Neutron Spectrum Averaged Cross Sections Research Articles

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.

\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.

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.

Calculations of Nuclear Astrophysics and Californium Fission Neutron Spectrum Averaged Cross Section Uncertainties Using ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and Low-fidelity Covariances

Nuclear astrophysics and californium fission neutron spectrum averaged cross sections and their uncertainties for ENDF materials have been calculated. Absolute values were deduced with Maxwellian and Mannhart spectra, while uncertainties are based on ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and Low-Fidelity covariances. These quantities are compared with available data, independent benchmarks, EXFOR library, and analyzed for a wide range of cases. Recommendations for neutron cross section covariances are given and implications are discussed.

Characterization of neutrons from a thick target 9Be(p, n) reaction by multiple foil activation technique for practical applications: measurement of integral cross sections and their comparison with integrated data from data files

Abstract The neutron spectrum in the energy range of 0.1 to 5.85 MeV in the 0○ direction with respect to the incident proton beam accelerated in a Tandem Accelerator of the Wakasa Wan Energy Research Center (WERC), Japan, has been characterized using the multiple foil activation technique. Flux mapping of thermal and epithermal neutrons at 0○, 20○, 40○ and 60○ at downstream of the beam has also been carried out. The fast neutron spectrum averaged cross sections of several threshold reactions were measured. Two monitor reactions, 58Ni(n, p)58Co and 115In(n, n′, γ)115mIn, were used to determine the fast neutron flux. The use of fast neutron flux in validating some excitation functions is described.

Fission neutron spectrum averaged cross sections for threshold reactions on arsenic

We have measured the cross sections, averaged over a 235U fission neutron spectrum, for the two high threshold reactions: 75As(n,p)75mGe and 75As(n,2n)74As. The measured averaged cross sections are 0.292±0.022 mb, referred to the 3.95±0.20 mb standard for the 27Al(n,p)27Mg averaged cross section, and 0.371±0.032 mb referred to the 111±3 mb standard for the 58Ni(n,p)58m+gCo 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 235U fission neutron spectrum. The calculated cross sections, though depending on analytical representation of the flux, agree with the measured values within the estimated uncertainties.

Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on zirconium: Production possibility of no-carrier-added 90Y in a nuclear reactor

Summary Employing the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several (n, p) and (n, α) reactions on isotopes of zirconium; our measurements constitute the first systematic study. The special interest was in the investigation of the 90Zr(n, p)90Y reaction for the production of no-carrier-added 90Y in a nuclear reactor. At a fast neutron flux density of 7.5 × 1013 cm−2 s−1 and an irradiation time of 120 h, using 100% enriched 90Zr target 0.548 GBq of no-carrier-added 90Y per batch could be produced. The method is thus suitable for medium-scale production of this radionuclide.

Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on zinc: small-scale production of no-carrier-added 64Cu in a nuclear reactor

Summary Fission neutron spectrum averaged cross sections of some threshold reactions like (n,p), (n, α) and (n,2n) on zinc were measured using the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy. Special attention was paid to the formation of 64Cu via the 64Zn(n,p) 64Cu reaction since 64Cu labeled compounds are potentially important radiopharmaceuticals. At a fast neutron flux density of 7.5 × 1013 cm-2s-1 and using a 100% enriched 64Zn target, 35 GBq activity of no-carrier-added 64Cu per batch could be produced. The process seems to be interesting from the medical application viewpoint, if a cyclotron is not available. A fast neutron source, on the other hand, would lead to significantly higher yield of 64Cu.

Fission neutron spectrum averaged cross sections of some threshold reactions on cadmium: production feasibility of no-carrier-added 103Pd in a nuclear reactor

Summary 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 103Pd via the 106Cd(n,α) 103Pd reaction since it is an important therapeutic radionuclide. At a fast flux neutron density of 7.5 × 1013 cm2s-1 and an irradiation time of 120 h, using 100% enriched 106Cd target 340 MBq of no-carrier-added 103Pd per batch could be produced. The method is thus suitable for medium-scale production of this radionuclide.

Fission spectrum averaged cross section measurements of some neutron threshold reactions of relevance to medical radionuclide production

Summary A resume is given of radiochemical measurements of nuclear reaction cross sections relevant to the production of some medically important radionuclides carried out at PINSTECH during the last decade. Systematic studies on fission neutron spectrum averaged cross sections of several threshold reactions, like (n, 2n), (n, p) and (n, α), on titanium, ruthenium, europium and dysprosium for the production of 45Ti, 96Tc, 153Sm and 153Gd are described.

Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on dysprosium: small-scale production of no-carrier-added153Gd in a nuclear reactor

SummaryEmploying the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several (n, 2n), (n,p) and (n, α) reactions on isotopes of dysprosium; our measurements constitute the first systematic studies. Of special interest was the investigation of the156Dy(n, α)153Gd reaction for the production of no-carrier-added153Gd in a nuclear reactor. Using 100% enriched156Dy target, 8633 KBq of153Gd per batch can be produced which is, however, not sufficient for medical application.

Fission neutron spectrum averaged cross sections of some threshold reactions on titanium: production feasibility of 45Ti in a nuclear reactor

Summary Systematic studies on fission neutron spectrum averaged cross sections of some threshold reactions like (n, 2n), (n, p) and (n, α) on titanium 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 45Ti via the 46Ti(n, 2n)45Ti reaction since 45Ti labeled compounds are potentially important radiopharmaceuticals. At a fast flux neutron density of 7.5 × 1013 cm−2 s−1 and using 100% enriched 46Ti target, 8.58 MBq saturation activity of 45Ti per batch can be produced. Due to rather low activity and low specific activity the procedure is not feasible from medical application point of view.

Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on europium: small scale production of no-carrier-added153Sm in a nuclear reactor

SummaryEmploying the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several (n, 2n), (n, p) and (n, α) reactions on isotopes of europium. Our measurements constitute the first systematic studies. Of special interest was the investigation of153Eu(n, p)153Sm reaction for the production of no-carrier-added153Sm in a nuclear reactor. Using 100% enriched153Eu target, 97.21 MBq153Sm per batch can be produced which is, however, not sufficient for medical application.

Use of the Absolute Parametric Method to Evaluate the Analytical Significance of Threshold Reactions on P, Si, Al and Mg by NAA

We investigated the use of threshold reactions on magnesium, aluminum, silicon and phosphorus to determine elemental concentrations and also to correct analytical interferences, using the parametric method. The reactor neutron spectrum averaged cross sections of the threshold reactions on these elements were referred to the (111±3) mb value for the 58Ni(n,p)58g+mCo reaction cross section. The results of the cross section measurements are: 25Mg(n,p)25Na, 1.57±0.12 mb; 28Si(n,p)28Al, 5.68±0.25 mb; 29Si(n,p)29Al, 3.02±0.15 mb; 30Si(n,α)27Mg, 0.1346±0.0090 mb; and 31P(n,α)28Al, 1.912±0.083 mb. The values for the cross sections of the following reactions: 24Mg(n,p)24Na, 1.441±0.092 mb; 27Al(n,p)27Mg, 3.84±0.18 mb; and 27Al(n,α)24Na 0.721±0.028 mb were re-measured to check the methodology. We present the application of these values to the analysis of some reference materials.

Determination of the fission averaged cross sections of some threshold reactions induced on chlorine, potassium, and calcium, and their role in the development of self-validated methods of neutron activation analysis

The feasibility of validating data obtained in the determination of chlorine, potassium, and calcium by neutron capture activation analysis, through their determination using threshold reactions has been studied. To this purpose the authors experimentally determined the following fission neutron spectrum averaged cross sections (in mb):37Cl(n,p)37S, 0.234±0.015;41K(n,α)38Cl, 0.740±0.044;42Ca(n,p)42K, 3.09±0.22;43Ca(n,p)43K, 2.27±0.12;44Ca(n,p)44K, 0.074±0.013. The application of these values to the analysis of some reference materials are presented. Also, the problem of mutual interferences has been addressed for the following pairs:36S(n,Γ)37S and37Cl(n,p)37S;37Cl(n,Γ)38Cl and41K(n,α)38Cl;41K(n,ψ)42K and42Ca(n,p)42K.

Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Tin: Small Scale Production of 111In in a Nuclear Reactor

Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Tin: Small Scale Production of ¹¹¹In in a Nuclear Reactor

Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Tellurium: Estimation of Radioactive Impurities in (n,y)-Produced 131Te-131I

Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Tellurium: Estimation of Radioactive Impurities in (n,y)-Produced ¹³¹Te-¹³¹I

Measurements of U-235 Fission Neutron Spectrum Averaged Cross Sections for Threshold Reactions

Measurements of U-235 Fission Neutron Spectrum Averaged Cross Sections for Threshold Reactions

Measurement of Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Nickel: Small Scale Production of 57Co in a Nuclear Reactor

Measurement of Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Nickel: Small Scale Production of ⁵⁷Co in a Nuclear Reactor

Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Molybdenum: Estimation of Radioactive Impurities in (n, γ)-Produced 99Mo-99mTc Generator System

Fission Neutron Spectrum Averaged Cross Sections of some Threshold Reactions on Molybdenum: Estimation of Radioactive Impurities in (n, γ)-Produced ⁹⁹Mo-^99mTc Generator System