Thermal Capture Cross Section Research Articles

Investigation of the Cd-Ratio difference method to measure the cross section of 186W(n,γ)187W at the Jordan Research and Training reactor

The neutron capture cross section of 186W(n,γ)187W was measured at different energies, including the thermal capture cross section (0.025 eV) and resonance integral. Neutron capture cross section values were determined using the activation technique and by applying the Cadmium-ratio difference method. Two irradiation locations at the Jordan Research and Training Reactor (JRTR) were used (NAA1 and NAA2). The thermal capture cross section values were determined to be 39.56 ± 0.69 ± 2.58b and 32.17 ± 3.72 ± 2.00b at NAA1 and NAA2, respectively. While resonance integral values were found to be 113.93 ± 8.20 ± 16.87b and 69.35 ± 26.63 ± 29.69b at NAA1 and NAA2, respectively. The present results of the resonance integral are not in good agreement with the most of the previous experimental and evaluated data as well as the theoretical calculations.

Robust determination of kerogen properties in organic-rich mudrocks via Raman Spectroscopy

Organic-rich mudrocks are an expansive resource of hydrocarbons that support the global energy demand. Effective characterization and appraisal of these resources relies upon accurate volumetric quantification of organic phases in the formation, which include producible hydrocarbons (oil and gas) and an abundant, nonproducible, insoluble solid (kerogen). A substantial uncertainty in this volumetric quantification comes from the variable, and nearly always unknown, composition and properties of kerogen. This paper demonstrates a novel method to quantify relevant properties of type II kerogen in organic-rich mudrocks from Raman spectroscopy measurements of bulk samples. The correlations are developed using a geographically and geologically diverse set of organic-rich mudrocks containing type II kerogens, with ages spanning more than 200 Ma and hydrocarbon generation stages ranging from immature to dry gas. The properties comprise kerogen’s carbon mass fraction, density, hydrogen index, apparent thermal neutron porosity, and macroscopic cross section for thermal neutron capture. These properties are estimated from empirical correlations between these kerogen property values quantified by laboratory reference techniques and kerogen spectral characteristics quantified from Raman measurements. The Raman-based method is a practical alternative for quantification of kerogen properties that are critical for the correct interpretation of common wellbore logging measurements made in organic-rich mudrocks.

Simulation-based studies on graphite absorption properties for ASTRA critical experiments

This work is focused on the improvement of the graphite impurity content characterization for the ASTRA critical facility. The equivalent boron content for the graphite reflectors of the ASTRA facility reported in the benchmark specifications was estimated using the results of the exponential experiments performed by graphite manufacturer. Because raw measured data are not available for these experiments, we tried to reassess the equivalent boron content for the ASTRA graphite reflectors using the results of analogous exponential experiments performed earlier on the empty RBMK critical facility at the National Research Center «Kurchatov Institute» and further simulation of the experiments utilizing the Monte Carlo code Serpent with ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0 cross-section libraries. The same grade graphite was used for manufacturing the graphite blocks for the RBMK critical facility and the ASTRA critical facility. The Monte Carlo simulations confirmed the accuracy of the exponential measurements performed for RBMK graphite characterization. It was also shown that the simulation must reproduce the graphite absorption cross-section estimated using the exponential experiments for the agreement between the measured and calculated distribution of thermal neutron flux. Therefore, the equivalent boron content should be assessed individually for each cross-section library used in the simulation. In case of the ASTRA facility the carbon capture cross section of 3.36mb at 0.0253 eV (ENDF/B-VII.0) was used to deduce the equivalent boron content of 1 ppmwt from the measured results. Thermal capture cross-section of carbon was significantly increased in the ENDF/B-VII.1 and other recent cross-section libraries. Therefore, when calculating the ASTRA benchmark with these cross-section libraries the equivalent boron content reported in the benchmark specification should be reassessed.

Determination of the Bi209(n,γ)Bi210g cross section using the NICE detector

The capture cross section of $^{209}$Bi(n,$\gamma$)$^{210g}$Bi was measured at different astrophysically energies including thermal capture cross section (25 meV), resonance integral, and the Maxwellian averaged cross section at a thermal energy of $kT$ = 30 keV. The partial capture cross section ($\sigma_g$) was determined using the activation technique and by measuring the $^{210}$Po activity. The newly developed and tested NICE detector setup was used to measure the $\alpha$-activity of the $^{210}$Po. Using this setup the thermal and resonance integral cross sections were determined to be $16.2\;\pm\;0.97$~mb and $89.81\;\pm\;8.0$~mb, respectively. And the Maxwellian average cross section was measured to be $2.01\;\pm\;0.38$~mb.

NICE - Neutron Induced Charged particle Emission

A new detector setup (NICE-detector) based on an organic plastic scintillator is presented. It will be used during experiments measuring neutron-induced reactions with a charged particle in the exit channel. The proposed design was tested at the Goethe University Frankfurt. One of the test cases was the capture cross-section of 209Bi at different astrophysically important energies, including thermal capture cross-section and resonance integral. This research presents the performance of the detector setup, as well as preliminary results for the calculated cross-sections. The preliminary results demonstrate that the newly developed NICE-detector can be used to determine capture cross-sections with sufficient accuracy, and might be adopted in the future as an alternative for charged particle measurements in different nuclear and astrophysical applications.

Measurement of neutron-proton capture in the SNO+ water phase

The SNO+ experiment collected data as a low-threshold water Cherenkov detector from September 2017 to July 2019. Measurements of the 2.2-MeV $\gamma$ produced by neutron capture on hydrogen have been made using an Am-Be calibration source, for which a large fraction of emitted neutrons are produced simultaneously with a 4.4-MeV $\gamma$. Analysis of the delayed coincidence between the 4.4-MeV $\gamma$ and the 2.2-MeV capture $\gamma$ revealed a neutron detection efficiency that is centered around 50% and varies at the level of 1% across the inner region of the detector, which to our knowledge is the highest efficiency achieved among pure water Cherenkov detectors. In addition, the neutron capture time constant was measured and converted to a thermal neutron-proton capture cross section of $336.3^{+1.2}_{-1.5}$ mb.

Validation of actinide nuclear data based on reactivity worth experiments in a MOX-LWR spectrum

OSMOSE was a CEA/DOE collaborative experiment dedicated to the improvement of actinide nuclear data. Based on reactivity worth measurements of pure isotopic samples of various actinides from 232Th to 245Cm, it represents a unique set experimental data to validate evaluated nuclear data.This paper presents the analysis of the experiments conducted in a MOXLWR-type core configuration, with TRIPOLI-4® Monte-Carlo calculations. The comparison between calculated and measured reactivity worth shows a good agreement with JEFF-3.1.1 for all the actinides, except 241Am. The combination of the C/E results with the sensitivity coefficients and prior covariances in a re-estimation process suggests that some improvement is required on the 241Am (n,g), towards an increase of +6% of the JEFF-3.1.1 thermal capture cross section. Based on the same reestimation process, improvements in the nuclear data covariances are pointed out, with posterior uncertainties on the reactivity worth significantly reduced for 238Pu, 239Pu, 241Am, 237Np and 244+245Cm.

Electroluminescence analysis of VOC degradation of individual subcell in GaInP/GaAs/Ge space solar cells irradiated by 1.0 MeV electrons

The open-circuit voltage (VOC) degradation for individual subcell in 1.0 MeV electrons-irradiated GaInP/GaAs/Ge solar cells has been analyzed using electroluminescence (EL) measurements. By using the relationship between VOC degradation and electron fluence, the capture cross section of the defects induced by electron irradiation in individual subcells were determined. Furthermore, by comparing the thermal activation energy and capture cross section, a defect located at Ev+0.55 ​eV styled H2 hole trap is nonradiative recombination centre in GaInP subcell, a defect located at EC-0.96 eV styled E5 electron trap is nonradiative recombination centre in GaAs subcell, and a defect located at EC-0.38 eV styled E-center electron trap is nonradiative recombination centre in Ge subcell.

Neutronic investigation of the thorium-based mixed-oxide as an alternative fuel in the TRIGA Mark-II research reactor – Part I: A beginning of life calculations

This article presents a neutronic investigation of the effect of replacing some uranium zirconium hydride (U-ZrH1.6) fuels with thorium-based mixed-oxide ((Th,U)O2) fuels in the TRIGA Mark-II research reactor without changing the core geometry, while the only variable is the fuel material.The possibility of utilizing thorium mixed oxide fuel in the TRIGA Mark-II reactor is assessed using the famous seed-blanket configuration or also known as Radkowsky Thorium Fuel (RTF) concept. Using the RTF concept, a long core lifetime can be achieved by enhancing the internal conversion rate of fertile to fissile materials, thanks to the thermal capture cross-section of 232Th that is around three times that of 238U. Additionally, on the fuel cycle aspect of using the thorium-based fuel is that the thorium cycle produces much less plutonium and other transuranic radioactive elements than uranium fuel cycle.Actually, neutronic and kinetic calculations, at the beginning of life conditions, for nine different seed-blanket configurations have been investigated using the MCNPX2.6 code. These calculations for all proposed configurations in comparison with the conventional U-ZrH1.6 fueled core were performed to pick out an optimal core(s) design ensuring the most efficient use of (Th,U)O2 fuel in it and with minimum changes in the main safety parameters. The parameters consist of effective multiplication factor keff, delayed neutron fraction βeff, neutron flux distribution, power distribution and radial hot rod power peaking factor (FHR), and shutdown margin (SDM).The calculation results exhibit that the highest initial criticality achieved is 1.07716. The results of the other neutronic investigations of thorium-based mixed-oxide fuel elements and their positions, including their numbers in the core are summarized and discussed in the following paper.

Improved 242Pu(n,$ \gamma$) thermal cross section combining activation and prompt gamma analysis

A good knowledge of the radiative capture cross section of 242Pu is required for innovative nuclear reactor studies, especially for MoX fuel reactors. However, the experimental data available show discrepancies in the energy regions of interest: the thermal point and the keV region. Previous experimental results of the thermal cross section deviate from each other by 20% and these discrepancies are reflected also in the evaluated libraries, each of them giving more credit to different data sets. A recent measurement by Genreith et al. did not succeed to solve the existing discrepancy due to the large uncertainties and correction factors in the analysis. This work presents a new measurement of the thermal capture cross section of 242Pu carried out in the Budapest Research Reactor using the same thin targets of a previous measurement at n_TOF-EAR1, each containing 30mg of 99.995% pure 242Pu . The combined analysis of the full prompt $ \gamma$-ray spectrum and the 243Pu decay has led to three compatible values for the thermal cross section. Their average value, 18.9(9)b, has an improved accuracy compared to recent measurements. Leaving aside the activation value of Genreith using an outdated intensity value for the 84 keV decay line of 243Pu , our average result is in very good agreement with the JEFF-3.2 evaluation and all the previous measurements, with the exception of the highest value 22.5(11)b reported by Marie et al., which has a strong influence in the ENDF evaluation.

Physical characteristics of a PWR core loaded with micro-cell UO2 pellet fuels

In this work, the physical characteristics of PWR cores are evaluated for several micro-cell UO2 pellets with Cr-coated Zircaloy-4 cladding through consistent fuel assembly (FA) and core-level analyses. The micro-cell UO2 pellets that are currently researched as accident tolerant fuel (ATF) candidates in the Korea Atomic Energy Research Institute (KAERI) are UO2-Cr, UO2-Mo, and UO2-SiO2-TiO2. In particular, a calculational matrix is used to consider not only the micro-cell UO2 pellets and the cladding coating, but also to estimate the achievable cycles within the usual 5.0 wt% uranium enrichment limit and the uranium enrichment required to achieve a reference cycle length of 480 effective full power days (EFPDs). The core analyses using the 3D nodal diffusion code MASTER showed that the cores loaded with micro-cell UO2 pellets (i.e., UO2-Cr and UO2-Mo) with Cr-alloy coating on Zircaloy-4 cladding have considerably shorter cycle lengths, lower discharge burnups, and more negative fuel temperature coefficients than the conventional UO2 fuel. The more negative fuel temperature coefficients lead to a more rapid stabilization under rod ejection accident with a substantially lower fuel temperature. In particular, the Mo-based micro-cell UO2 fueled cores have a much shorter cycle length and more negative moderator temperature coefficient than Cr-based one due to the large thermal and resonance capture cross sections of Mo isotopes and due to its lower critical boron concentration (CBC). However, the effect of Cr-alloy coating on core performance is not significant in comparison with the effects of the micro-cell UO2 pellets.

Assembly-level analyses of accident-tolerant cladding concepts for a long-life civil marine SMR core using micro-heterogeneous duplex fuel

In this reactor physics study, we examine the neutronic performance of accident-tolerant fuel (ATF) claddings – austenitic type 310 stainless steel (310SS), ferritic Fe-20Cr-5Al (FeCrAl), advanced powder metallurgic ferritic (APMT), and silicon carbide (SiC)-based materials – as alternative cladding materials compared with Zircaloy-4 (Zr) cladding. The cores considered use 18% 235U enriched micro-heterogeneous ThO2-UO2 duplex fuel and, for purposes of comparison, 15% 235U enriched homogeneously mixed all-UO2 fuel, loaded into 13×13 pin arrays. A constant cladding coating thickness of 655 μm is assumed. We use the WIMS reactor physics code to analyse the associated reactivity, achievable discharge burnup, spectral variations, rim effect and reactivity feedback parameters for the candidate cladding materials at the assembly level.The results show that candidate fuels with 310SS cladding exhibit a ∼13% discharge burnup penalty compared to Zr due to the presence of a very high nickel (Ni) concentration. The high neutron absorption cross-sections of iron (Fe) in the FeCrAl and APMT claddings also lead to a ∼10% discharge burnup penalty. The fuels with SiC cladding can achieve a ∼1% higher discharge burnup compared to Zr due to the low thermal neutron absorption cross-sections of its constituents and the softer neutron spectrum. The claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. For both candidate fuels, the end-of-life 239Pu (for UO2 fuel) and 233U (for duplex fuel) inventories are higher for the claddings (Fe-based: FeCrAl, APMT and steel-based: 310SS) with higher thermal capture cross-sections, unlike for SiC and Zr, where SiC provides higher end-of-life 239Pu and 233U inventories despite having lower capture cross-section than that of the Zr. Reactivity feedback parameter values (moderator and fuel temperature coefficients) are more negative for the duplex fuel than the UO2 fuel for all the candidate claddings, with claddings with harder spectra exhibiting more negative values. The duplex fuel yields a softer spectrum than the UO2 fuel with the candidate claddings, which improves neutron economy and thus discharge burnup.

Neutron Transmission and Capture Measurements of 133Cs from 0.01 to 600 eV

Neutron capture and transmission measurements were carried out from 0.01 to 600 eV on both solid and liquid samples containing elemental cesium (133Cs). Only s-wave resonances were observed in these measurements. These data were analyzed for resonance parameters utilizing the SAMMY Bayesian analysis code to simultaneously fit both the capture and transmission data. Parameters were obtained for 31 cesium resonances up to 600 eV. The thermal capture cross section and capture resonance integral were determined. The thermal capture cross section is 10% larger than the ENDF, JENDL, and JEFF evaluated values but lies within the uncertainty of the most recent measurement by Yoon and Lee [New Phys.: Sae Mulli (Korean Phys. Soc.)., Vol. 61, p. 7 (2011)]. The capture resonance integral has a statistical 1σ error of 2% and lies 1.4σ above the JENDL value, 5.5σ above the ENDF value, and 3.9σ above the JEFF value. The s-wave strength function was determined.

Measurements of neutron total and capture cross sections of 241Am with ANNRI at J-PARC

ABSTRACT Neutron total and capture cross sections of 241Am have been measured with a new data acquisition system and a new neutron transmission measurement system installed in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex. The neutron total cross sections of 241Am were determined by using a neutron time-of-flight (TOF) method in the neutron energy region from 4 meV to 2 eV. The thermal total cross section of 241Am was derived with an uncertainty of 2.9%. A pulse-height weighting technique was applied to determine neutron capture yields of 241Am. The neutron capture cross sections were determined by the TOF method in the neutron energy region from the thermal to 100 eV, and the thermal capture cross section was obtained with an uncertainty of 4.1%. The evaluation data of JENDL-4.0 and JEFF-3.2 were compared with the present results.

Resonance analysis of 151,153Eu from neutron capture cross section measurements in the energy range from 1 to 20 eV

ABSTRACT The neutron capture cross sections of Europium-151 and Europium-153 have been measured by the time-of-flight method in the energy range from 0.005 to 100 eV using the Kyoto University Research Reactor Institute-Linear Accelerator (KURRI-LINAC). An assembly of Bismuth Germanate (BGO) scintillators was used to detect the prompt capture of γ rays. The absolute values of the neutron capture cross sections of 151Eu and 153Eu were deduced by normalizing the thermal capture cross sections in JENDL-4.0 and ENDF/B-VII.1, respectively. Then, we have obtained the resonance parameters of 20 resonances in 151Eu and 17 resonances in 153Eu using the code SAMMY. For the 3.36-eV resonance of 151Eu, the evaluated resonance peak area in JENDL-4.0 is about 95% smaller than the present result. For the 7.00-, 7.22-, and 7.42-eV resonance; we confirmed that there are significant differences between the measured peaks and evaluated peaks in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2. For the 153Eu, the evaluated resonance peak areas in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2 are about 15% larger than the measured resonance peak areas at the 2.46-, 3.29-, and 3.94-eV resonances.

Rhenium resonance parameters from neutron capture and transmission measurements in the energy range 0.01 eV to 1 keV

Rhenium is a refractory metal with potential uses in nuclear reactor applications, particularly those at very high temperatures. Measurements have been made using natural samples. Natural rhenium consists of two isotopes: 185Re (37.40%) and 187Re (62.60%). The electron linear accelerator (LINAC) at the Rensselaer Polytechnic Institute (RPI) Gaerttner LINAC Center was used to explore neutron interactions with rhenium in the energy region from 0.01 eV to 1 keV. Neutron capture and transmission measurements were performed by the time-of-flight technique. Two transmission measurements were performed at flight paths of 15 m and 25 m with 6Li glass scintillation detectors. The neutron capture measurements were performed at a flight path of 25 m with a 16-segment sodium iodide multiplicity detector. Resonance parameters were extracted from the data using the multilevel R-matrix Bayesian code SAMMY. A table of resonance parameters and their uncertainties is presented. The uncertainties in resonance parameters were propagated from a number of experimental quantities using a Bayesian analysis. Uncertainties were also estimated from fitting each Re sample measurement individually. The measured neutron capture resonance integral for 185Re is (4 ± 1)% larger than ENDF/B-VII.1. The capture resonance integral for 187Re is (3 ± 1)% larger than ENDF/B-VII.1. Other findings from these measurements include: a decrease in the thermal capture cross section for 185Re of (2 ± 2)% from ENDF/B-VII.1; a decrease in the thermal capture cross section for 187Re of (3 ± 4)% from ENDF/B-VII.1; a decrease in the thermal total cross section for 185Re of (2 ± 2)% from ENDF/B-VII.1; and a decrease in the thermal total cross section for 187Re of (6 ± 5)% from ENDF/B-VII.1. Considering the uncertainties, none of the indicated changes in thermal cross sections represents a statistically significant change from ENDF/B-VII.1.

Thermal neutron capture cross-section and resonance integral measurements of 139La(n, $ \gamma$ γ )140La and 140Ce(n, $ \gamma$ γ )141Ce using a Am-Be neutron source

Thermal neutron capture cross-sections and resonance integrals of 139La(n,\( \gamma\))140La and 140Ce (n,\( \gamma\))141Ce are measured with respect to reference reactions 197Au(n,\( \gamma\))198Au and 55Mn(n,\( \gamma\))56Mn using the neutron activation technique. Measurements are carried out using neutrons from an Am-Be source located inside a concrete bunker. Two different methods are used for determining self-shielding factors of activation foils as well as for finding the epithermal neutron spectrum shape factor. For 139 La with reference to 197 Au and 55 Mn the measured thermal cross sections are \( 9.24 \pm 0.25\) b and \( 9.28 \pm 0.37\) b, respectively, while the measured resonance integrals are \( 12.18 \pm 0.67\) b and \( 11.81 \pm 0.94\) b, respectively. For 140 Ce with reference to 197 Au and 55 Mn the measured thermal cross sections are \( 0.44 \pm 0.01\) b and \( 0.44 \pm 0.02\) b, respectively, while the measured resonance integrals are \( 0.55 \pm 0.03\) b and \( 0.54 \pm 0.04\) b, respectively. The present measurements are compared with earlier measurements and evaluations. Presently estimated values confirm the established 139La(n,\( \gamma\))140La cross-sections. The presently measured thermal capture cross-section 140Ce(n,\( \gamma\))141Ce , though lower than the evaluated data, is having higher accuracy compared to previous measurements with large uncertainties. The resonance integral measured is higher (like most previous measurements) than most evaluations requiring a revision of the evaluated data.

Neutron transmission and capture measurements and analysis of Dy from 0.01 to 550 eV

Neutron capture and transmission measurements were carried out from 0.01 to over 600 eV on both natural Dy and samples highly enriched in 164Dy. These data were analyzed for resonance parameters utilizing the SAMMY Bayesian analysis code to simultaneously fit both the capture and transmission data. Parameters were obtained for 17 resonances in Dy isotopes up to 18 eV and for the 164Dy resonances near 147, 450 and 540 eV. The thermal capture cross section (at 0.0253 eV) and capture resonance integral were determined for 164Dy.

Thermal-neutron cross sections and resonance integrals of 138Ba and 141Pr using Am-Be neutron source

The thermal-neutron capture cross sections and resonance integrals of 138Ba(n,\( \gamma\))139Ba and 141Pr(n,\( \gamma\))142Pr were measured by activation method using an isotopic Am-Be neutron source. The estimations were with respect to that of 55Mn(n,\( \gamma\))56Mn and 197Au(n,\( \gamma\))198Au reference monitors. The measured thermal-capture cross section of 138 Ba with respect to 55 Mn is 0.410±0.023 b and with respect to 197 Au is 0.386±0.019 b. The measured thermal-capture cross section of 141 Pr with respect to 55 Mn is 11.36±1.29 b and with respect to 197 Au is 10.43±1.14 b. The resonance integrals for 138 Ba are 0.380±0.033 b (55 Mn) and 0.364±0.027 b (197 Au) and for 141 Pr are 21.05±2.88 b (55 Mn) and 15.27±1.87 b (197 Au). The comparison between the present measurements and various reported values are discussed. The cross sections corresponding to the selected isotopes are measured using an Am-Be source facility for the first time.

Correction of the thermal neutron capture cross section of 241Am obtained by the Westcott convention

ABSTRACTThere is large discrepancy among the reported experimental data of the thermal neutron capture cross section of 241Am, where the activation measurements provided larger cross sections than those in the time-of-flight ones. The Westcott convention has been widely used for the derivation of the thermal neutron capture cross section in the activation measurements. We have estimated that this large discrepancy is due to the existence of the resonances below the cadmium cut-off energy (ECd ∼ 0.5 eV). By reviewing the Westcott convention, we developed the correction method taking account of the contribution of the resonances near or below ECd. The correction term was evaluated using the JENDL-4.0. Application of the present method successfully improved the existing discrepancy of the thermal capture cross section of 241Am.