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Ta 181 ( n , γ ) cross section and average resonance parameter measurements in the unresolved resonance region from 24 to 1180 keV using a filtered-beam technique
Abstract
A new array of four Deuterated Benzene (C6D6) detectors has been installed at the Gaerttner Linear Accelerator Center at Rensselaer Polytechnic Institute for the purpose of measuring neutron capture cross sections in the keV region. Measurements were performed on samples of Ta181 in the unresolved resonance region (URR) using a filtered-beam technique, by which a 30 cm iron filter was placed in a white-spectrum neutron beam to remove all time-dependent γ-ray background and all neutrons except those transmitted through resonance-potential interference “windows” in the iron. The resulting filtered beam was effectively a quasimonoenergetic neutron source, which was used for performing measurements on isotopes with narrow level spacings in the URR. The capture cross-section results obtained for two thicknesses of tantalum are in agreement with those documented in the JEFF-3.2 library, as are the average resonance parameters obtained via a fit to the data using the sammy-fitacs code.
... Monitor reactions are commonly used in offline measurements and provide a measurement of the overall irradiation fluence. In the thermal region, 197 Au and 94 Zr are common monitors [6,7,23], and in the RRR smooth cross sections like 10 B(n, α) and 6 Li(n, t) are used [22,24]. Monitor reactions can also be used in the HER, such as 56 Fe(n, n ) [25]. ...
... A schematic of a TED detector setup is shown in Figure 3b. C 6 D 6 detector systems have been used at many facilities, including RPI [24], the Geel Electron LINear Accelerator (GELINA) [57], CERN [58], KURRI [59], and ORELA [60], and NaI systems have been used at J-PARC [61]. ...
... Flux normalization. Normalization of the count rates into capture yields is similar to TAS measurements, often using a saturated resonance in the isotope being measured [24] or in 197 Au [67], or by simultaneous fitting of transmission and capture yield data [51]. ...
This paper provides a template of expected uncertainties and correlations for measurements of neutron-induced capture and charged-particle production cross sections. Measurements performed in-beam include total absorption spectroscopy, total energy detection, γ-ray spectroscopy, and direct charged-particle detection. Offline measurements include activation analysis and accelerator mass spectrometry. The information needed for proper use of the datasets in resonance region and high energy region evaluations is described, and recommended uncertainties are provided when specific values are not available for a dataset.
... 180 Ta is produced by two minor branchings in the s-process along the stable hafnium isotopes which is discussed by Kappeler et al. 5 and Malatji et al. 6 . 181 Ta is produced by the s-process, its ( n, γ ) cross sections and MACS at 30 keV are of great significance in nuclear astrophysics for understanding the reaction path of the s-process 7,8 . However, according to the EXFOR library, high-precision, continuous measurements of capture cross sections in the resolved resonance region are not sufficient. ...
The neutron capture cross section of [Formula: see text]Ta is relevant to s-process of nuclear astrophysics, extraterrestrial samples analysis in planetary geology and new generation nuclear energy system design. The [Formula: see text]Ta([Formula: see text]) cross section had been measured between 1 eV and 800 keV at the back-streaming white neutron facility (Back-n) of China spallation neutron source(CSNS) using the time-of-flight (TOF) technique and [Formula: see text] liquid scintillator detectors. The experimental results are compared with the data of several evaluated libraries and previous experiments in the resolved and unresolved resonance region. Resonance parameters are extracted using the R-Matrix code SAMMY in the 1-700 eV region. The astrophysical Maxwell average cross section(MACS) from kT = 5 to 100 keV is calculated over a sufficiently wide range of neutron energies. For the characteristic thermal energy of an astrophysical site, at kT = 30keV the MACS value of [Formula: see text]Ta is 834 ± 75 mb, which shows an obvious discrepancy with the Karlsruhe Astrophysical Database of Nucleosynthesis in Stars (KADoNiS) recommended value 766 ± 15 mb. The new measurements strongly constrain the MACS of [Formula: see text]Ta([Formula: see text]) reaction in the stellar s-process temperatures.
... The 54 Fe(n,p), 54 Fe(n,a) and 89 Y(n,2n) were selected because all are in IRDFF (International Reactor Dosimetry and Fusion Files) nuclear data library and moreover, iron is an important construction material. The 181 Ta(n,g) was selected due to reported discrepancies in 181 Ta cross section (McDermott et al., 2017). ...
Validation of the selected materials cross sections is an important task for their use as flux monitors within experimental measurements. As the reaction rate is a function of neutron spectrum and material cross section, the neutron spectrum must be well defined for the correct cross section validation. This criterion is met in the special core of the LR-0 reactor, where the reaction rates of 181 Ta(n,c), 54 Fe(n,p), 54 Fe (n,a) and 89 Y(n,2n) reactions were measured. The 197 Au(n,c) and 58 Ni(n,p) reactions were used as flux monitors during measurements. Results of calculation show that in case of 181 Ta(n,c) reaction, the values reported in JEFF-3.2 provide a good agreement, while ENDF/B-VII.1 is underpredicting the experiment. In case of 54 Fe(n,p) reaction, JEFF-3.2 is in satisfactory agreement, while ENDF/B-VII.1 overpredicts the result by about 10%. The 54 Fe(n,a) shows significant discrepancies, JEFF-3.2 under predicts by about 20%, while ENDF/B-VII.1 overpredicts result by 40%. The high threshold 89 Y(n,2n) reaction is interesting from the point of view of spectral averaged cross section. It is worth noting that the resulted spectral averaged cross section, being 0.169 with 1r uncertainty of 0.007 mb, is in very good agreement with previous result, being 0.172 ± 0.006 mb.
Cr-filtered keV-neutron experiments were performed in the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) beamline in the Materials and Life Science (MLF) facility of the Japan Proton Accelerator Research Complex (J-PARC) to measure the neutron capture cross-section of ¹⁹⁷Au. The energy range of the neutron filtering system at ANNRI was extended through the use of 15 cm of natCr as filter material to tailor quasi-monochromatic neutron peaks with averaged neutron energies of 133.4 and 45.0 keV. The performance of the natCr filter assembly was evaluated by means of experimental capture and transmission analyses, together with the use of Monte-Carlo simulations. The present ¹⁹⁷Au neutron capture cross-section results provide agreement within uncertainties with the JENDL-4.0 evaluated library and the IAEA standard data library further demonstrating the capabilities of the neutron filtering system at ANNRI.
A neutron filtering system has been introduced in the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) beamline in the Materials and Life Science (MLF) facility of the Japan Proton Accelerator Research Complex (J-PARC) in order to produce quasi-monoenergetic neutron beams. The filtered neutron spectrum by the filter assemblies was analyzed by means of capture and transmission measurements and also by Monte Carlo simulations using PHITS. The characteristics of the filtered neutron beam are discussed alongside its viability in future applications for neutron cross-section measurements in the fast neutron region.
New evaluated data file for Ta-181 irradiated with neutrons at energies up to 200 MeV has been prepared.
The data evaluation has been done using the results of calculations, measured data, systematics predictions, and covariance information. Calculations have been performed using a special version of the TALYS code implementing the geometry dependent hybrid model and models for the non-equilibrium light cluster emission.
The TEFAL code and the FOX code from the BEKED package have been used for the formatting of the data.
Often discrepancies can be found in the corresponding cross sections of different evaluated nuclear data libraries. Traditional integral benchmarks that are used to validate such libraries are sensitive to cross-section values across many different energies. This means an erroneously low cross section at one energy may compensate for an erroneously high cross section at another energy, and the integral benchmark value may still be met. While the evaluated cross section may agree with that single benchmark, it could affect other systems differently. To reduce the potential for this error, an energy differential validation method is proposed herein for continuous energy Monte Carlo neutron transport models in the resolved resonance region and the unresolved resonance region (URR). The proposed method exposes the underlying physics of the URR and validates both the average cross section and resonance self-shielding effect driven by the fluctuations in that cross section. This is done by measuring the neutron transmission of a thick sample that, by its nature, exaggerates the resonance self-shielding effect. This validation method is shown to be very sensitive to the cross-section model used (resolved versus unresolved) and the fluctuation correction employed, allowing it to probe the validity of the previously mentioned cross-section evaluations. Tantalum-181 is used as an example to demonstrate the impact of different resonance evaluations. It was found that the JEFF-3.3 and JENDL-4.0u evaluations made reasonable choices for cross-section models of ¹⁸¹Ta; none of the current evaluations, however, can be used to properly model the validation transmission over all energies. It was also found that updating resonance parameters in the URR provided better agreement with the validation transmission.
Purpose: To estimate the contribution of the secondary neutron flux to the total radiation flux during the operation of Trilogy linear medical accelerator and Varian’s Clinac 2100 accelerator for assessment of impact on the health of patients and medical personnel. High-energy linear accelerators operating at energies higher than 8 MeV generate neutron fluxes when interacting with accelerator elements and with structural materials of the room for treating patients. Neutrons can form at the accelerator head (target, collimators, smoothing filter, etc.), the procedure room, and directly in the patient’s body. Because of the high radiobiological hazard of neutron radiation, its contribution to the total beam flux, even at a level of few percent, substantially increases the dose received by the patient. Material and methods: Secondary neutron fluxes were investigated during the process of the linear medical accelerators Trilogy and Clinac 2100 of Varian operation by the photoactivation method using (γ, n) and (n, γ) reactions on the detection target of natural 181Ta. In addition, measurements of neutron spectra were carried out directly in the room during the operation of a medical accelerator using a spectrometer-dosimeter SDMF-1608. Results: It was determined that the neutron flux on the tantalum target is 16 % of the gamma-ray flux on the same target when the accelerator is operated with a 18 MeV bremsstrahlung energy and 5 % when the accelerator is operated with a 20 MeV excluding thermal neutrons. Conclusion: Finally, it may be noted that, taking into account the coefficient of relative biological efficiency (RBE) of neutron radiation for neutrons with energies of 0.1–200 keV equal to 10 compared with the RBE coefficient for gamma quanta (equal to 1), even preliminary analysis demonstrates significant underestimation of the contribution of neutrons dose to the total dose received by the patient in radiation therapy using bremsstrahlung of 18 and 20 MeV.
Flows of secondary electrons are studied using a medical electron accelerator with an energy of 20 MeV. The electrons are generated on the structural materials of the accelerator itself and in the treatment rooms. The experiments are based on means of activation based on (γ, n) and (n, γ) reactions on a detecting target of natural ¹⁸¹Ta tantalum. The irradiated tantalum targets are probed using a spectrometer with a large-volume ultrapure germanium detector. The energy distribution of neutrons is obtained using a spectrometer–dosimeter equipped with an organic scintillator. It is established that the flow of neutrons accounts for 7% of the flow of braking gamma-quanta.
Due to the complexity of nuclear reaction models, current nuclear data evaluations must rely on experimental observations to constrain models and provide the accuracy needed for applications. For criticality applications, the accuracy of nuclear data needed is higher than what is currently possible from differential experiments alone, and integral measurements are often used for data adjustment within the uncertainties of differential experiments. This approach does not necessarily result in physically correct cross sections or other adjusted quantities because compensation between different materials is hard to avoid. One of the objectives of the recent CIELO project [M. Chadwick et al., Nucl. Data Sheets 118, 1 (2014)] was simultaneous evaluation of important materials in an attempt to minimize the effects of compensation. Improvement to the evaluation process depends on obtaining new experimental data with high accuracy and lower uncertainty that will help constrain the evaluations for certain important reactions. Improved experiments are accomplished by careful design with the objective of achieving high accuracy and lower uncertainty, and by designing new innovative experiments. New and unconventional experiments do not necessarily provide differential data but instead nuclear data that evaluators will find useful to constrain the evaluation and reduce the uncertainty. This also means that closer information exchange and collaboration between experimentalists and evaluators is important. For conventional experiments such as neutron transmission or capture measurements, it is important to understand the sources of uncertainty and address them in the experiment design. Such a process can also lead to the design of innovative methods. For example, the filtered beam method minimizes uncertainties due to background, and the Quasi-Differential Neutron Scattering method simplifies the experiment and data analysis and results in lower experimental uncertainty. A review of the sources of uncertainty in various experiments and examples of experimental techniques that help reduce experimental and evaluation uncertainty and increase accuracy will be discussed.
The National Aeronautics and Space Administration is considering nuclear power sources for space exploration. A series of critical mass experiments was designed to address the development, performance, and design of a space nuclear reactor being considered to support the Prometheus project. These experiments consisted of interlacing the refractory metals rhenium (Re), molybdenum (Mo), tantalum2.5 wt% tungsten (Ta-2.5W), and niobium-1 wt% zirconium (Nb-1Zr) with moderating materials (graphite or polyethylene) and were fueled by highly enriched uranium plates. These experiments are designed to assess the adequacy of and uncertainty in refractory metal neutron cross-section evaluations for use in Prometheus nuclear reactor design work. The critical experiments were designed in the energy spectrum closely resembling or bracketing that in the proposed space reactor. For each material (Re, Mo, Ta-2.5W and Nb-1Zr), four critical configurations were designed and performed to measure the sensitivity of keff to the material under four different and progressively softer neutron spectra (core center spectrum, harder than core average spectrum, softer than core average spectrum, and accident flooded spectrum). The thicknesses of the graphite or polyethylene moderator and reflector plates were adjusted to achieve the desired neutron spectrum. One critical and 18 subcritical experiments provided for measurements of material neutronic behavior in a simple cylindrical geometry configuration that was modeled in MCNP with ENDF/B-VI.6 cross-section data and compared to the extrapolated or predicted critical mass for all the experiments. These experiments were performed at the Los Alamos National Laboratory using the Planet vertical lift critical assembly at the Los Alamos Critical Experiment Facility.
The neutron capture cross sections of Nb, In, I, Ho, Ta, Th and U were measured using the Fe-filtered beam. A 15-cm thick Fe filter was placed in the neutron beam produced by the KUR 46-MeV electron Linac and capture prays were detected by two C6F6 scintillation detectors located at an 11.7 m-flight path. The pulse-height weighting technique was used to determine the relative capture pray detection efficiency. The neutron flux was measured by the same detectors, whose detection efficiency for the 480-keV pray from the B(n, α1) reaction was calibrated by the saturated resonance capture in Ag at 5.2-eV. Self-shielding and multiple scattering corrections were applied to the data. The results of 24-keV capture cross sections are 340, 770, 780, 1,280, 880, 520 and 520 mb for Nb, In, I, Ho, Ta, Th and U, respectively. Total errors are 5 to 8%, with an estimated systematic error of 4%. The discrepancy between the present results and other data measured recently is within 10%.
The biennial review of atomic-weight determinations and other cognate data has resulted in changes for the standard atomic weights of five elements. The atomic weight of bromine has changed from 79.904(1) to the interval [79.901, 79.907], germanium from 72.63(1) to 72.630(8), indium from 114.818(3) to 114.818(1), magnesium from 24.3050(6) to the interval [24.304, 24.307], and mercury from 200.59(2) to 200.592(3). For bromine and magnesium, assignment of intervals for the new standard atomic weights reflects the common occurrence of variations in the atomic weights of those elements in normal terrestrial materials.
This paper presents new measurements of the carbon and beryllium neutron total cross section in the energy range of 24 to 950 keV. The measurements were done using a pulsed neutron source driven by an electron LINAC. The neutron beam passed through a 30-cm-thick iron filter, which results in neutron transmission only in energies where resonance scattering and potential interference exist. The neutron filter removes most of the neutrons at other energies and significantly attenuates the gamma background resulting in 20 energy windows and a high signal-to-background ratio. The filtered beam was used for transmission measurements through graphite that results in ;1% accurate total cross sections that are in excellent agreement with current evaluations. The carbon measurement provides a verification of the accuracy of the filtered beam method. Measurements of three samples of different thicknesses of beryllium resulted in accurate total cross-section values that agree with one previous measurement and show dis-crepancies from current evaluations. The high accuracy of the new measurements can be used for improve-ment of future total cross-section evaluations of beryllium.
Accurate isotopic molybdenum nuclear data are important because molybdenum can exist in nuclear reactor components including fuel, cladding, or as a high yield fission product. High-resolution time-of-flight neutron transmission measurements on highly enriched isotopic metallic samples of Mo95, Mo96, Mo98, and Mo100 were performed in the resonance energy range from 1 to 620keV. The measurements were taken with the newly developed modular Li6-glass transmission detector positioned at the 100-m experimental flight station. In the unresolved energy region (URR), new comprehensive methods of analysis were developed and validated in order to obtain accurate neutron total cross-section data from the measurement by correcting for background and transmission enhancement effects. Average parameters and fits to the total cross section for Mo95 were obtained using the Hauser-Feshbach statistical model code fitacs, which is currently incorporated into the sammy code. The fits to the experimental data deviate from the current evaluated nuclear data file/B-VII.1 isotopic Mo evaluations by several percent in the URR. © 2015 American Physical Society.
The total cross section and the differential cross section for the inelastic scattering of neutrons are considered. It is assumed that the compound nucleus is sufficiently excited so that the statistical model may be applied. If the statistical model may be applied as well to the residual nucleus, it is shown that the angular distribution of the inelastically scattered neutrons is isotropic. If only a few levels of the target nucleus can be excited, the angular distribution is anisotropic. Tables are provided which permit the calculation of the angular distribution if the incident and emergent neutron angular momenta are less than or equal to 3ℏ. Examples of the evaluation of total cross sections are given, providing examples of the sensitivity of the results to the quantum numbers of the excited state.
Neutron-radiative-capture cross sections have been measured at various energies from 30-220 keV using a new "total-energy-detector" technique of high sensitivity, applicable to small samples of separated stable isotopes. Samples studied are: V, Fe, Ni, 86Sr, 87Sr, Y, Nb, Rh, Ag, 122Te, 123Te, 124Te, 125Te, 126Te, 128Te, 130Te, I, Eu, 175Lu, 176Lu, W, Au, 204Pb, and 208Pb. The odd-odd isotope 176Lu shows a substantially higher cross section than those of neighboring odd-even and even-odd nuclei. The results for europium (2 b near 65 keV) emphasize the peak expected from previous studies of systematics. The Sr and Te isotopic results strongly confirm the predictions of s-process stellar nucleosynthesis.
Neutron capture by a tantalum sample was measured at the Oak Ridge Electron Linear Accelerator pulsed neutron time-of-flight facility on a 40-m flight path. Average cross sections for /sup 181/Ta(n,..gamma..) in the energy range from 2.6 to 1900 keV were derived. The partially resolved region from 2620 to 4000 eV was fitted in terms of resonance parameters by least-squares adjustment.
Hydrogen-free deuterated benzene C6D6 liquid scintillators are widely used for high resolution neutron capture cross-section measurements at time-of-flight facilities, using the total energy detection principle in combination with the Pulse Height Weighting Technique (PHWT). The quality of the data deduced from such measurements depends on the accuracy of the detector response that is used in the calculation of the weighting function and on the normalization procedure. In addition, for nuclei with small capture to scattering ratios, i.e. light and near neutron magic nuclei, a correction for the sensitivity of the capture detector to the scattered neutrons is required. The MCNP code was used to simulate both the γ-ray and neutron transport in a C6D6 detection system including its surroundings. The weighting functions and neutron sensitivity were then deduced from the simulations and validated by experiments. The simulations have also been used to identify the sources of uncertainties in performing capture cross-section measurements with C6D6 detectors.
A new simulation of nuclear γ cascades by the Monte Carlo method is described. It makes it possible to generate artificially individual events of the γ-cascade decay of an isolated, highly excited initial level in a medium and heavy nucleus. A broad class of quantities, associated with the process of γ-cascade de-excitation, can be modelled. The main advantage of the method is the possibility of a full quantitative control over the influence of the Porter–Thomas fluctuations of partial radiation widths on uncertainties of the modelled cascade-related quantities. For assessment of these uncertainties and a control over the accuracy of the method, a special statistical formalism has been developed.
A scintillation gamma ray detector is described whose efficiency for detecting neutron capture events depends only on the total gamma ray energy released. The detector is cheap and fast; it has a low background and a low sensitivity to neutrons in the energy range studied (up to ∼ 100 keV).A description is given of the use of the detector with the Harwell Electron Linac neutron time of flight system, and results are presented of capture cross section measurements on rhodium, silver, indium, tantalum and gold, over the energy range 100 eV to 50 keV.
A new evaluation of the neutron cross section standards was recently completed. The results of this work became the standards for the new ENDF/B-VII library. This evaluation was performed to include new experiments on the standards that have been made since the ENDF/B-VI standards evaluation was completed and to improve the evaluation process. The evaluations include the H(n,n), 6Li(n,t), 10B(n,α), 10B(n,α1γ), 197Au(n,γ), 235U(n,f), and 238U(n,f) standard reactions. Evaluations were not done for the 3He(n,p) and C(n,n) standards which were carried over from ENDF/B-VI. The need for standards above 20 MeV led to the extension of the 235U(n,f) and 238U(n,f) cross sections to 200 MeV. In addition evaluations were produced for the non-standard 238U(n,γ) and 239Pu(n,f) reactions. Results of this work and comparisons with previous evaluations are presented.