Spectrum Averaged Cross Sections Research Articles

The spectrum-averaged cross section investigation of 117Sn(n,n')117mSn and 67Zn(n,p)67Cu reactions

The spectrum averaged cross sections (SACS) in standard neutron field, e.g. 252Cf(s.f.), is a preferable tool for cross section evaluation and validation. A set of reaction measurements with high energy thresholds was previously performed. The presented work focuses on lower energy threshold reactions, namely on the inelastic scattering of the tin foil, more specifically the reaction 117Sn(n,n')117mSn, and the zinc foil reaction, namely 67Zn(n,p)67Cu. These reactions are of special interest due to their intermediate energy range, which is essential in classical reactor dosimetry and fast reactor dosimetry. The experiments were carried out in a standard neutron field formed by 252Cf(s.f.) source in Řež. The experimental results were compared with calculations using MCNP6.2, ENDF/B-VII.1 transport library, and ENDF/B-VIII.0 and IRDFF-II cross section data library. Additionally, the calculations using CEA code DARWIN/PEPIN2 using JEFF-3.0/A were executed. The obtained experimental SACS of previously measured reactions were in good agreement with the SACS calculated using the IRDFF-II library. Additionally, the calculational reaction rate of 67Zn(n,p)67Cu was in accordance with the experimental data in case of ENDF/B-VIII.0 nuclear data library. Moreover, the calculational results of 117Sn(n,n')117mSn obtained by DARWIN/PEPIN2 code (using JEFF-3.0/A nuclear data library) are closest to the experimental results.

Measurement of spectrum-averaged cross sections of the (n,p) and (n,n′) reactions on strontium by fast neutrons of a TRIGA reactor: comparison with integrated data from excitation functions of various data libraries

Abstract Integral cross sections of the reactions 84Sr(n,p)84Rb, 86Sr(n,p)86Rb and 87Sr(n,n′)87mSr, over the fast neutron spectrum of a TRIGA reactor above 0.5 MeV, were measured for the first time. The results were used for integral tests of the evaluated excitation functions in the data libraries, ENDF/B-VIII.0, ROSFOND-2010, JEFF-3.3, JENDL-5 and TENDL-2021. The integral measurements are generally consistent with the integrated values within 12 %, except for a few cases. The two data libraries ENDF/B-VIII.0 and ROSFOND-2010 appear to need some improvement with regard to the reactions investigated in this work.

Study of 239Pu(n,f) prompt fission neutron spectrum

The experimental data concerning 239Pu prompt fission neutron spectrum above 10 MeV are sparse, and no spectrum averaged cross section has been measured in this neutron field yet. The first measurement was performed at the IBR-2 reactor, which uses plutonium dioxide fuel. The present work aimed at the measurement of spectral averaged cross sections for high threshold reactions such as 46Ti(n,p)46Sc, 54Fe(n,p)54Mn, 58Ni(n,X)57Co, 89Y(n,2n)88Y, and 209Bi(n,3n)207Bi to explore the high energy region of the 239Pu prompt fission neutron spectrum. Measurement of spectrum averaged cross section for the 89Y(n,2n)88Y reaction is included in NEA high priority request list. The 58Ni(n,p)58Co reaction was employed as a monitor reaction. The experimental spectral averaged cross sections were derived from reaction rates measured by means of high purity Germanium spectrometry. The experimental values were compared with spectrum averaged cross sections compiled from either ENDF/B-VIII.0, GMA or JEFF-3.3 239Pu prompt fission neutron spectrum and IRDFF-II cross sections. The best agreement, i.e. within uncertainties, was achieved using GMA 239Pu prompt fission neutron spectrum for all reactions under study. Calculation results using JEFF-3.3 and ENDF/B-VIII.0 evaluations of the 239Pu prompt fission neutron spectrum agree well with the experimental data except for the high energy region reflected in the disagreement with the 209Bi(n,3n)207Bi reaction.

Experimental spectrum averaged cross sections (SACS) in 252Cf(sf) neutron field and its impact on the evaluation of neutron standards

A new evaluation of spectrum averaged cross sections (SACS) of 235U, 238U, and 239Pu measured in the 252Cf(sf) reference neutron field is presented and found to be consistent with the original Mannhart SACS evaluation in IRDF-2002. The comprehensive vetted experimental database that includes SACS ratio and absolute SACS measurements of major actinides is being used to update the SACS database employed as input of the new GMApy code to derive the Neutron Standards. An update of the current neutron standards based on Time Projection Chamber (TPC) shape data, a new comprehensive uncertainty quantification, and revised SACS experimental database is proposed which result in a 0.7% increase of the evaluated 239Pu(n,f)/235U(n,f) crosssection ratio in the 1–5 MeV energy region. The increase is due to a 0.3% reduction of the standard 235U(n,f) cross section and a 0.4% increase of the 239Pu(n,f) reference cross section in the 1–5 MeV energy region. Those changes are well within estimated USU fission cross-section uncertainties of 1.2%, but are relevant for the evaluated mean values.

Direct reactions for nuclear astrophysics

The neutron activation technique is a well established method to measure neutron capture cross sections relevant for the s-process. The7Li(p,n) reaction at Ep= 1912 keV is often used as a neutron source since the energy distribution of the emitted neutrons closely resembles a Maxwell-Boltzmann spectrum ofkBT= 25 keV, mimicking the22Ne(α,n) phase in TP-AGB stars. The weak s-process, which takes place in massive stars, can reach energies up tokBT = 90 keV. Neutron spectra corresponding to a Maxwell-Boltzmann distribution withkBT> 25 keV cannot be produced by the7Li(p,n) reaction directly. We developed a method to obtain quasi-Maxwellian neutron capture cross sections over a wide energy range by combining a set of spectrum average cross sections measured at six different proton energies and distances between the lithium target and the sample. The measured spectrum averaged cross section can be used to calculate the Maxwellian-Averaged cross-section (MACS) fromkBT= 25 keV tokBT= 90 keV. Over the last two years neutron capture cross sections on over 20 isotopes have been measured at Goethe University Frankfurt using this methodology. An overview of the current experimental method, challenges during data analysis and the first results are presented.

Simultaneous evaluation of uranium and plutonium fast neutron fission cross sections up to 200 MeV for JENDL-5 and its updates

Simultaneous evaluation of 233,235,238U and 239,240,241Pu fission cross sections for fast neutrons up to 200 MeV was performed for the JENDL-5 library. Experimental covariances were estimated for each experimental dataset of cross sections or cross section ratios extracted from the EXFOR library, and they were stored in an experimental database dedicated to the new evaluation. The cross sections were expressed by Schmittroth’s roof functions, and the values on defined incident energy grids were adjusted by the least-squares method to reproduce the experimental cross sections and their ratios. The newly evaluated cross sections were validated using the spectrum averaged cross sections measured in the 252Cf spontaneous fission neutron standard field. The evaluation adopted by the JENDL-5 library was further updated by addition of two datasets and deletion of one dataset.

Impact of reactor neutron spectrum on measured spectrum averaged cross sections

The cross section averaged over 235U thermal-neutron induced fission spectrum is a fundamental quantity that can be used in evaluation and validation of nuclear data. Many experiments focused on the determination of Spectrum Averaged Cross Sections (SACS) in 235U(nth,f) Prompt Fission Neutron Spectrum (PFNS) in light water reactors using enriched uranium fuel. In these reactors, there is already some amount of water moderator between the uranium fuel and the irradiated sample. Due to the decrease of hydrogen cross-section with neutron energy, the high energy tail of the reactor spectrum in cores with water moderator may be harder than the pure PFNS. This paper aims to compare the shape of the actual reactor spectrum in various core positions of a research light-water reactor differing each from other by the effective water thickness. The spectrum shape is determined both by calculations and experimentally using various high energy threshold reactions.The impact of the photo-nuclear reactions (γ,n) competing with (n,2n) in production of the same residual nucleus was shown to be less than a percent for most of studied dosimeters. An important exception was found for 197Au(n,2n)196Au dosimeter irradiated in the outcore channel where a notable photo-neutron contribution to the production of 196Au is caused by the neutron production from the high energy γ-rays from thermal-neutron capture in 54Fe. The corresponding ENDF/B-VIII.0 data turned out to underestimate such γ-yield by 40 % in comparison with ENDF/B-VI.8. This has improved but however not resolved the disagreement between our measurement and calculations. The remaining deficiency was attributed to the underestimation of the evaluated cross section IAEA/PD-2019 for the 197Au(γ,n) cross section near the reaction threshold. The later was confirmed by comparison with existing measured data.

Spectrum averaged cross section measurements of lutetium using standard 252Cf neutron source

The spectrum averaged cross section (SACS) in a standard neutron field is a preferable tool for cross section validation. The presented work uses only neutron standard, i.e., 252Cf(sf) reaction neutron field, for validation of lutetium threshold cross sections. SACS were inferred from gamma spectrometry derived reaction rates. The SACS which were derived include 175Lu (n,2n)174Lu, 175Lu (n,3n)173Lu, 175Lu (n,p)175Yb, and 176Lu (n,n')176m1Lu reactions. All these reactions SACS were measured for the first time. MCNP6.2 calculations using JEFF-3.3 or ENDF/B-VIII.0 libraries for lutetium cross sections were compared with experimental data. The agreement was found very poor for all reactions under study. Thus there is a need for their improvement. The presented data can be also used for validation of the various theoretical models.

EXFOR-based simultaneous evaluation of neutron-induced uranium and plutonium fission cross sections for JENDL-5

ABSTRACT The neutron-induced fission cross sections were simultaneously evaluated for the JENDL-5 library for 233,235U and 239,241Pu from 10 keV to 200 MeV and for 238U and 240Pu from 100 keV to 200 MeV. Evaluation was performed by least-squares fitting of Schmittroth’s roof function to the logarithms of the experimental cross sections and cross section ratios in the EXFOR library. A simultaneous evaluation code SOK was used with its extension to data in arbitrary unit. The outputs of the code were adopted as the evaluated cross sections without any further corrections. The newly obtained evaluated cross sections were compared with the evaluated cross sections in the JENDL-4.0 library and the IAEA Neutron Data Standards 2017. The evaluated cross sections were also validated against the californium-252 spontaneous fission neutron spectrum averaged cross sections, (coupled thermal/fast uranium and boron carbide spherical assembly) neutron spectrum averaged cross sections, and small-sized LANL fast system criticalities. The changes in the obtained evaluated cross sections from those in the JENDL-4.0 library are within 4% (241Pu), 3% (233U, 240Pu), or 2% (235U, 239Pu). The newly evaluated 235U, 238U and 239Pu cross sections agree with the IAEA Neutron Data Standards 2017 within 2% with some exceptions.

An overview of activation cross-section measurements of some neutron and charged-particle induced reactions in Bangladesh

Abstract Over the last several decades, our research group has been working on nuclear data measurements using a J-25 Neutron Generator, a TRIGA Mark II Research Reactor and a Tandem Accelerator at Savar, as well as an Am/Be neutron source at Rajshahi. The spectrum of fast neutrons having energies from 0.5 to 20 MeV in the core of the TRIGA reactor as well as from 1.5 to 11 MeV of the Am/Be source was well characterized for validation of evaluated data of neutron threshold reactions. Spectrum-averaged cross sections were determined by integral measurements. The integral values measured at both the reactor and the Am/Be source were compared with the integrated data deduced from measured and evaluated excitation functions of some nuclear reactions given in some data libraries. As example, the details on the integral and integrated cross sections of the 64Zn(n,p)64Cu and 60Ni(n,p)60Co reactions are discussed. In both cases, integral test shows that further improvement of the evaluated excitation function reported in data libraries is needed. The use of a newly developed facility at the 3 MV Tandem Accelerator in measurement of proton induced reaction cross sections in the energy region below 5 MeV is outlined, and new results on the 64Ni(p,n)64Cu and natNi(p,x)61Cu processes below 3 MeV are discussed.

\u2009Prompt and delayed gamma rays induced by epithermal and fast neutrons with indium

The emission of prompt and delayed gamma rays from (n,γ) and (n,n´γ) reactions induced by irradiation of indium with epithermal and fast neutrons was investigated with the instrument FaNGaS operated at Heinz-Maier-Leibnitz Zentrum (MLZ) in Garching. The average neutron energy of the neutron spectrum was 2.30 MeV. The measurement was done at an angle of 90° between neutron beam and detector. A total of 136 prompt gamma lines from which 42 are related to the capture of epithermal and fast neutrons and 94 to the inelastic scattering of fast neutrons were detected together with the delayed gamma lines of the activation products 113mIn, 114m2In, 115mIn, 116m2In and 116mIn. Intensities and neutron spectrum averaged isotopic partial cross section of the gamma lines are presented. Additionally the neutron spectrum averaged cross sections of the reactions, 113In(n,n´)113mIn, 113In(n,γ)114m2In, 115In(n,n´)15mIn, 115In(n, γ)116m2In and 115In(n, γ)116mIn were determined from the corresponding delayed gamma rays of the formed isotopes as 143 ± 22, 288 ± 13 194 ± 18, 201 ± 10 and 508 ± 24 mb respectively. The various results obtained were found consistent with the literature data. However, our measurement indicate the need to reevaluate the cross section of the 115In(n,γ)116m2In reaction for thermal neutrons.

Validation of IRDFF-II library in VR-1 reactor field using thin targets

Cross section data are fundamental quantities which affect the accuracy of all calculations in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations with full uncertainty quantification was developed by the International Atomic Energy Agency and released in January 2020 (https://www-nds.iea.org/IRDFF). A previous version, IRDFF-1.05, was released in 2014 and experimental validation of the newly released cross section by spectrum averaged cross section (SACS) measurements is a high priority task. For such purpose, a neutron dosimeter set containing 5 target foils was activated in 2 independent experiments at the VR-1 reactor of the Czech Technical University in Prague. Care was taken to derive SACS with low uncertainties. New experimental evaluation method is in good agreement with previous approaches based on relative measurements using monitor foils. Good agreement of measured SACS and evaluated IRDFF-II cross sections is observed. Slight overestimation of evaluated ENDF/B-VIII.0 235U(nth,f) PFNS above 10 MeV is discussed.

Neutron activation of Ga69 and Ga71 at kBT≈25keV

Background: About 50% of heavy elements are produced by the slow neutron capture process ($s$ process) in stars. The element gallium is mostly produced during the weak $s$ process in massive stars.Purpose: Our activation at ${k}_{\mathrm{B}}T\phantom{\rule{4pt}{0ex}}\ensuremath{\approx}25$ keV is the first experiment in a series of activation and time-of-flight measurements on $^{69}\mathrm{Ga}$ and $^{71}\mathrm{Ga}$ relevant for astrophysics.Methods: We activated $^{69}\mathrm{Ga}$ and $^{71}\mathrm{Ga}$ with a neutron distribution that corresponds to a quasistellar distribution with ${k}_{\mathrm{B}}T=25\phantom{\rule{4pt}{0ex}}\mathrm{keV}$ at the Joint Research Centre (JRC), Geel, Belgium. Protons were provided by an electrostatic Van de Graaff accelerator to produce neutrons via the reaction $^{7}\mathrm{Li}(\mathrm{p},\mathrm{n})$. The produced activity was measured via the $\ensuremath{\gamma}$ emission by the decaying product nuclei by high-purity germanium detectors.Results: We provide spectrum-averaged cross sections (SACS) and ratios of the cross sections ${\ensuremath{\sigma}}_{\text{Ga}}/{\ensuremath{\sigma}}_{\text{Au}}$ for the neutron spectrum of the activation. We obtain values of ${\ensuremath{\sigma}}_{69\mathrm{Ga},\mathrm{SACS}}=(186\phantom{\rule{0.16em}{0ex}}\ifmmode\pm\else\textpm\fi{}\phantom{\rule{0.16em}{0ex}}12)\phantom{\rule{4pt}{0ex}}\mathrm{mb}$ and ${\ensuremath{\sigma}}_{71\mathrm{Ga},\mathrm{SACS}}=(112\phantom{\rule{0.16em}{0ex}}\ifmmode\pm\else\textpm\fi{}\phantom{\rule{0.16em}{0ex}}7)\phantom{\rule{4pt}{0ex}}\mathrm{mb}$, and cross-section ratios of ${\ensuremath{\sigma}}_{\text{69Ga}}/{\ensuremath{\sigma}}_{\text{Au}}=0.29\phantom{\rule{4pt}{0ex}}\ifmmode\pm\else\textpm\fi{}\phantom{\rule{0.16em}{0ex}}0.02$ and ${\ensuremath{\sigma}}_{\text{71Ga}}/{\ensuremath{\sigma}}_{\text{Au}}=0.17\phantom{\rule{4pt}{0ex}}\ifmmode\pm\else\textpm\fi{}\phantom{\rule{0.16em}{0ex}}0.01$.Conclusions: Our data disagree with the available evaluated data provided by KADoNiS v0.3, our cross-section ratio is about 20% higher for $^{69}\mathrm{Ga}$ and about 20% lower for $^{71}\mathrm{Ga}$.

MEASUREMENT OF INTEGRAL CROSS SECTIONS OF SELECTED DOSIMETRY REACTIONS IN LR-0 REACTOR

The cross section is a fundamental quantity which affects the accuracy of Monte Carlo simulations widely used in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations that include full uncertainty quantification is being developed by the International Atomic Energy Agency and expected to be released in January 2020; a preliminary version IRDFF-1.05 was released in 2014 and is being tested in this work. Validation of the cross-section evaluations proposed for this library is a high priority task. The validation can be realized using integral cross sections measured in standard and/or reference neutron benchmark fields. Integral quantities feature significantly lower uncertainties than differential nuclear data. If the neutron spectrum where the cross section is measured is well characterized, then the Spectrum Averaged Cross Section can be used for validating of existing evaluations.

Neutron radiative capture cross section for sodium with covariance analysis

The neutron radiative capture cross sections measurement has been carried out for the $$^{23}$$ Na nucleus in the neutron energy region from 0.6 to 3.2 MeV using the neutron activation technique followed by off-line $$\gamma $$ -ray spectrometry. The measurement was made relative to the $$^{115}$$ In(n,n $$\prime $$ $$\gamma $$ ) $$^{115}\hbox {In}^{m}$$ reference monitor reaction cross section. The neutrons were produced via the $$^{7}$$ Li(p,n) $$^{7}$$ Be reaction. Detailed uncertainty propagation has been performed using the covariance analysis, and the measured cross sections are being reported with their uncertainties, covariance, and correlation matrix. The necessary corrections have been made for the low background neutron energy contribution, $$\gamma $$ -ray true coincidence summing, and self-attenuation process. The obtained neutron spectrum averaged cross sections of $$^{23}$$ Na(n, $$\gamma $$ ) $$^{24}$$ Na are discussed and compared with the existing cross sections data retrieved from the EXFOR database. EMPIRE-3.2 and TALYS-1.9 calculations were performed in order to determine the radiative capture cross section in this energy region. The present results are also compared with the evaluated nuclear data from ENDF/B-VIII.0, TENDL-2019, IRDFF-1.05, JENDL-4.0, and JEFF-3.3. The obtained cross section results are in good agreement with existing experimental data, evaluated libraries, and reaction models for the highest energy points (2.11 and 3.13 MeV), while the lowest-energy point at 0.61 MeV underestimates them.

Methodology of 63Cu(n,2n)62Cu Reaction Rate Measurement

Abstract This paper presents the measurement of the spectrum-averaged cross section (SACS) of 63Cu(n,2n)62Cu reaction in 252Cf spontaneous fission neutron spectrum. The SACS in the 252Cf spectrum was chosen as a validation tool since 252Cf is the only standard neutron field and 62Cu isotope is not easy to measure by gamma spectroscopy since the gamma line of interest is an annihilation peak, which is also produced by 64Cu isotope. Fortunately, contributions to the annihilation peak from these isotopes can be distinguished due to the very different half-lives. SACS was inferred from the experimental reaction rate. The SACS in the 252Cf spontaneous fission neutron field for the 63Cu(n,2n)62Cu reaction was determined as equal to (0.1763 ± 0.0077 mb). This value agrees with value (0.183 ± 0.007) × 10−3 b within uncertainty presented by W. Mannhart. However, it differs by 12.7% from IRDFF-II value, which is equal to (0.19874 ± 8.954 10−3) × 10−3 b. Furthermore, reasonable agreement is not achieved with ENDF/B-VIII.0, JEFF-3.3, CENDL-3.1, ROSFOND-2010, nor JENDL-4.0 nuclear data libraries.

Ratio of spectral averaged cross sections measured in standard 252Cf(sf) and 235U(nth,f) neutron fields

The results of systematic evaluations of spectrum averaged cross section (SACS) measurements in the fission neutron fields of 252Cf and 235U are presented. The data form a complete database of high-threshold experimental SACS measured in the same installation under the same conditions and using the same high purity germanium gamma spectrometer. This is crucial to reduce the uncertainty of the ratio and the data scattering and therefore, to minimize discrepancies compared to cross section measured under different conditions in different laboratories. This new dataset complements and extends earlier experimental evaluations. The total emission of the 252Cf neutron source during the experiments varied from 9.5E8 to 4.5E8 neutrons per second. The emission was derived in accordance to the data in the Certificate of Calibration involving absolute flux measurements in a manganese sulphate bath. Concerning 235U fission neutron field, the irradiations were carried out in a specifically designed core assembled in the zero power light water LR-0 reactor. This special core has a well described neutron field. After the irradiation, the low volume irradiated samples to be measured by gamma spectrometry were placed directly on the upper cap of a coaxial high purity germanium (HPGe) detector in a vertical configuration (ORTEC GEM35P4). High volume samples were homogenized and strewn into the Marinelli beaker. The HPGe detector is surrounded by the lead shielding box with a thin inner copper cladding and covered with rubber for suppression of background signal and bremsstrahlung. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas (NPA) measured using the semiconductor HPGe detector. The measured reaction rates are used to derive the spectrum-averaged cross sections. Furthermore, measured reaction rates are also compared with MCNP6 calculations using various nuclear data libraries, in particular IRDFF evaluations.

IRDFF-II: A New Neutron Metrology Library

High quality nuclear data is the most fundamental underpinning for all neutron metrology applications. This paper describes the release of version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) that contains a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60 MeV neutron energy. The library is intended to support: a) applications in research reactors; b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new IRDFF-II library includes 119 metrology reactions, four cover material reactions to support self-shielding corrections, five metrology metrics used by the dosimetry community, and cumulative fission products yields for seven fission products in three different neutron energy regions. In support of characterizing the measurement of the residual nuclei from the dosimetry reactions and the fission product decay modes, the present document lists the recommended decay data, particle emission energies and probabilities for 68 activation products. It also includes neutron spectral characterization data for 29 neutron benchmark fields for the validation of the library contents. Additional six reference fields were assessed (four from plutonium critical assemblies, two measured fields for thermal-neutron induced fission on 233U and 239Pu targets) but not used for validation due to systematic discrepancies in C/E reaction rate values or lack of reaction-rate experimental data. Another ten analytical functions are included that can be useful for calculating average cross sections, average energy, thermal spectrum average cross sections and resonance integrals. The IRDFF-II library and comprehensive documentation is available online at www-nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at www-nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications.

Study of Astrophysical s-Process Neutron Capture Reactions at the High-Intensity SARAF-LiLiT Neutron Source

We report on recent experiments at the Soreq Applied Research Accelerator Facility Liquid-Lithium Target (SARAF-LiLiT) laboratory dedicated to the study of s-process neutron capture reactions. The kW-power proton beam at 1.92 MeV (1-2 mA) from SARAF Phase I yields high-intensity 30 keV quasi-Maxwellian neutrons (3-5×1010 n/s). The high neutron intensity enables Maxwellian averaged cross sections (MACS) measurements of low-abundance or radioactive targets. Neutron capture reactions on the important s-process branching points 147Pm and 171Tm were investigated by activation in the LiLiT neutron beam and γ-measurements of their decay products. MACS values at 30 keV extracted from the experimental spectrum-averaged cross sections are obtained and will be discussed. The Kr region, at the border between the so-called weak and strong s-process was also investigated. Atom Trap Trace Analysis (ATTA) was used for the first time for the measurement of a nuclear reaction cross section. After activation in the quasi-Maxwellian neutron flux at SARAF-LiLiT, isotopic ratios were determined for 81Kr(230 ky)/80Kr and 85gKr(10.8 y)/84Kr. The latter ratio was confirmed both by low-level β counting and γ spectrometry. The shorter-lived capture products 79,85m,87Kr were detected by γ -spectrometry and the corresponding neutron-capture MACS of the respective target nuclei 78,84,86Kr were determined. The MACS of the 80Kr(n, γ)81Kr and 84Kr(n, γ)85gKr reactions are still under study. The partial MACS leading to 85mKr(4.5 h) measured in this experiment has interesting implications since this state decays preferentially by γ decay (79%) to 85Rb on a faster time scale than does 85gKr and behaves thus as an s-process branching point.

Integral measurement of spectrum-averaged cross sections of a few threshold reactions induced by fast neutrons of a TRIGA reactor: comparison with integrated data from excitation functions given in various data libraries

Abstract Integral cross sections of the reactions 24Mg(n,p)24Na, 27Al(n,p)27Mg, 27Al(n,α)24Na, 58Ni(n,d + np)57Co and 60Ni(n,p)60Co were measured for the first time using the fast neutron spectrum of a TRIGA reactor extending from 0.5 to 20 MeV. The values obtained in this work were comparable with the recommended values for a pure 235U prompt fission spectrum. The measured integral value was utilized for integral test of excitation function of each reaction given in some data libraries, namely ENDF/B-VIII.0, TENDL-2017, IRDFF-1.05 and ROSFOND-2010. The integral measurements are generally consistent with the integrated values within 5 %, except for a few cases, e. g. the reaction 60Ni(n,p)60Co, where the data libraries appear to need improvement.