through an excellent high current electron linear accelerator. The feasibility studies are being conducted on the transmutation system using it through an electron accelerator. (authors)
COMBINE/PC - a portable neutron spectrum and cross-section generation program

International Nuclear Information System (INIS)

Nigg, D.W.; Grimesey, R.A.; Curtis, R.L.

1990-01-01

Use of personal computers and engineering workstations for complex scientific computations has expanded rapidly in the past few years. This trend is expected to continue in the future with the introduction of increasingly sophisticated microprocessors and microcomputer systems. In response to this, an integrated system of neutronics and radiation transport software suitable for operation in an IBM personal computer (PC)-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past 3 years. A key component of this system will be module to produce application-specific multigroup cross-section libraries that can be used in various neutron transport and diffusion theory code modules. This software module, referred to as COMBINE/PC, was recently completed at INEL and is the subject of this paper. COMBINE/PC was developed to provide an ENDF/B-based neutron cross-section generation capability of sufficient sophistication to handle a wide variety of practical fission and fusion-related applications while maintaining a compact machine-independent structure
Spectrum-averaged cross-section measurement of /sup 103/Rh(n,n)/sup 103m/Rh in the /sup 252/Cf fission neutron spectrum

International Nuclear Information System (INIS)

Lamaze, G.P.; Schima, F.J.; Eisenhauer, C.M.; Spiegel, V.

1988-01-01

Because of the similarity in energy dependence of the /sup 103/Rh(n,n') differential cross section to the kerma muscle response function for neutrons, rhodium may be useful as a neutron kerma monitor. In support of its use as a neutron monitor, the spectrum-averaged cross section σ-bar has been measured for a /sup 252/Cf fission neutron spectrum. Pairs of thin rhodium samples were irradiated on opposite sides of a thinly encapsulated /sup 252/Cf neutron source. The neutron emission rate of the /sup 252/Cf source was determined by the manganous sulfate (MnSO/sub 4/) bath technique. In this method, the californium source emission rate is determined by comparison to the known emission rate of NBS-I, a standard radium-beryllium neutron source. The neutron fluence incident on the rhodium samples is determined from the californium source strength, average sample-to-source distance, and the duration of the irradiation. Corrections are made for neutron scattering saturation of activity, and attenuation of the X rays by the sample during counting. The X rays were detected with an intrinsic germanium detector designed specifically for low-energy X-ray detection. The activity was not determined by absolute counting so that the final results depend on the value of P/sub Κx/, to total Κ X-ray emission probability. The results of five separate irradiations yield a value of σ-bar . P/sub Κx/ = 62.3 +- 1.9 mb. Using the most recently published value of P/sub Κx/ gives a value of σ-bar = 739 +- 22 mb. A discussion of systematic uncertainties is given
Fission threshold determination of 209Bi and sup(204,206,207,208)Pb by electrofission

International Nuclear Information System (INIS)

Tuerck, D.

1975-01-01

At the Darmstadt electron linear accelerator the cross sections for the electrofission of 209 Bi were measured for electron energies between 24 and 70 MeV, for the separated lead isotopes sup(204,206,207,208)Pb between 38 and 50 MeV. For the determination of the fission thresholds the cross sections were examined by the virtuel photon method using calculations in first Born approximation for the point nucleus with Coulomb wave functions. The analytic functions fitting the fission probability were based on level densities after the Fermi-gas-model. (orig./WL) [de
Application of Gauss quadratures to the calculation of resolved resonance contribution in multigroup cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1989-01-01

A program (RESQ) based on quadratures that evaluates, from ENDF/B data, the resolved resonance contribution in group-averaged cross sections (capture, fission and scattering) was developed. Single and Multilevel Breit-Wigner parameters are accepted. Constant weighting function and zero Kelvin were considered. To assure convergence, different quadrature orders may be analysed. Results are compared with other codes' reconstruction and integration methods. (author) [pt
Fission of highly excited nuclei investigated in complete kinematic measurements

International Nuclear Information System (INIS)

Rodriguez-Sanchez, J. L.; Benlliure, J.; Taieb, J.; Avarez-Pol, H.; Audouin, L.; Ayyad, Y.; Belier, G.; Boutoux, G.; Casarejos, E.; Chatillon, A.; Cortina-Gil, D.; Gorbinet, T.; Heinz, A.; Kelic-Heil, A.; Kurz, N.; Laurent, B.; Martin, J. F.; Paradela, C.; Pellereau, E.; Pietras, B.; Prochazka, A.; Ramos, D.; Rodriguez-Tajes, C.; Rossi, D.; Simon, H.; Tassan-Got, L.; Vargas, J.; Voss, B.

2013-01-01

Fission is an extremely complex mechanism that requires a dynamical approach to describe the evolution of the process in terms of intrinsic and collective excitations of the nuclear constituents. In order to determine these effects a complex experimental setup was mounted at GSI, which allowed us for the first time the full identification in charge and mass of all fission fragments thanks to a magnetic separation and the use of the inverse kinematic technique. Moreover, we also measured the neutron multiplicities and the light-charged particles emitted in coincidence with fission. These complete kinematic measurements will be used to define sensitive observables to dissipative and transient effects in fission. In this manuscript we present the first results for the total fission cross sections. (authors)
Recent improvements in the calculation of prompt fission neutron spectra: Preliminary results

International Nuclear Information System (INIS)

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

1989-01-01

We consider three topics in the refinement and improvement of our original calculations of prompt fission neutron spectra. These are an improved calculation of the prompt fission neutron spectrum N(E) from the spontaneous fission of 252 Cf, a complete calculation of the prompt fission neutron spectrum matrix N(E,E n ) from the neutron-induced fission of 235 U, at incident neutron energies ranging from 0 to 15 MeV, and an assessment of the scission neutron component of the prompt fission neutron spectrum. Preliminary results will be presented and compared with experimental measurements and an evaluation. A suggestion is made for new integral cross section measurements. (author). 45 refs, 12 figs, 1 tab
Transmutation analysis considering and explicit fission product treatment based on a coupled Hammer-Technion and Cinder-2 system

International Nuclear Information System (INIS)

Abe, A.Y.

1989-01-01

This work presents a study about neutron absorption in a typical PWR cell by considering an explicit treatment for the fission products. The proposed methodology to treat fission product neutron absorption in a lattice calculation combines the HAMMER-TECHNION and CINDER-2 codes. The fission product chain treatment considers nearly 99% of all original CINDER-2 neutron absorption chain treatment. Parallel to the explicit treatment, a cross section library in the HAMMER-TECHNION code multigroup structure for the fission products was generated using the ENDF/B-V fission product library and processed by NJOY and AMPX-II processing codes. The methodology validation was investigated against two available benchmarks and it was obtained excellent results for the K-Infinity (IAEA-TECDOC-233) as function of burnup and enrichment and for the aggregate quantity sup(σ)2200 in units of barns/fission cross sections (OKAZAKI and SOKOLOWSKI). This work contributed for a better understanding of the fission product neutron absorption in a typical PWR cell and showed that the explicit fission product treatment can be successfully achieved. Besides that the performance of the ENDF/B-V fission product library was accessed. (author)
User's manual for computer code RIBD-II, a fission product inventory code

International Nuclear Information System (INIS)

Marr, D.R.

1975-01-01

The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)
Experimental data on fission and (n,xn) reactions; Donnees experimentales de fission et de reactions (n,xn)

Energy Technology Data Exchange (ETDEWEB)

Belier, G.; Chatillon, A.; Granier, T.; Laborie, J.M.; Laurent, B.; Ledoux, X.; Taieb, J.; Varignon, C.; Bauge, E.; Bersillon, O.; Aupiais, J.; Le Petit, G. [CEA Bruyeres-le-Chatel, 91 (France); Authier, N.; Casoli, P. [CEA Valduc, 21 - Is-sur-Tille (France)

2011-07-15

Investigations on neutron-induced fission of actinides and the deuteron breakup are presented. Neutron-induced fission has been studied for 10 years at the WNR (Weapons Neutron Research) neutron facility of the Los Alamos Neutron Science Center (LANSCE). Thanks to this white neutron source the evolution of the prompt fission neutron energy spectra as a function of the incident neutron energy has been characterized in a single experiment up to 200 MeV incident energy. For some isotopes the prompt neutron multiplicity has been extracted. These experimental results demonstrated the effect on the mean neutron energy of the neutron emission before scission for energies higher than the neutron binding energy. This extensive program ({sup 235}U and {sup 238}U, {sup 239}Pu, {sup 237}Np and {sup 232}Th were measured) is completed by neutron spectra measurements on the CEA 4 MV accelerator. The D(n,2n) reaction is studied both theoretically and experimentally. The cross-section was calculated for several nucleon-nucleon interactions including the AV18 interaction. It has also been measured on the CEA 7 MV tandem accelerator at incident neutron energies up to 25 MeV. Uncertainties lower than 8% between 5 and 10 MeV were obtained. In particular these experiments have extended the measured domain for cross sections. (authors)
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
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.
Measurements of Fission Cross Sections for the Isotopes relevant to the Thorium Fuel Cycle

CERN Multimedia

2002-01-01

The present concern about a sustainable energy supply is characterised by a considerable uncertainty: the green house effect and foreseeable limits in fossil fuel resources on the one hand, the concern about the environmental impact of nuclear fission energy and the long term fusion research on the other hand, have led to the consideration of a variety of advanced strategies for the nuclear fuel cycle and related nuclear energy systems. The present research directories concern such strategies as the extension of the life span of presently operating reactors, the increase of the fuel burn-up, the plutonium recycling, and in particular the incineration of actinides and long-Lived fission products, the accelerator driven systems (ADS), like the "Energy Amplifier" (EA) concept of C. Rubbia, and the possible use of the Thorium fuel cycle. The detailed feasibility study and safety assessment of these strategies requires the accurate knowledge of neutron nuclear reaction data. Both, higher fuel burn-up and especiall...
NSLINK, Coupling of NJOY Cross-Sections Generator Code to SCALE-3 System

International Nuclear Information System (INIS)

De Leege, P.F.A

1991-01-01

1 - Description of program or function: NSLINK (NJOY - SCALE - LINK) is a set of computer codes to couple the NJOY cross-section generation code to the SCALE-3 code system (using AMPX-2 master library format) retaining the Nordheim resolved resonance treatment option. 2 - Method of solution: The following module and codes are included in NSLINK: XLACSR: This module is a stripped-down version of the XLACS-2 code. The module passes all l=0 resonance parameters as well as the contribution from all other resonances to the group cross-sections, the contribution from the wings of the l=0 resonances, the background cross-section and possible interference for multilevel Breit-Wigner resonance parameters. The group cross-sections are stored in the appropriate 1-D cross-section arrays. The output file has AMPX-2 master format. The original NJOY code is used to calculate all other data. The XLACSR module is included in the NJOY code. MILER: This code converts NJOY output (GENDF format) to AMPX-2 master format. The code is an extensively revised version of the original MILER code. In addition, the treatment of thermal scattering matrices at different temperatures is included. UNITABR: This code is a revised version of the UNITAB code. It merges the output of XLACSR and MILER in such a way that contributions from the bodies of the l=0 resonances in the resolved energy range, calculated by XLACSR, are subtracted from the 1-D group cross-section arrays for fission (MT=18) and neutron capture (MT=102). The l=0 resonance parameters and the contributions from the bodies of these resonances are added separately (MT=1023, 1022 and 1021). The total cross-section (MT=1), the absorption cross- section (MT=27) and the neutron removal cross-section (MT=101) values are adjusted. In the case of Bondarenko data, infinite dilution values of the cross-sections (MT=1, 18 and 102) are changed in the same way as the 1-D cross-section. The output file of UNITABR is in AMPX-2 master format and
Tables and figures from JNDC Nuclear Data Library of fission products, version 2

International Nuclear Information System (INIS)

Ihara, Hitoshi

1989-11-01

The content of JNDC (Japanese Nuclear Data Committee) FP (Fission Product) Nuclear Data Library version 2 for 1227 fission products is presented in the form of tables and figures. The library is inclusive of evaluated decay data such as decay constant, Q-value, average energies of beta, gamma and internal conversion electron, spin-parity, branching ratio of each decay mode and fission yield. The neutron capture cross-sections are also contained for 166 nuclides. The mass number of the fission product nuclides ranges from A = 66 to A = 172. (author)
Evaluation of the neutron cross sections of 235U in the thermal energy region. Final report

International Nuclear Information System (INIS)

Leonard, B.R. Jr.; Kottwitz, D.A.; Thompson, J.K.

1976-02-01

The objective of this work has been to improve the knowledge of the thermal cross sections of the fissile nuclei as a step toward providing a standard data base for the nuclear industry. The methodology uses a form of the Adler-Adler multilevel-fission theory and Breit-Wigner multilevel-scattering theory. It incorporates these theories in a general nonlinear least-squares (LSQ) fitting program SIGLEARNThe analysis methodology in this work was applied to the thermal data on 235 U. A reference data file has been developed which includes most of the known data of interest. The first important result of this work is the assessment of the shape uncertainties of the partial cross sections. The results of our studies lead to the following values and error estimates for 235 U g factors in a thermal (20.44 0 C) energy spectrum: g/sub f/ = 0.97751 (+-0.11%); g/sub γ/ = 0.98230 (+-0.14%). A second important result of this study is the development of a recommended set of 2200 m/s (0.0253 eV) values of the parameters and the probable range of further adjustment which might be made. The analysis also provides the result of a common interpretation of energy-dependent absolute cross-section data of different measurements to yield a consistent set of experimental 0.0253 eV values with rigorous error estimates. It also provides normalization factors for relative fission and capture cross sections on a common basis with rigorous error estimates. The results of these analyses provide a basis for deciding what new measurements would be most beneficial. The most important of these would be improved direct capture data in the thermal region
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.)
Cross sections for the production of 11C in C targets by 3.65 AGeV projectiles

International Nuclear Information System (INIS)

Kozma, P.; Tolstov, K.D.; Yanovskij, V.V.

1989-01-01

The absolute cross sections for the production of 11 C in C targets by 3.65 AGeV protons, deuterons, 4 He- and 12 C-ions were measured. Annihialtion radiation from 11 C was counted using a large volume NaI(Tl) and BaF 2 detectors. The flux measurement technique based on registration of charged particles by means of a thin nuclear emulsion layer rotating in a beam as well as fission chamber was used. The results are compared with earlier measurements of the cross sections in carbon targets using high-energy projectiles and Glauber theoretical prediction, as well. 10 refs.; 3 figs.; 1 tab
Elliptical cross section fuel rod study II; Estudio de barras combustibles de seccion eliptica II

Energy Technology Data Exchange (ETDEWEB)

Taboada, H; Marajofsky, A [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Combustibles Nucleares

1997-12-31

In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab.
Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)
Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)
Characterization of the fusion-fission process in light nuclear systems

International Nuclear Information System (INIS)

Anjos, R.M. dos.

1992-01-01

Fusion cross sections measurements of highly damped processes and elastic scattering were performed for the 16, 17, 18 O + 10, 11 B and 19 F + 9 Be, in the incident energy interval 22 ≤ E LAB ≤ 64 MeV. Evidences are presented that highly damped binary processes observed in these systems are originated from a fusion-fission process rather than a dinuclear ''orbiting'' mechanism. The relative importance of the fusion-fission process in these very light systems is demonstrated both by the experimental results, which indicate a statistically balanced compound nucleus fission process occurrence, and theoretical calculations. (L.C.J.A.)
Study of fast neutron scattering. The displacement cross-section (1962); Etude de la diffusion des neutrons rapides. Section efficace de deplacement (1962)

Energy Technology Data Exchange (ETDEWEB)

Millot, J P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1962-07-01

We propose a method for calculating the biological efficiency of fast neutrons emitted by in-pile fission sources. This method justifies the empirical theory of Albert and Welton. In making simple assumptions concerning the cross-sections, we have supposed that the propagation can ben reduced to a mono-kinetic problem. A system of orthonormal functions is then set up making it possible to calculate the flux leaving a planar source. This method generalises the results obtained by Platzek to the case where the elastic cross-sections are not isotropic, and make it possible in particular to define a displacement cross-section: extension of the diffusion coefficient. This method can be generalised to the case of neutron diffraction as a function of time, and to the study of slowing-down. Numerical results are given in an appendix for the following: H{sub 2}O, D{sub 2}O, Fe, Be, Pb, CH, CH{sub 2}. These cross-sections have been verified experimentally in water and in graphite for neutrons of 2.5 and 14 MeV using a SAMES accelerator and a 2 MeV Van De Graaff. (author) [French] Nous proposons une methode permettant de calculer l'efficacite biologique des neutrons rapides issus des sources de fission dans la protection d'une pile. Cette methode justifie la theorie empirique d'Albert et Welton. En faisant des hypotheses simples sur les sections efficaces, nous avons suppose que la propagation pouvait etre ramenee a un probleme monocinetique. Nous construisons alors un systeme de fonctions orthonormales qui permet de calculer le flux issu d'une source plane. Cette methode generalise les resultats obtenus par Platzek au cas ou les sections efficaces elastiques ne sont pas isotropes et en particulier permet de definir une section efficace de deplacement: extension du coefficient de diffusion. Cette methode peut etre generalisee a la diffusion des neutrons en fonction du temps et a l'etude du ralentissement. Les resultats numeriques sont donnes en annexe pour les corps. H{sub 2
Photofission cross-section ratio measurement of 235U/238U using monoenergetic photons in the energy range of 9.0-16.6 MeV

Science.gov (United States)

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

2017-05-01

Photofission cross-section ratios of 235U and 238U have been measured using monoenergetic photon beams at the HIγS facility of TUNL. These measurements have been performed in small energy steps between 9.0 and 16.6 MeV using a dual-fission ionization chamber. Measured cross-section ratios are compared with the previous experimental data as well as with the recent evaluated nuclear data library ENDF.
Fission physics experiments at the time-of-flight spectrometer GNEIS in Gatchina (PNPI)

International Nuclear Information System (INIS)

Shcherbakov, O.A.

1994-01-01

The outline of and fission physics experiments at the Gatchina neutron spectrometer GNEIS based on the 1 GeV PNPI proton synchrotron are presented. The prefission gamma-ray spectrum of the (n, gamma f) reaction were investigated. The capture gamma-ray spectra for 721.6 eV and 1211.4 eV resonances in U-238 were measured and the nature of the 721.6 eV resonance in U-238 were examined. The forward-backward asymmetry in slow neutron fission of U-235 and energy dependence of the forward-backward and instrumental asymmetry coefficients were obtained. Fission cross section ratios for Th-232 to U-235 and for U-238 to U-235 in the energy range up to 200 MeV were measured. The results of the cross section ratios agreed well with those of Behrens et al. and Difilippo et al. (T.H.)
Photon and proton induced fission on heavy nuclei at intermediate energies

Directory of Open Access Journals (Sweden)

Andrade-II E.

2014-04-01

Full Text Available We present an analysis of fission induced by intermediate energy protons or photons on actinides. The 660 MeV proton induced reactions are on 241Am, 238U, and 237Np targets and the Bremmstrahlung-photons with end-point energies at 50 MeV and 3500 MeV are on 232Th and 238U targets. The study was performed by means of the Monte Carlo simulation code CRISP. A multimodal fission extension was added to the code within an approach which accounts for the contribution of symmetric and asymmetric fission. This procedure allowed the investigation of fission cross sections, fissility, number of evaporated nucleons and fission-fragment charge distributions. The comparison with experimental data show a good agreement between calculations and experiments.
Fifty years of nuclear fission: Nuclear data and measurements series

International Nuclear Information System (INIS)

Lynn, J.E.

1989-06-01

This report is the written version of a colloquium first presented at Argonne National Laboratory in January 1989. The paper begins with an historical preamble about the events leading to the discovery of nuclear fission. This leads naturally to an account of early results and understanding of the fission phenomena. Some of the key concepts in the development of fission theory are then discussed. The main theme of this discussion is the topography of the fission barrier, in which the interplay of the liquid-drop model and nucleon shell effects lead to a wide range of fascinating phenomena encompassing metastable isomers, intermediate-structure effects in fission cross-sections, and large changes in fission product properties. It is shown how study of these changing effects and theoretical calculations of the potential energy of the deformed nucleus have led to broad qualitative understanding of the nature of the fission process. 54 refs., 35 figs
Measurement of (n,2n) and (n,3n) cross sections at incident energies between 8 and 15MeV

Energy Technology Data Exchange (ETDEWEB)

Frehaut, J; Mosinski, G

1975-01-01

Cross sections of (n,2n) and (n,3n) reactions were measured for several nuclides between 8 and 15MeV using a large liquid scintillator to count the neutrons directly. Measurements relative to the fission cross section of /sup 238/U were made for the Nd isotopes 142-144-146-148-150, for the Sm isotopes 148-150-152-154 and for the natural elements Ti, V, Cr, Cu, Zr, Mo and Pb. The relative accuracy of the measurements was generally of the order of 5 to 10%.
Nuclear level density effects on the evaluated cross-sections of nickel isotopes

International Nuclear Information System (INIS)

Garg, S.B.

1995-01-01

A detailed investigation has been made to estimate the effect of various level density options on the computed neutron induced reaction cross-sections of Ni-58 and Ni-60 covering the energy range 5-25 MeV in the framework of the multistep Hauser-Feshbach statistical model scheme which accounts for the pre-equilibrium decay according to the Kalbach exciton model and gamma-ray competition according to the giant dipole radiation model of Brink and Axel. Various level density options considered in this paper are based on the Original Gilbert-Cameron, Improved Gilbert-Cameron, Back-Shifted Fermi gas and the Ingatyuk-Smirenkin-Tishin approaches. The effect of these different level density prescriptions is brought out with special reference to (n,p) (n,2n) (n,α) and total production cross-sections for neutron, hydrogen, helium and gamma-rays which are of technological importance for fission and fusion based reactor systems. (author). 18 refs, 2 figs
Pion-induced fission of 209Bi and 119Sn: measurements, calculations, analyses and comparison

International Nuclear Information System (INIS)

Rana, M.A.; Sher, G.; Manzoor, S.; Shehzad, M.I.

2011-01-01

Cross-sections for the π - -induced fission of 209 Bi and 119 Sn have been measured using the most sensitive CR-39 solid-state nuclear track detector. In experiments, target–detector stacks were exposed to negative pions of energy 500, 672, 1068, and 1665 MeV at the Brookhaven National Laboratory, USA. An important aspect of the present paper is the comparison of pion-induced fission fragment spectra of above mentioned nuclei with the spontaneous fission fragment spectra of 252 Cf. This comparison is made in terms of fission fragment track lengths in the CR-39 detectors. Measurement results are compared with calculations of Monte Carlo and statistical weight functions methods using the computer code CEM95. Agreement between measurements and calculations is fairly good for 209 Bi target nuclei whereas it is indigent for the case of 119 Sn. The possibilities of the trustworthy calculations, using the computer code CEM95, comparable with measurements of pion-induced fission in intermediate and heavy nuclei are explored by employing various systematics available in the code. Energy dependence of pion-induced fission in 119 Sn and 209 Bi is analyzed employing a newly defined parameter geometric-size-normalized fission cross-section (χ f g ). It is found that the collective nuclear excitations, which may lead to fission, become more probable for both 209 Bi and 119 Sn nuclei with increasing energy of negative pions from 500 to 1665 MeV. (author)

Cross section of ternary fission of Al, Ti, Co and Zr nuclei induced by 0,8 - 1,8 Gev photons

International Nuclear Information System (INIS)

Lima, D.A. de; Sousa, E.V. de; Milomen, W.C.C.; Tavares, O.A.P.

1988-01-01

A research on ternary fission of Al, Ti, Co, and Zr nuclei induced by bremsstrahlung photons of 0,8, 1,0, 1,4, and 1,8 Gev end-point energies has been carried out using makrofol polycarbonate and CR-39 polymer as fission-track detectors. Results are discussed and compared with other ternary fission data. (M.W.O.) [pt
Study of a new mechanism of reaction between heavy ions: the quasi-fission

International Nuclear Information System (INIS)

Ngo, Christian.

1975-01-01

A new type of deep inelastic reaction between two heavy ions (quasi-fission) has been discovered and studied when the product Z 1 Z 2 between the two ion atomic numbers is greater than or approximately equal to 1500. This mechanism is mainly binary, the total kinetic energy of the products is the one expected for a binary fission giving the same products, most of the products have masses very close to the initial masses, the angular distribution of the light products is peaked slightly forwards the projectile grazing angle (when the bombarding energy is not too much above the interaction barrier), at last, the total cross section for this process is a large part of the total reaction cross section. These results have been interpreted on the one hand using a static model and on the other hand using a dynamic model. An interaction potential between the two heavy ions has been derived using the energy density formalism within the framework of the sudden approximation. It has been shown that the nuclear part satisfies a scaling law which allows to factorize it in one term which depends on the two ion masses and another term which is independent of the system (universal function). Using the critical distance notion, the static calculations reproduce the quasi-fission cross sections. With regards to the dynamical calculations, the previously described potential has been introduced within the framework of Deubler and Dietrich's model. It is a classical dynamical calculation including dissipative terms. The vibration degrees of freedom of each ion have been explicitely taken into account. This calculation nicely reproduces both the energy loss in the relative motion, the focusing effect of the angular distribution, and the quasi-fission cross sections [fr
Neutron-induced cross-sections via the surrogate method

International Nuclear Information System (INIS)

Boutoux, G.

2011-11-01

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

International Nuclear Information System (INIS)

Belier, G.; Chatillon, A.; Granier, T.; Laborie, J.M.; Laurent, B.; Ledoux, X.; Taieb, J.; Varignon, C.; Bauge, E.; Bersillon, O.; Aupiais, J.; Le Petit, G.; Authier, N.; Casoli, P.

2011-01-01

Investigations on neutron-induced fission of actinides and the deuteron breakup are presented. Neutron-induced fission has been studied for 10 years at the WNR (Weapons Neutron Research) neutron facility of the Los Alamos Neutron Science Center (LANSCE). Thanks to this white neutron source the evolution of the prompt fission neutron energy spectra as a function of the incident neutron energy has been characterized in a single experiment up to 200 MeV incident energy. For some isotopes the prompt neutron multiplicity has been extracted. These experimental results demonstrated the effect on the mean neutron energy of the neutron emission before scission for energies higher than the neutron binding energy. This extensive program ( 235 U and 238 U, 239 Pu, 237 Np and 232 Th were measured) is completed by neutron spectra measurements on the CEA 4 MV accelerator. The D(n,2n) reaction is studied both theoretically and experimentally. The cross-section was calculated for several nucleon-nucleon interactions including the AV18 interaction. It has also been measured on the CEA 7 MV tandem accelerator at incident neutron energies up to 25 MeV. Uncertainties lower than 8% between 5 and 10 MeV were obtained. In particular these experiments have extended the measured domain for cross sections. (authors)
Fusion-fission in Ar-heavy nuclei collisions

International Nuclear Information System (INIS)

Zaric, Alexandre

1984-01-01

Fusion-fission products have been studied for three reactions: Ar + Au, Ar + Bi and Ar + U (5.25-7.5 MeV/u). By measuring symmetric fragmentation components (fission-like events), cross sections for fusion were deduced and compared with the prediction of static and dynamic models. With increasing projectile energy, the width of the mass distributions strongly increases for the two lighter systems. By contrast, for Ar + U it remains essentially constant at a very large value. These results clearly demonstrate that the large increase of the width of the mass distribution cannot be attributed simply to large values of the angular momentum. However, they can be explained by the occurrence of a different dissipative process, fast fission, which can be expected if there is no barrier to fission. For the reaction Ar + Au, the total kinetic-energy distributions were also studied in detail. In this case fast fission occurs only at high incident energy. The average total kinetic energy (TKE) was found to be constant with increasing energy. (author) [fr
Binary and ternary fission yields induced by 12C and 20Ne ions on 238U targets

International Nuclear Information System (INIS)

Otto, R.J.

1974-01-01

Evidence for ternary fission of 250 Cf* and 258 No* compound nuclei has been found. Relative cross section data for nuclides with masses between 24 Na and 161 Tb have been determined for 12 C bombardments of natural uranium at laboratory energies of 122 MeV, 113 MeV and 105 MeV. Relative cross section data for 8 nuclides between 24 Na and 66 Ni were sought for 20 Ne bombardments of natural uranium at 150 MeV laboratory energies. The binary fission fragment mass distribution for 238 U( 12 C,f) was determined by analysis of fission fragment recoil collection foils using radiochemical techniques and high resolution gamma ray spectroscopy. The results indicated the existence of a ternary fission branch similar to mass distributions obtained for He induced fission of Th, U, and Pu nuclei at intermediate energies. Comparison of the data with He induced ternary fission data obtained previously in this laboratory indicated an increase in the ternary fission probability with increasing Z 2 /A of the compound nucleus and with excitation energy. A shift of the binary-ternary fission product intersection point to lower mass numbers with increasing Z 2 /A and excitation energy of the compound nucleus was also observed. (Diss. Abstr. Int., B)
Microscopic description of production cross sections including deexcitation effects

Science.gov (United States)

Sekizawa, Kazuyuki

2017-07-01

Background: At the forefront of the nuclear science, production of new neutron-rich isotopes is continuously pursued at accelerator laboratories all over the world. To explore the currently unknown territories in the nuclear chart far away from the stability, reliable theoretical predictions are inevitable. Purpose: To provide a reliable prediction of production cross sections taking into account secondary deexcitation processes, both particle evaporation and fission, a new method called TDHF+GEMINI is proposed, which combines the microscopic time-dependent Hartree-Fock (TDHF) theory with a sophisticated statistical compound-nucleus deexcitation model, GEMINI++. Methods: Low-energy heavy ion reactions are described based on three-dimensional Skyrme-TDHF calculations. Using the particle-number projection method, production probabilities, total angular momenta, and excitation energies of primary reaction products are extracted from the TDHF wave function after collision. Production cross sections for secondary reaction products are evaluated employing GEMINI++. Results are compared with available experimental data and widely used grazing calculations. Results: The method is applied to describe cross sections for multinucleon transfer processes in 40Ca+124Sn (Ec .m .≃128.54 MeV ), 48Ca+124Sn (Ec .m .≃125.44 MeV ), 40Ca+208Pb (Ec .m .≃208.84 MeV ), 58Ni+208Pb (Ec .m .≃256.79 MeV ), 64Ni+238U (Ec .m .≃307.35 MeV ), and 136Xe+198Pt (Ec .m .≃644.98 MeV ) reactions at energies close to the Coulomb barrier. It is shown that the inclusion of secondary deexcitation processes, which are dominated by neutron evaporation in the present systems, substantially improves agreement with the experimental data. The magnitude of the evaporation effects is very similar to the one observed in grazing calculations. TDHF+GEMINI provides better description of the absolute value of the cross sections for channels involving transfer of more than one proton, compared to the grazing
Development of modern CANDU PHWR cross-section libraries for SCALE

International Nuclear Information System (INIS)

Shoman, Nathan T.; Skutnik, Steven E.

2016-01-01

Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of "1"5"4Eu and "1"5"5Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the "1"5"4Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a typical
A possible mechanism in heavy ion induced reactions: 'fast fission process'

International Nuclear Information System (INIS)

Borderie, B.; Gardes, D.; Berlanger, M.

1980-01-01

The influence of the orbital angular momentum l on the mass distribution of fission fragments is studied, both on previously available data on heavy ion induced fission and in new specifically planned experiments: systems 40 Ar + 165 Ho and 24 Mg + 181 Ta at bombarding energies ranging from 180 up to 391 MeV and leading to the same fissionning nucleus 205 At wigh different l distributions. When l values corresponding to a vanished fission barrier are reached, the mass distribution broadens. This suggest the existence of a specific process, 'fast fission', at l-values leading to compound nucleus formation and deep inelastic collisions, respectively. This process and its conditions of occurrence are discussed; of special interest are the correlated differences between the limitations to the fission cross-section and the fission mass distributions broadenings, respectively, for the Ar + Ho and Mg + Ta systems
Dynamic effects in neutron induced fission of 230Th and 232Th

International Nuclear Information System (INIS)

Trochon, J.; Frehaut, J.; Pranal, Y.; Simon, G.; Boldeman, J.W.

1982-09-01

The fission fragment characteristics of the two thorium isotopes 230 Th and 232 Th have been measured in an attempt to study the evolution of the fissioning nucleus from saddle point to scission. The partial fission channel at the saddle point have been deduced from a fission fragment angular distribution and fission cross section analysis. Changes with energy in the average number of prompt neutron (νsub(p)) emitted per fission and the total fragment kinetic energy (TKE) have been observed in the fission threshold region. A rather good fit of νsub(p) and TKE values has been obtained on the basis of a correlation of these quantities and the partial fission channel ratios. This leads to expect for these isotopes a passage from saddle point to scission sufficiently rapid for the coupling between collective and intrinsic excitation to be very weak [fr
Evaluation of ETOG-3Q, ETOG-3, FLANGE-II, XLACS, NJOY and LINEAR/RECENT/GROUPIE computer codes concerning to the resonance contribution and background cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1988-12-01

The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analysed. (author) [pt
Measurements of the {sup 235}U(n,f) cross section in the 3 to 30 MeV neutron energy region

Energy Technology Data Exchange (ETDEWEB)

Carlson, A.D.; Wasson, O.A. [National Institute of Standards and Technology, Gaithersburg, MD (United States); Lisowski, P.W. [Los Alamos National Lab., NM (United States)] [and others

1991-12-31

To improve the accuracy of the {sup 235}U(n,f) cross section, measurements have been made of this standard cross section at the target 4 facility at Los Alamos National Laboratory (LANL). The data were obtained at the 20-meter flight path of that facility. The fission reaction rate was determined with a fast parallel plate ionization chamber and the neutron fluence was measured with an annular proton recoil telescope. The measurements provide the shape of the {sup 235}U(n,f) cross section relative to the hydrogen scattering cross section for neutron energies from about 3 to 30 MeV neutron energy. The data have been normalized to the very accurately known value near 14 MeV. The results are in good agreement with the ENDF/B-VI evaluation up to about 15 MeV neutron energy. Above this energy differences as large as 5% are observed.
Spectrum integrated (n,He) cross section comparisons and least squares analyses for 6Li and 10B in benchmark fields

International Nuclear Information System (INIS)

Schenter, R.E.; Oliver, B.M.; Farrar, H. IV.

1986-06-01

Spectrum integrated cross sections for 6 Li and 10 B from five benchmark fast reactor neutron fields are compared with calculated values obtained using the ENDF/B-V Cross Section Files. The benchmark fields include the Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory, the 10% Enriched U-235 Critical Assembly (BIG-10) at Los Alamos National Laboratory, the Sigma-Sigma and Fission Cavity fields of the BR-1 reactor at CEN/SCK, and the Intermediate Energy Standard Neutron Field (ISNF) at the National Bureau of Standards. Results from least square analyses using the FERRET computer code to obtain adjusted cross section values and their uncertainties are presented. Input to these calculations include the above five benchmark data sets. These analyses indicate a need for revision in the ENDF/B-V files for the 10 B and 6 Li cross sections for energies above 50 keV
Contribution to uncertainties evaluation for fast reactors neutronic cross sections

International Nuclear Information System (INIS)

Privas, Edwin

2015-01-01

The thesis has been motivated by a wish to increase the uncertainty knowledge on nuclear data, for safety criteria. It aims the cross sections required by core calculation for sodium fast reactors (SFR), and new tools to evaluate its.The main objective of this work is to provide new tools in order to create coherent evaluated files, with reliable and mastered uncertainties. To answer those problematic, several methods have been implemented within the CONRAD code, which is developed at CEA of Cadarache. After a summary of all the elements required to understand the evaluation world, stochastic methods are presented in order to solve the Bayesian inference. They give the evaluator more information about probability density and they also can be used as validation tools. The algorithms have been successfully tested, despite long calculation time. Then, microscopic constraints have been implemented in CONRAD. They are defined as new information that should be taken into account during the evaluation process. An algorithm has been developed in order to solve, for example, continuity issues between two energy domains, with the Lagrange multiplier formalism. Another method is given by using a marginalization procedure, in order to either complete an existing evaluation with new covariance or add systematic uncertainty on an experiment described by two theories. The algorithms are well performed along examples, such the 238 U total cross section. The last parts focus on the integral data feedback, using methods of integral data assimilation to reduce the uncertainties on cross sections. This work ends with uncertainty reduction on key nuclear reactions, such the capture and fission cross sections of 238 U and 239 Pu, thanks to PROFIL and PROFIL-2 experiments in Phenix and the Jezebel benchmark. (author) [fr
Fission Fragment Mass Distributions and Total Kinetic Energy Release of 235-Uranium and 238-Uranium in Neutron-Induced Fission at Intermediate and Fast Neutron Energies

Energy Technology Data Exchange (ETDEWEB)

Duke, Dana Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-12

This Ph.D. dissertation describes a measurement of the change in mass distributions and average total kinetic energy (TKE) release with increasing incident neutron energy for fission of ²³⁵U and ²³⁸U. Although fission was discovered over seventy-five years ago, open questions remain about the physics of the fission process. The energy of the incident neutron, En, changes the division of energy release in the resulting fission fragments, however, the details of energy partitioning remain ambiguous because the nucleus is a many-body quantum system. Creating a full theoretical model is difficult and experimental data to validate existing models are lacking. Additional fission measurements will lead to higher-quality models of the fission process, therefore improving applications such as the development of next-generation nuclear reactors and defense. This work also paves the way for precision experiments such as the Time Projection Chamber (TPC) for fission cross section measurements and the Spectrometer for Ion Determination in Fission (SPIDER) for precision mass yields.
MICROX-2 cross section library based on ENDF/B-VII

International Nuclear Information System (INIS)

Hou, J.; Ivanov, K.; Choi, H.

2012-01-01

New cross section libraries of a neutron transport code MICROX-2 have been generated for advanced reactor design and fuel cycle analyses. A total of 386 nuclides were processed, including 10 thermal scattering nuclides, which are available in ENDF/B-VII release 0 nuclear data. The NJOY system and MICROR code were used to process nuclear data and convert them into MICROX-2 format. The energy group structure of the new library was optimized for both the thermal and fast neutron spectrum reactors based on Contributon and Point-wise Cross Section Driven (CPXSD) method, resulting in a total of 1173 energy groups. A series of lattice cell level benchmark calculations have been performed against both experimental measurements and Monte Carlo calculations for the effective/infinite multiplication factor and reaction rate ratios. The results of MICROX-2 calculation with the new library were consistent with those of 15 reference cases. The average errors of the infinite multiplication factor and reaction rate ratio were 0.31% δk and 1.9%, respectively. The maximum error of reaction rate ratio was 8% for 238 U-to- 235 U fission of ZEBRA lattice against the reference calculation done by MCNP5. (authors)
Methodological study of the adjustment of multigroup cross sections for fast reactors, by means of integral data

International Nuclear Information System (INIS)

Bastos, H.F.B.N.

1979-01-01

In this work a study of the methodology of the adjustment of multigroup cross sections by means of integral data is presented. A synthesis of the principal methods existent and the mathematical development of the adaptation of one of them are made. A calculational system is built from this reference method, with the basic conditions for the operation of the process of adjustment. In order to test the system developed and analyze several problems related to the adjustment, a series of trial adjustments was made with the value of the U 235 fission cross section from the infinite dilution library used in the calculational system for fast reactors of the Instituto de Engenharia Nuclear. (author)
Photon and proton induced fission on heavy nuclei at intermediate energies

Energy Technology Data Exchange (ETDEWEB)

Andrade-II, E.; Karapetyan, G.S.; Deppman, A.; Guimaraes, V. [Universidade de Sao Paulo (USP), Sao Paulo, SP (Brazil). Instituto de Fisica; Balabekyan, A.R. [Yerevan State University, Alex Manoogian 1, Yerevan (Armenia); Demekhina, N.A. [Yerevan Physics Institute, Alikhanyan Brothers 2, Yerevan (Armenia); Joint Institute for Nuclear Research (JINR), Flerov Laboratory of Nuclear Reactions (LNR), Moscow (Russian Federation)

2014-07-01

We present an analysis of fission induced by intermediate energy protons or photons on actinides. The 660 MeV proton induced reactions are on {sup 241}Am, {sup 238}U, and {sup 237}Np targets and the Bremsstrahlung-photons with end-point energies at 50 MeV and 3500 MeV are on {sup 232}Th and {sup 238}U targets. The study was performed by means of the Monte Carlo simulation code CRISP. A multimodal fission extension was added to the code within an approach which accounts for the contribution of symmetric and asymmetric fission. This procedure allowed the investigation of fission cross sections, fissility, number of evaporated nucleons and fission-fragment charge distributions. The comparison with experimental data show a good agreement between calculations and experiments. (author)
Optimal systematics of single-humped fission barriers for statistical calculations

International Nuclear Information System (INIS)

Mashnik, S.G.

1993-01-01

A systematic comparison of the existing phenomenological approaches and models for describing single-humped fast-computing fission barriers are given. The experimental data on excitation energy dependence of the fissility of compound nuclei are analyzed in the framework of the statistical approach by using different models for fission barriers, shell and pairing corrections and level-density parameter in order to identify their reliability and region of applicability for Monte Carlo calculations of evaporative cascades. The energy dependence of fission cross-sections for reactions induced by intermediate energy protons has been analyzed in the framework of the cascade-exiton model. 53 refs., 15 figs., 3 tabs
Fission product model for lattice calculation of high conversion boiling water reactor

International Nuclear Information System (INIS)

Iijima, S.; Yoshida, T.; Yamamoto, T.

1988-01-01

A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Energy Technology Data Exchange (ETDEWEB)

Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

2011-12-01

The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also
Mass distribution of fission-like fragments formed in 20Ne + 165Ho system at Elab≈ 8.2 MeV/A

International Nuclear Information System (INIS)

Singh, D.; Linda, Sneha Bharti; Giri, Pankaj K.

2017-01-01

In the present work, an attempt has been made to study CFF and IFF in 20 Ne + 165 Ho system at projectile energy ≈ 8.2 MeV/A. Twelve fission like fragments (FLF) produced through complete fusion-fission (CFF) and/or incomplete fusion-fission (IFF) in the present system have been identified. The production cross-sections of identified fission like fragments have been measured and the mass distribution of fission like fragments studied
Study of U{sup 235} neutron fission spectrum by the knowledge of cross sections average over that spectrum; Estudio del espectro de neutrones de fision del {sup 235}U a traves del conocimiento de secciones eficaces promediadas sobre dicho espectro

Energy Technology Data Exchange (ETDEWEB)

Suarez, P M [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

1998-12-31

A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the {sup 45}Sc(n,2n) {sup 44g}Sc and {sup 45}Sc(n,2n) {sup 44m}Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, {sup 235}U. The results of the calculations mentioned above lead us to propose an analytical form for the {sup 235}U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole
238U and 237Np nuclear fission by 90-270 MeV electrons

International Nuclear Information System (INIS)

Kuznetsov, V.L.; Nedorezov, V.G.; Nikitina, N.V.; Noga, V.I.; Ranyuk, Yu.N.; Telegin, Yu.N.; Smirnov, A.N.; Ehjsmont, V.P.

1981-01-01

A technique for measuring cross sections of 238 U and 237 Np nuclei fission caused by 90-270 MeV electrons is described. Measurement results are given. The results obtained are discussed on the basis of the virtual photon method. It is shown that the difference in cross sections of 238 U and 237 Np electrofission is due to the different contribution of the giant resonance [ru
The correction of pebble bed reactor nodal cross sections for the effects of leakage and depletion history

Science.gov (United States)

Hudson, Nathanael Harrison

An accurate and computationally fast method to generate nodal cross sections for the Pebble Bed Reactor (PBR) was presented. In this method, named Spectral History Correction (SHC), a set of fine group microscopic cross section libraries, pre-computed at specified depletion and moderation states, was coupled with the nodal nuclide densities and group bucklings to compute the new fine group spectrum for each node. The relevant fine group cross-section library was then recollapsed to the local broad group cross-section structure with this new fine group spectrum. This library set was tracked in terms of fuel isotopic densities. Fine group modulation factors (to correct the homogeneous flux for heterogeneous effects) and fission spectra were also stored with the cross section library. As the PBR simulation converges to a steady state fuel cycle, the initial nodal cross section library becomes inaccurate due to the burnup of the fuel and the neutron leakage into and out of the node. Because of the recirculation of discharged fuel pebbles with fresh fuel pebbles, a node can consist of a collection of pebbles at various burnup stages. To account for the nodal burnup, the microscopic cross sections were combined with nodal averaged atom densities to approximate the fine group macroscopic cross-sections for that node. These constructed, homogeneous macroscopic cross sections within the node were used to calculate a numerical solution for the fine group spectrum with B1 theory. This new fine spectrum was used to collapse the pre-computed microscopic cross section library to the broad group structure employed by the fuel cycle code. This SHC technique was developed and practically implemented as a subroutine within the PBR fuel cycle code PEBBED. The SHC subroutine was called to recalculate the broad group cross sections during the code convergence. The result was a fast method that compared favorably to the benchmark scheme of cross section calculation with the lattice
Contribution to the study of the unresolved resonance range of the neutrons cross sections

International Nuclear Information System (INIS)

Noguere, Gilles

2014-01-01

This document presents the statistical description of neutron cross sections in the unresolved resonance range. The modeling of the total cross section and of the 'shape - elastic' cross section is based on the 'average R-Matrix' formalism. The partial cross sections describing the radiative capture, elastic scattering, inelastic scattering and fission process are calculated using the Hauser-Feshbach formalism with width fluctuation corrections. In the unresolved resonance range, these models depend on the average resonance parameters (neutron strength function Sc, mean level spacing D c , average partial reaction widths Γ c , channel radius a c , effective radius R' and distant level parameter R-bar c ∞ ). The codes (NJOY, CALENDF...) dedicated to the processing of nuclear data libraries (JEFF, ENDF/B, JENDL, CENDL, BROND... ) use the average parameters to take into account the self-shielding phenomenon for the simulation of the neutron transport in Monte-Carlo (MCNP, TRIPOLI... ) and deterministic (APOLLO, ERANOS...) codes. The evaluation work consists in establishing a consistent set of average parameters as a function of the total angular momentum J of the system and of the orbital moment of the incident neutron l. The work presented in this paper aims to describe the links between the S-Matrix and the 'average R-Matrix' formalism for the calculation of Sc, R-bar c ∞ , ac and R'. (author) [fr
Cross section for the 103Rh(n,n')103Rhm reaction in the energy range 5.7 endash 12 MeV

International Nuclear Information System (INIS)

Miah, M.M.; Strohmaier, B.; Vonach, H.; Mannhart, W.; Schmidt, D.

1996-01-01

The 103 Rh(n,n ' ) 103 Rh m cross section was measured by the activation method in the neutron energy range 5.7 endash 12 MeV with an uncertainty of ≅5%. Monoenergetic neutrons produced by the D(d,n) 3 He reaction were used to irradiate metallic Rh samples at 0 degree relative to the deuteron beam. The K x rays from 103 Rh m were measured with a calibrated Si detector, and the neutron fluence was determined by means of a 238 U fission chamber. The measured cross sections resolve the discrepancies in previous data and agree with the results of recent statistical model calculations of the fast-neutron cross sections of rhodium. copyright 1996 The American Physical Society
Cross sections for atmospheric corrections

International Nuclear Information System (INIS)

Meyer, J.P.; Casse, M.; Westergaard, N.

1975-01-01

A set of cross sections for spallation of relativistic nuclei is proposed based on (i) the best available proton cross sections, (ii) an extrapolation to heavier nuclei of the dependence on the number of nucleons lost of the 'target factor' observed for C 12 and O 16 by Lindstrom et al. (1975), in analogy with Rudstam's formalism, and (iii) on a normalization of all cross sections to the total cross sections for production of fragments with Asub(f) >= 6. The obtained cross sections for peripheral interactions are not inconsistent with simple geometrical considerations. (orig.) [de
Systematic evaluation of prompt neutron spectra in fission

International Nuclear Information System (INIS)

Osawa, Takaaki

1995-01-01

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

Energy Technology Data Exchange (ETDEWEB)

Hecht, Adam [Univ. of New Mexico, Albuquerque, NM (United States)

2017-11-21

Within the 3 year POP we propose to continue to test and further develop the fission spectrometers, to do development tests and full data acquisition run at the national laboratory neutron beam facilities, to measure correlated fission fragment yields at low neutron energies with 235 U fission targets, and make these data available to the nuclear community. The spectrometer development will be both on the university based r\\prototype and on the National Laboratory Spectrometer, and measurements will be performed with both. Over the longer time frame of the collaboration, we will take data over a range of low energies, and use other fission targets available to the laboratory. We will gather energy specific fragment distributions and reaction cross sections. We will further develop the data acquisition capabilities to take correlated fission fragment'gamma ray/neurton data, all on an event-by-event basis. This really is an enabling technology.
Neutron emission in fission of highly excited californium nuclei (E* = 76 MeV)

International Nuclear Information System (INIS)

Blinov, M.V.; Bordyug, V.M.; Kozulin, E.M.; Levitovich, M.; Mozhaev, A.N.; Muzychka, Yu.A.; Penionzhkevich, Yu.E.; Pustyl'nik, B.I.

1990-01-01

Differential cross sections for neutron production have been measured in fission of excited californium nuclei produced in the reaction 238 U + 12 C (105 MeV). It follows from analysis of the experimental results that the number of neutrons emitted before fission considerably exceeds the number obtained in the framework of the standard statistical model. On the basis of the multiplicity of neutrons they authors have estimated the time of fission of the excited nucleus. The dependence of the number of neutrons and their average energies on the mass of the fragments is determined
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)
Measurements and analysis of the 127I and 129I neutron capture and total cross sections

International Nuclear Information System (INIS)

Noguere, G.

2005-01-01

Most of the experimental work on the interaction of neutrons with matter has focused on materials important to reactor physics and reactor structures. By comparison, the corresponding data for minor actinides or long-lived fission products are poor. A significant demand has developed for improved neutron cross-section data of these little-studied nuclides due to the surge of interest in the transmutation of nuclear waste. With 400 kg of 129 I produced yearly in the reactors of the EU countries and a very long β - half-life of 1.57 x 10 7 years, iodine requires disposal strategies that will isolate this isotope from the environment for long periods of time. Therefore, 129 I is potentially a key long-lived fission product for transmutation applications, since 129 I transmutes in 130 I after a single neutron capture and decays to 130 Xe with a 12.36 h half-life. Accurate capture cross sections would help to reduce uncertainties in waste management concepts. For that purpose, Time-Of-Flight measurements covering the [0.5 eV-100 keV] energy range have been carried out at the 150 MeV pulsed neutron source GELINA of the Institute for Reference Materials and Measurements (IRMM). Two types of experiments have been performed at the IRMM, namely capture and transmission experiments. They are respectively related to the neutron capture and total cross sections. Since the PbI 2 samples used in this work contain natural and radioactive iodine, extensive measurements of 129 I have been carried out under the same experimental conditions as for the 129 I. The data reduction process was performed with the AGS system, and the resonance parameters were extracted with the SAMMY and REFIT shape analysis codes. In a last step, the parameters have been converted into ENDF-6 format and processed with the NJOY code to produce point-wise and multigroup cross sections, as well as MCNP and ERANOS libraries. (author)
A fast-fission component with small mass drift in the reaction 84Kr + 27Al at ELab = 5.9 MeV/u

International Nuclear Information System (INIS)

Heusch, B.; Freiesleben, H.; Schneider, W.F.W.; Kohlmeyer, B.; Stege, H.; Puehlhofer, F.

1985-01-01

All reaction products in the range from target- and projectile-like to fission-like fragments were measured for the system 84 Kr + 27 Al at 5.9 MeV/u beam energy. They are assigned to the various reaction mechanisms on the basis of experimental signatures (energy dissipation, mass and angular distribution). The sum of the measured partial cross sections, including the evaporation residue yield obtained previously, agrees with the total reaction cross section derived from elastic scattering. A small fast-fission component was found, discernible from deep-inelastic reactions by its 1/sinθ angular distribution, and distinguished from compound-nucleus fission by an incomplete mass asymmetry relaxation
Development of Indian cross section data files for Th-232 and U-233 and integral validation studies

International Nuclear Information System (INIS)

Ganesan, S.

1988-01-01

This paper presents an overview of the tasks performed towards the development of Indian cross section data files for Th-232 and U-233. Discrepancies in various neutron induced reaction cross sections in various available evaluated data files have been obtained by processing the basic data into multigroup form and intercomparison of the latter. Interesting results of integral validation studies for capture, fission and (n,2n) cross sections for Th-232 by analyses of selected integral measurements are presented. In the resonance range, energy regions where significant differences in the calculated self-shielding factors for Th-232 occur have been identified by a comparison of self-shielded multigroup cross sections derived from two recent evaluated data files, viz., ENDF/B-V (Rev.2) and JENDL-2, for several dilutions and temperatures. For U-233, the three different basic data files ENDF/B-IV, JENDL-2 and ENDL-84 were intercompared. Interesting observations on the predictional capability of these files for the criticality of the spherical metal U-233 system are given. The current status of Indian data file is presented. (author) 62 ref
Evaluation of cross-section uncertainties using physical constraints for 238U, 239Pu

International Nuclear Information System (INIS)

De Saint Jean, Cyrille; Privas, Edwin; Archier, Pascal; Noguere, Gilles; Litaize, Olivier; Leconte, Pierre; Bernard, David

2014-01-01

Neutron-induced reactions between 0 eV and 20 MeV are based on various physical properties such as nuclear reaction models, microscopic and integral measurements. Most of the time, the evaluation work is done independently between the resolved resonance range and the continuum, giving rise to mismatches for the cross-sections, larger uncertainties on boundary and no cross-correlation between high-energy domain and resonance range. In addition the use of integral experiment is sometimes only related to central values (evaluation is 'working fine' on a dedicated set of benchmarks) and reductions of uncertainties are not straightforward on cross-sections themselves: working fine could be mathematically reflected by a reduced uncertainty. As the CIELO initiative is to bring experts in each field to propose/discuss these matters, after having presented the status of 238 U and 239 Pu cross-sections covariances evaluation (for JEFF-3.2 as well as the WPEC SG34 subgroup), this paper will present several methodologies that may be used to avoid such effects on covariances. A first idea based on the use of experiments overlapping two energy domains appeared in the near past. It was reviewed and extended to the use of systematic uncertainties (normalisation for example) and for integral experiments as well. In addition, we propose a methodology taking into account physical constraints on an overlapping energy domain where both nuclear reaction models are used (continuity of both cross-sections and derivatives for example). The use of Lagrange multiplier (related to these constraints) in a classical generalised least square procedure will be exposed. Some academic examples will then be presented for both point-wise and multi-group cross-sections to present the methodologies. In addition, new results for 239 Pu will be presented on resonance range and higher energies to reduce capture and fission cross-section uncertainties by using integral experiments (JEZEBEL experiment as
Neutron total cross section measurements in the energy region from 47 keV to 20 MeV

International Nuclear Information System (INIS)

Poenitz, W.P.; Whalen, J.F.

1983-05-01

Neutron total cross sections were measured for 26 elements. Data were obtained in the energy range from 47 keV to 20 MeV for 11 elements in the range of light-mass fission products. Previously reported measurements for eight heavy and actinide isotopes were extended to 20 MeV. Data were also obtained for Cu (47 keV to 1.4 MeV) and for Sc, Zn, Nd, Hf, and Pt (1.8 to 20 MeV). The present work is part of a continuing effort to provide accurate neutron total cross sections for evaluations and for optical-model parameteriztions. The latter are required for the derivation of other nuclear-data information of importance to applied programs. 37 references
Photofission cross-section ratio measurement of {sup 235}U/{sup 238}U using monoenergetic photons in the energy range of 9.0–16.6 MeV

Energy Technology Data Exchange (ETDEWEB)

Krishichayan, E-mail: krishi@tunl.duke.edu [Department of Physics, Duke University, Durham, NC 27708 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States); Bhike, Megha; Finch, S.W.; Howell, C.R. [Department of Physics, Duke University, Durham, NC 27708 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States); Tonchev, A.P. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Tornow, W. [Department of Physics, Duke University, Durham, NC 27708 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States)

2017-05-11

Photofission cross-section ratios of {sup 235}U and {sup 238}U have been measured using monoenergetic photon beams at the HIγS facility of TUNL. These measurements have been performed in small energy steps between 9.0 and 16.6 MeV using a dual-fission ionization chamber. Measured cross-section ratios are compared with the previous experimental data as well as with the recent evaluated nuclear data library ENDF.
Approach synthesis of superheavy nuclei from some aspects of cross section calculations

International Nuclear Information System (INIS)

Liu Zuhua

2003-01-01

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

Energy Technology Data Exchange (ETDEWEB)

Lee, Amy Sarah [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-05-16

Neutron reaction cross sections play a vital role in tracking the production and destruction of isotopes exposed to neutron fluence. They are central to the process of reconciling the initial and final atom inventories. Measurements of irradiated samples by radiochemical methods in tangent with an algorithm are used to evaluate the fluence a sample is exposed to over the course of the irradiation. This algorithm is the Isotope Production Code (IPC) created and used by the radiochemistry data assessment team at Los Alamos National Laboratory (LANL). An integral result is calculated by varying the total neutron fluence seen by a sample. A sample, irradiated in a critical assembly, will be exposed to a unique neutron flux defined by the neutron source and distance of the sample from the source. Neutron cross sections utilized are a function of the hardness of the neutron spectrum at the location of irradiation. A spectral index is used an indicator of the hardness of the neutron spectrum. Cross sections fit forms applied in IPC are collapsed from a LANL 30-group energy structure. Several decades of research and development have been performed to formalize the current IPC cross section library. Basis of the current fission spectrum neutron reaction cross section library is rooted in critical assembly experiments performed from the 1950’s through the early 1970’s at LANL. The focus of this report is development of the spectral index used an indicator of the hardness of the neutron spectrum in the fission spectrum energy regime.

Multitrajectory eikonal cross sections

International Nuclear Information System (INIS)

Turner, R.E.

1983-01-01

With the use of reference and distorted transition operators, a time-correlation-function representation of the inelastic differential cross section has recently been used to obtain distorted eikonal cross sections. These cross sections involve straight-line and reference classical translational trajectories that are unaffected by any internal-state changes which have occurred during the collision. This distorted eikonal theory is now extended to include effects of internal-state changes on the translational motion. In particular, a different classical trajectory is associated with each pair of internal states. Expressions for these inelastic cross sections are obtained in terms of time-ordered cosine and sine memory functions using the Zwanzig-Feshbach projection-operator method. Explicit formulas are obtained in the time-disordered perturbation approximation
A set-up for measuring neutron cross sections and radiation multiplicity from neutron-nucleus interaction

International Nuclear Information System (INIS)

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

1988-01-01

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

Data.gov (United States)

Minnesota Department of Natural Resources — FEMA Cross Sections are required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally...
Spectrum integrated (n,He) cross section comparison and least squares analysis for /sup 6/Li and /sup 10/B in benchmark fields

International Nuclear Information System (INIS)

Schenter, R.E.; Oliver, B.M.; Farrar, H. IV

1987-01-01

Spectrum integrated cross sections for /sup 6/Li and /sup 10/B from five benchmark fast reactor neutron fields are compared with calculated values obtained using the ENDF/B-V Cross Section Files. The benchmark fields include the Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory, the 10% Enriched U-235 Critical Assembly (BIG-10) at Los Alamos National Laboratory, the Sigma Sigma and Fission Cavity fields of the BR-1 reactor at CEN/SCK, and the Intermediate-Energy Standard Neutron Field (ISNF) at the National Bureau of Standards. Results from least square analyses using the FERRET computer code to obtain adjusted cross section values and their uncertainties are presented. Input to these calculations include the above five benchmark data sets. These analyses indicate a need for revision in the ENDF/B-V files for the /sup 10/B cross section for energies above 50 keV
Review of the neutron capture process in fission reactors

International Nuclear Information System (INIS)

Poenitz, W.P.

1981-07-01

The importance of the neutron capture process and the status of the more important cross section data are reviewed. The capture in fertile and fissile nuclei is considered. For thermal reactors the thermal to epithermal capture ratio for 238 U and 232 Th remains a problem though some improvements were made with more recent measurements. The capture cross section of 238 U in the fast energy range remains quite uncertain and a long standing discrepancy for the calculated versus experimental central reaction rate ratio C28/F49 persists. Capture in structural materials, fission product nuclei and the higher actinides is also considered
A Cross-Section Adjustment Method for Double Heterogeneity Problem in VHTGR Analysis

International Nuclear Information System (INIS)

Yun, Sung Hwan; Cho, Nam Zin

2011-01-01

Very High Temperature Gas-Cooled Reactors (VHTGRs) draw strong interest as candidates for a Gen-IV reactor concept, in which TRISO (tristructuralisotropic) fuel is employed to enhance the fuel performance. However, randomly dispersed TRISO fuel particles in a graphite matrix induce the so-called double heterogeneity problem. For design and analysis of such reactors with the double heterogeneity problem, the Monte Carlo method is widely used due to its complex geometry and continuous-energy capabilities. However, its huge computational burden, even in the modern high computing power, is still problematic to perform wholecore analysis in reactor design procedure. To address the double heterogeneity problem using conventional lattice codes, the RPT (Reactivityequivalent Physical Transformation) method considers a homogenized fuel region that is geometrically transformed to provide equivalent self-shielding effect. Another method is the coupled Monte Carlo/Collision Probability method, in which the absorption and nu-fission resonance cross-section libraries in the deterministic CPM3 lattice code are modified group-wise by the double heterogeneity factors determined by Monte Carlo results. In this paper, a new two-step Monte Carlo homogenization method is described as an alternative to those methods above. In the new method, a single cross-section adjustment factor is introduced to provide self-shielding effect equivalent to the self-shielding in heterogeneous geometry for a unit cell of compact fuel. Then, the homogenized fuel compact material with the equivalent cross-section adjustment factor is used in continuous-energy Monte Carlo calculation for various types of fuel blocks (or assemblies). The procedure of cross-section adjustment is implemented in the MCNP5 code
Statistical Modelling of Resonant Cross Section Structure in URR, Model of the Characteristic Function

International Nuclear Information System (INIS)

Koyumdjieva, N.

2006-01-01

A statistical model for the resonant cross section structure in the Unresolved Resonance Region has been developed in the framework of the R-matrix formalism in Reich Moore approach with effective accounting of the resonance parameters fluctuations. The model uses only the average resonance parameters and can be effectively applied for analyses of cross sections functional, averaged over many resonances. Those are cross section moments, transmission and self-indication functions measured through thick sample. In this statistical model the resonant cross sections structure is accepted to be periodic and the R-matrix is a function of ε=E/D with period 0≤ε≤N; R nc (ε)=π/2√(S n *S c )1/NΣ(i=1,N)(β in *β ic *ctg[π(ε i - = ε-iS i )/N]; Here S n ,S c ,S i is respectively neutron strength function, strength function for fission or inelastic channel and strength function for radiative capture, N is the number of resonances (ε i ,β i ) that obey the statistic of Porter-Thomas and Wigner's one. The simple case of this statistical model concerns the resonant cross section structure for non-fissile nuclei under the threshold for inelastic scattering - the model of the characteristic function with HARFOR program. In the above model some improvements of calculation of the phases and logarithmic derivatives of neutron channels have been done. In the parameterization we use the free parameter R l ∞ , which accounts the influence of long-distant resonances. The above scheme for statistical modelling of the resonant cross section structure has been applied for evaluation of experimental data for total, capture and inelastic cross sections for 232 Th in the URR (4-150) keV and also the transmission and self-indication functions in (4-175) keV. The set of evaluated average resonance parameters have been obtained. The evaluated average resonance parameters in the URR are consistent with those in the Resolved Resonance Region (CRP for Th-U cycle, Vienna, 2006
Integral measurement of fission products capture in fast breeder reactors

International Nuclear Information System (INIS)

Martin Deidier, Loick.

1979-12-01

For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr
Fission and nuclear fragmentation of silver and bromine nuclei by photons of 1-6 GeV

International Nuclear Information System (INIS)

Pinheiro Filho, J.D.

1983-01-01

The studies of fission and fragmentation of silver and bromine nuclei by Bremsstrahlung photons of 1.6 GeV energy range are presented. The Il ford-KO nuclear emulsion submitted to Bremsstrahlung beams in Deutsches Elektronen Synchrotron (DESY) with total doses of 10'' equivalent photons, was used for nuclear fragment detection. The discrimination of fission and fragmentation events was done analysing angular distribution, range and angles between fragments. The results of fragment range distributions, angular distributions, distributions of angles between fragments, distributions of ratio between range, velocity distributions forward/backward ratio, cross sections of fission and fragmentation, nuclear fissionability and ternary fission frequency are presented and discussed. (M.C.K.)
Measurement of Plutonium-240 Angular Momentum Dependent Fission Probabilities Using the Alpha-Alpha' Reaction

Science.gov (United States)

Koglin, Johnathon

Accurate nuclear reaction data from a few keV to tens of MeV and across the table of nuclides is essential to a number of applications of nuclear physics, including national security, nuclear forensics, nuclear astrophysics, and nuclear energy. Precise determination of (n, f) and neutron capture cross sections for reactions in high- ux environments are particularly important for a proper understanding of nuclear reactor performance and stellar nucleosynthesis. In these extreme environments reactions on short-lived and otherwise difficult-to-produce isotopes play a significant role in system evolution and provide insights into the types of nuclear processes taking place; a detailed understanding of these processes is necessary to properly determine cross sections far from stability. Indirect methods are often attempted to measure cross sections on isotopes that are difficult to separate in a laboratory setting. Using the surrogate approach, the same compound nucleus from the reaction of interest is created through a "surrogate" reaction on a different isotope and the resulting decay is measured. This result is combined with appropriate reaction theory for compound nucleus population, from which the desired cross sections can be inferred. This method has shown promise, but the theoretical framework often lacks necessary experimental data to constrain models. In this work, dual arrays of silicon telescope particle identification detectors and photovoltaic (solar) cell fission fragment detectors have been used to measure the fission probability of the 240Pu(alpha, alpha'f) reaction - a surrogate for the 239Pu(n, f) - and fission of 35.9(2)MeV at eleven scattering angles from 40° to 140° in 10° intervals and at nuclear excitation energies up to 16MeV. Within experimental uncertainty, the maximum fission probability was observed at the neutron separation energy for each alpha scattering angle. Fission probabilities were separated into five 500 keV bins from 5:5MeV to
A new approach to make collapsed cross section for burnup calculation of subcritical system

International Nuclear Information System (INIS)

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

2008-01-01

A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)
Cross-Sections for Low-Energy Neutron-Induced Fission; Sections Efficaces de Fission pour des Neutrons de Faible Energie; 0421 0415 0427 0415 041d 0418 042f 0414 0415 041b 0415 041d 0418 042f , 0412 042b 0417 0412 0410 041d 041d 041e 0413 041e 041d 0415 0419 0422 0420 041e 041d 0410 041c 0418 041d 0418 0417 K 041e 0419 042d 041d 0415 0420 0413 0418 0418 ; Secciones Eficaces en la Fision Inducida por Neutrones de Baja Energia

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-15

The cross-sections of the heavy nuclei for neutron-induced fission are of fundamental importance to the technology of nuclear energy production. The manner in which these cross-sections vary with the neutron energy and with the mass and charge of the target nuclei also provides much information on the structure of heavy nuclei. In the case of a thermally fissile target nucleus, as the neutron energy is increased from thermal the cross-section first exhibits an inverse velocity dependence, followed by a region in which sharp resonance peaks appear, and finally a continuum region where the cross-section exhibits relatively smooth steps and breaks. All these phenomena can be explained in principle, but certain features of the data have proved very difficult to explain quantitatively. Recent cross-section measurements, stimulated by the needs of reactor technology, have concentrated on improving the energy resolution and accuracy of the data on fuel materials, and this has led to a more detailed study of the resonance region, which is of considerable interest for reactor Doppler effect calculations. More accurate measurements at rather higher energies have established the existence of appreciable fission cross-sections below the so-called fission threshold in certain cases. Careful measurements of this nature have in turn stimulated interest in the interpretation of the cross-sections in terms of nuclear models on a firm quantitative basis. This paper outlines the main features of neutron-induced-fission cross-sections and their interpretation. Some emphasis is placed on recent improvements in the quality of the measurements and in attempts at quantitative interpretation of certain aspects of the data. (author) [French] Les sections efficaces de fission des noyaux lourds ont une importance essentielle pour la technologie de la production d'energie d'origine nucleaire. La maniere dont ces sections efficaces varient Selon l'energie des neutrons et Selon la masse et la
Signatures of fission dynamics in highly excited nuclei produced in 197AU(800 A MeV) on proton collisions

International Nuclear Information System (INIS)

Benlliure, J.; Armbruster, P.; Bernas, M.

2001-09-01

197 Au(800 A MeV)-on-proton collisions are used to investigate the fission dynamics at high excitation energy. The kinematic properties together with the isotopic identification of the fission fragments allow to determine the mass, charge and excitation energy of the fissioning nucleus at saddle. The comparison of these observables and the measured total fission cross section with model calculations evidences a clear hindrance of fission at high excitation energy that can be explained in terms of nuclear dissipation. Assuming a statistical evaporation for other de-excitation channels than fission, an estimated value of the transient time of fission of (3 ± 1) . 10 -21 s is obtained. (orig.)
Comparison Between Calculated and Measured Cross Section Changes in Natural Uranium Irradiated in NRX

Energy Technology Data Exchange (ETDEWEB)

Ahlstroem, P E

1961-03-15

It is desirable to obtain an experimental check of the reliability of the methods currently used to determine reactivity changes in a reactor and, with a view to meeting this requirement to some extent, a preliminary comparison has been made between calculated and measured cross-section changes in rods of natural uranium irradiated in NRX. The measurements were made at Harwell in the GLEEP reactor and a description has been given by, inter alia, Ward and Craig. The theory of the calculations, which is briefly described in this report, has been indicated by Littler. The investigation showed that the methods for calculating burn up used at present provides a good illustration of the long-term variations in isotope contents. A satisfactory agreement is obtained with experimental results when calculating apparent cross-section changes in uranium rods due to irradiation if the fission cross- section for {sup 239}Pu is set to 780 b. This is 34 b higher than the figure quoted in BNL - 325 (1958). However, in order to get a good idea as to whether the calculated long-term variations in reactivity really correspond to reality, it is necessary to make further investigations. For this reason the results quoted in this report should be regarded as preliminary.
Proton-fission for the accelerator production of Mo-99

International Nuclear Information System (INIS)

Lagunas-Solar, M.C.; Jungerman, J.A.; Castaneda, C.M.

1993-01-01

The production of Mo-99 (66.0 h) via de U-238(p,f) Mo-99 fission reaction is proposed as a non-reactor source of this essential precursor of 6.6-h Tc-99m, an isotope of wide use of diagnostic nuclear medicine applications. Measurements of the total excitation function for the U-238(p,f) reaction indicated a maximum and fairly constant cross section of 1.4 barns at > 30 MeV. Combining the advances of high-current (mA) H-accelerators with dual beam (dual target) operation, and assuming a 5% fission yield, estimates of Mo-99 reaches 5 to 14 Ci/h at 1 mA. The proton fission production of Mo-99 appears to more advantageous than the reactor produced via evaporation neutron-induced fission. An accelerator method could allow securing ample supply of Mo-99 independently of the current scarce reactor operation, while also simplifying the associated waste management problems as well as some of the environmental concerns
Comparison of fission probabilities with emission of long range particles under the action of slow and fast neutrons on various materials; Probabilites comparees de fission avec emission de particules de long parcours pour divers materiaux sous l'action des neutrons lents et rapides

Energy Technology Data Exchange (ETDEWEB)

Netter, F; Faraggi, H; Garin-Bonnet, A; Julien, J; Corge, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Turkiewicz, J [Institut de Recherches Nucleaire de Varsovie (Poland)

1958-07-01

The authors describe relative cross-section measurements of fission of the isotopes of uranium and plutonium (more particularly {sup 235}U, {sup 238}U, {sup 239}Pu), with emission of long range particles, under the action of neutrons of various energies: thermal neutrons, pile neutrons, neutrons produced with the Van de Graaff accelerator by reaction of protons on tritium. The measurements are carried out: 1) with the aid of photographic plates, by submitting to the action of the neutrons a layer of fissile material coupled with an Ilford nuclear emulsion of 200 microns; a tin sheet laying between the plate and the layer stops the {alpha} particles and the fission fragments. By an appropriate development the tracks of the long range particles can be distinguished in the emulsion, from the tracks of the recoil protons resulting of fission neutrons, or of the last primary neutrons. For neutrons of energy under 1 MeV, the compared frequency of the tracks of long range particles and of the recoils caused by the fission neutrons gives a measurement of the fission cross-section with emission of long range particles relative to the product of the fission cross-section by the mean number of neutrons emitted by fission. For neutrons of higher energy, one measures only the frequency of the tracks of long range particles, comparatively with the flux of primary neutrons. Some precautions are taken to eliminate the action of thermal neutrons in the measurements with fast neutrons. 2) with the aid of a system of ionization chamber and proportional counter, the rate of coincidence between the impulsions caused by the long range particles and the impulsions provided by one of the fission fragments is measured comparatively with the counting rate of fission fragme (author) [French] Les auteurs decrivent des mesures relatives a la section efficace de fission des isotopes de l'uranium et du plutonium (notamment {sup 235}U, {sup 238}U, {sup 239}Pu) avec emission de particules de long
[Fast neutron cross section measurements]: Progress report

International Nuclear Information System (INIS)

1988-01-01

As projected in our previous proposal, the past year on the cross section project at the University of Michigan has been one primarily of construction and assembly of our 14 MeV pulsed Neutron Facility. All the components of the system have now been either purchased or fabricated in our shop facilities and have been assembled in their final configuration. We are now in the process of testing the rf components that have been designed to deliver voltage to both the pulser and buncher stages. We expect that the system will be operational by the end of the current contract year. We have also accomplished the design and construction of several other major pieces of equipment that are needed to begin fast neutron time-of-flight measurements. These include the primary proton recoil detector, and a californium fission chamber needed in the efficiency calibration of the primary detector. We have also added considerable concrete shielding designed to lower the neutron background in the experimental area. 10 figs., 5 tabs
Relativistic photon-Maxwellian electron cross sections

International Nuclear Information System (INIS)

Wienke, B.R.; Lathrop, B.L.; Devaney, J.J.

1986-01-01

Temperature corrected cross sections, complementing the Klein-Nishina set, are developed for astrophysical, plasma, and transport applications. The set is obtained from a nonlinear least squares fit to the exact photon-Maxwellian electron cross sections, using the static formula as the asymptotic basis. Two parameters are sufficient (two decimal places) to fit the exact cross sections over a range of 0-100 keV in electron temperature, and 0-1 MeV in incident photon energy. The fit is made to the total cross sections, yet the parameters predict both total and differential scattering cross sections well. Corresponding differential energy cross sections are less accurate. An extended fit to (just) the total cross sections, over the temperature and energy range 0-5 MeV, is also described. (author)
Fission-residues produced in the spallation reaction 238U + p at 1 A GeV

International Nuclear Information System (INIS)

Bernas, M.; Armbruster, P.; Benlliure, J.; Boudard, A.; Casarejos, E.; Czajkowski, S.; Enqvist, T.; Legrain, R.; Leray, S.; Mustapha, B.; Napolitani, P.; Pereira, J.; Rejmund, F.; Ricciardi, M.V.; Schmidt, K.H.; Stephan, C.; Taieb, J.; Tassan-Got, L.; Volant, C.

2003-04-01

Fission fragments from 1 A GeV 238 U projectiles irradiating a hydrogen target were investigated by using the fragment separator FRS for magnetic selection of reaction products including ray-tracing and ΔE-ToF techniques. The momentum spectra of identified fragments were analysed to provide isotopic production cross sections, fission-fragment velocities and recoil momenta of the fissioning parent nuclei. Besides their general relevance, these quantities are also demanded for applications. Calculations and simulations with codes commonly used and recently developed or improved are compared to the data. (orig.)
Integral nucleus-nucleus cross sections

International Nuclear Information System (INIS)

Barashenkov, V.S.; Kumawat, H.

2003-01-01

Expressions approximating the experimental integral cross sections for elastic and inelastic interactions of light and heavy nuclei at the energies up to several GeV/nucleon are presented. The calculated cross sections are inside the corridor of experimental errors or very close to it. Described in detail FORTRAN code and a numerical example of the cross section approximation are also presented

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