Fission Cross Sections Research Articles

New Method for 18FDG generating using Geant4/ Gate7

Fluorine 18-deoxyglucose is often used in Positron Emission Tomography devices. Positron Emission Tomography imaging is one of the useful tool which is used for cancer detection and its management. Positron Emission Tomography growth is limited due to problems that depend on the production of Fluorine-18. Imaging results are strongly depending on the information of nuclear reaction cross section data. This study is presented to calculate different quantities such as stopping power, CSDA range, simulated and distributed absorbed dose of Fluorine-18 in water. In order to access these goals, we use Geant4/Gate7 simulation and the Bethe-Bloch model. The results of this simulation and theoretical model presented are in good agreement with each other. The important point of this paper is the presentation of a theoretical approach in order to the production of Fluorine-18 using protons generated through the main D(d;p)Tand side 3 He(d;p)4 He nuclear fusion reaction in which uses Helium-3 is catalyzed.

Positron spectroscopy of γ-flashes in the earth atmoshear

The registration of the fast thermonuclear transformation in Earth’s atmosphere is considered. One of them is γ-flashes from thunder lightning. The atmospheric chemical composition and bound conditions in the thunder lightning plasma tubes were associated with thermonuclear transformation. Nuclear reaction channel cross-sections were selected for the calculation of the γ-ray spectra. It was found that under existing conditions the γ-spectrum is formed due to the formation of unstable proton-excess nuclei with their further decay. It has been shown that electron-atom collisions are important and only initiate the formation of proton and fully ionized He (after sufficient acceleration, we call them α-particles) current flows. In this case, protons and α-particles initiate nuclear transformation after collisions with atmospheric He, C, N, O, and Ne atoms. The time scale of γ-fluorescence is estimated and related to observation with traditional γ-ray detectors. Separately calculated contribution of pollution admixture from dust into γ-spectra.

Optimisation of the Tripathi model using a nuclear reaction cross-section database

Nuclear reaction cross-sections are an essential ingredient to reliable deterministic and stochastic radiation transport codes used for radiation protection in space and heavy-ion therapy applications. A recent study compared the existing literature data compiled within the open-access GSI-ESA-NASA cross-section database to the models implemented in the transport codes most commonly used for radiation protection in space and heavy-ion therapy applications. The outcome of the comparison was that none of the models fit well the experimental data for all projectile-target systems at all energy ranges. Therefore, the literature data were exploited to optimise the Tripathi–Cucinotta–Wilson model as reported in this work. This model is used as default in FLUKA, TRiP, and SpaceTRiP, it is part of the hybrid-Kurotama (HK) model used in particle and heavy ion transport code (PHITS), and it is implemented in Geant4. The consequences of using the proposed Tripathi–Cucinotta–Wilson optimisation in the HK model are also analysed.

Assessing Criteria to Pick Ideal Moderators for Nuclear Fission Reactors

The goal of this study is to find the best and most feasible compound(s) to use in a nuclear reactor as a moderator in order to achieve a chain reaction with maximum efficiency. To obtain a chain reaction with the most energy output per volume material used, neutrons need to be able to collide with radioactive material in a manner that will maximize the probability of fission. For this to happen, the cross section for fission must be exceptionally large. This is attainable by slowing down the speed of neutrons by use of a moderator. A number of criteria must be met for a moderator to be considered the most feasible. For one, it must be able to quickly thermalize neutrons from the MeV range down to a few eV. In this paper, we will define thermalization as reducing a neutron's energy from 2 MeV to 0.025 eV [3]. Second, it mustn’t have a high affinity for absorbing neutrons. And lastly, it must be cheap and abundant. If these criteria are met, one has found a good moderator. This study will primarily explore two substances: Zirconium Hydride and Yttrium Hydride, and their abilities to act as moderators for slow water reactors.

EXFOR-based simultaneous evaluation for neutron-induced fission cross section of plutonium-242

ABSTRACT The 242Pu neutron-induced fission cross section was evaluated from 100 keV to 200 MeV. The experimental 242Pu and 235U fission cross sections and their ratios in the EXFOR library were reviewed and analysed by the least-squares method. Additional simultaneous evaluation was performed by including the experimental database of the 233,238U and 239,240,241Pu fission cross sections and their ratios developed for JENDL-5 evaluation. The 242Pu fission cross sections from our evaluation and JENDL-5 evaluation are close to each other below 1 MeV while they systematically differ from each other above 10 MeV. The cross section from our evaluation is systematically lower than the JENDL-4.0 cross section in the prompt fission neutron spectrum peak region (∼5% lower around 1 MeV). The newly evaluated 242Pu fission cross section was verified against the cross section measured in the 252Cf spontaneous fission neutron field and criticalities of small-sized LANL fast systems and demonstrated better performance than the JENDL-4.0 cross section on the same level with the JENDL-5 cross section.

Nuclear Level Density Effect for Induced Fission Cross-section Reactions on Some Even and Odd-A Americium Nuclei

Nuclear Level Density Effect for Induced Fission Cross-section Reactions on Some Even and Odd-A Americium Nuclei

Measurement of cross section for the 232Th(n, f) reaction using a time projection chamber

Accurate cross sections of neutron induced fission reactions are needed in the development of nuclear science and technology. The time projection chamber (TPC) is viewed as the potential detector to improve the measurement accuracy of the fission cross sections based on its track reconstruction and particle identification capabilities. The TPC developed by the back-streaming white neutron source (Back-n) team of China Spallation Neutron Source (CSNS) was introduced to verify its ability in measuring the cross section of fission reaction. Using our TPC, and based on the mono-energetic neutron source at Peking University and 232Th/238U samples, we have measured the cross section of the 232Th(n, f) reaction at 5.0 MeV. The fission events have been well distinguished from the background alpha events. The present cross section is consistent with the evaluation data from different libraries, which indicates that the fission cross section measurement using our TPC is feasible. By reducing the distance between the two fission samples, the large uncertainty of the present result will be reduced in future experiments.

The multigroup library processing method for coupled neutron and photon heating calculation of fast reactor

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

Manipulation of Giant Multipole Resonances via Vortex γ Photons.

Traditional photonuclear reactions primarily excite giant dipole resonances, making the measurement of isovector giant resonances with higher multipolarities a great challenge. In this Letter, the manipulation of collective excitations of different multipole transitions in even-even nuclei via vortex γ photons is investigated. We develop the calculation method for photonuclear cross sections induced by the vortex γ photon beam using the fully self-consistent random-phase approximation plus particle-vibration coupling (RPA+PVC) model based on Skyrme density functional. We find that the electromagnetic transitions with multipolarity J<|m_{γ}| are forbidden for vortex γ photons due to the angular momentum conservation, with m_{γ} being the projection of total angular momentum of γ photon on its propagation direction. For instance, this allows for probing the isovector giant quadrupole resonance without interference from dipole transitions using vortex γ photons with m_{γ}=2. Furthermore, the electromagnetic transition with J=|m_{γ}|+1 vanishes at a specific polar angle. Therefore, the giant resonances with specific multipolarity can be extracted via vortex γ photons. Moreover, the vortex properties of γ photons can be meticulously diagnosed by measuring the nuclear photon-absorption cross section. Our method opens new avenues for photonuclear excitations, generation of coherent γ photon laser and precise detection of vortex particles, and consequently, has significant impact on nuclear physics, nuclear astrophysics and strong laser physics.

EXFOR-based simultaneous evaluation for fast neutron-induced fission cross section of thorium-232

ABSTRACT The 232Th neutron-induced fission cross section was evaluated from 500 keV to 200 MeV. The experimental 232Th fission cross sections and their ratios to the 235,238U fission cross sections in the EXFOR library were reviewed and analysed by using the least-squares method. The newly published 232Th/235U fission cross-section ratios from the time-of-flight measurements at the CERN n_TOF and CSNS Back-n facilities were compiled in EXFOR. Additional simultaneous evaluation was performed by including the experimental 233,238U and 239,240,241Pu fission cross sections and their ratios. The new evaluation provides the 232Th fission cross section systematically lower than the JENDL-5 cross section. The reduction is 4% in the plateau region between 2 and 6 MeV and more significant in the subthreshold fission region. The present evaluation reduces the 232Th fission cross section averaged over the 252Cf spontaneous fission neutron spectrum from the JENDL-5 evaluation by 4%, which is closer to the other general purpose libraries but underestimates Grundl et al.’s measurement by 11%.

A detector system for `absolute' measurements of fission cross sections at n_TOF in the energy range below 200 MeV

A new measurement of the 235U(n,f) cross section was performed at the neutron time-of-flight facility n_TOF at CERN. The experiment focused on neutron energies from 20 MeV to several hundred MeV, and was normalized to neutron scattering on hydrogen. This is a measurement first of its kind at this facility, in an energy range that was until now not often explored, so the detector development phase was crucial for its success. Two detectors are presented, a parallel plate fission chamber (PPFC) and a recoil proton telescope (RPT), both dedicated to perform measurements in the incident neutron energy range from 30 MeV to 200 MeV. The experiment was designed to minimize statistical uncertainties in the allocated run time. Several efforts were made to ensure that the systematic effects were understood and under control. The results show that the detectors are suited for measurements at n_TOF above 30 MeV, and indicate the path for possible future lines of development.

Cross-Evaluation of the PU-MET-FAST criticality benchmark experiments in ICSBEP handbook (by the end of 2022)

Criticality experiments play an important role in the fields of criticality safety, neutron transport program validation and nuclear data validation. ICSBEP handbook collects about 5000 criticality benchmark experiments. Most of the criticality benchmark experiments in ICSBEP handbook are retrospective evaluation experiments, they are evaluated long after the experiments and their evaluations may be based on inaccurate or incomplete data. When these experiments are used in nuclear data validations and cross section adjustments, contradictions between some criticality benchmark experiments have been found. Therefore, it is necessary to analyze whether these retrospective evaluation criticality benchmark experiments are applicable in the fields of nuclear data validations (including cross section adjustments) and validations of nuclear system design. To solve this problem, we have developed cross-evaluation method of criticality benchmark experiments, which could evaluate the biases of criticality benchmark experiments from the point of statistics. In this article, we used this method to cross-evaluate the PU-MET-FAST criticality benchmark experiments in ICSBEP handbook. The results indicate that there are many contradictions between the retrospective evaluation experiments in the PU-MET-FAST category. Because of the limited number of experiments in this category, the number of experiments that are similar to each other after cluster analysis is small, which reduces the effectiveness of the cross-evaluation. There are only 1 or 2 experiments in some clusters. Even if there are contradictions in these clusters, it is impossible to judge which experiment has a higher quality. Based on these results, it is recommended to do new PU-MET-FAST criticality benchmark experiments to resolve contradictions between these retrospective evaluation experiments and to help validate the nuclear data more effectively.

Measurement of fast neutron spectrum in the radial channel of VR-1 research reactor

Having knowledge of the spectrum of fast neutrons in the irradiation position of a reactor plays an important role in many experimental settings. If well described, it can be used as a reference neutron field for validating and testing measuring devices or detectors. This paper compares model calculations with measurements of the neutron spectrum in the research reactor VR-1. The fast neutron spectrum was measured in the radial channel of VR-1 reactor using a stilbene scintillation detector and the NGA-01 measuring device. The NGA-01 is a two-parameter spectrometric system for neutron and gamma radiation fields. The calculations were performed in a critical mode of the MCNP6 code using ENDF/B-VII.1 nuclear data library. Generally, a good agreement was achieved. However, a slight discrepancy in the neutron spectrum between the calculated and measured values was found in the 2–3 MeV region. In other regions, the differences between calculated and experimental values are comparable with uncertainties. Since the detector is distanced from the core, the neutron beam can be assumed parallel and therefore possible issues with stilbene anisotropy are eliminated. Thus, this experiment can be used for validation of a reactor leakage spectrum and nuclear cross sections in the energy region above 6 MeV, which is of interest for the nuclear data section of International Atomic Energy Agency.

Classification of events from [formula omitted]-induced reactions in the MUSIC detector via statistical and ML methods

The Multi-Sampling Ionization Chamber (MUSIC) detector is typically used to measure nuclear reaction cross sections relevant for nuclear astrophysics, fusion studies, and other applications. From the MUSIC data produced in one experiment scientists carefully extract an order of 103 events of interest from about 109 total events, where each event can be represented by an 18-dimensional vector. However, the standard data classification process is based on expert-driven, manually intensive data analysis techniques that require several months to identify patterns and classify the relevant events from the collected data. To address this issue, we present a method for the classification of events originating from specific α-induced reactions by combining statistical and machine learning methods that require significantly less input from the domain scientist, relative to the standard technique. We applied the new method to two experimental data sets and compared our results with those obtained using traditional methods. With few exceptions, the new method yields results such that the percent change with respect to results obtained with traditional methods are within ±20%. With the present method, which is the first of its kind for the MUSIC data, we have established the foundation for the automated extraction of physical events of interest from experiments using the MUSIC detector.

Determination of the number of 235U target nuclei in the irregular target using a fission time projection chamber

Based on multiple measurements of ionization loss, the Time Projection Chamber (TPC) combines strong tracking ability with particle identification ability in a large momentum range, which is an important advantage of TPC detection technology over traditional ionization measurement technology. According to these two characteristics of TPC, applying it to the measurement of fission cross-section can greatly improve the measurement accuracy. During the measurement of the fission cross-section, the number of target nuclei is required to be accurately measured. So this paper introduces a method for measuring the number of 235U target nuclei using a fission TPC system. The measurement result agrees with the reference value, and relative error is around 1 %.

4th-Order-SENS: A Software Module for Efficient and Exact 4th-Order Sensitivity Analysis of Neutron Transport

This work presents a software module called 4th-Order-SENS, which enables the efficient computation of exactly obtained expressions for all sensitivities, up to and including the 4th order, of a functional of the particle flux (e.g., the leakage of particles out of a body) with respect to nuclear parameters (total, scattering, and fission cross sections; nu, chi, sources; and number densities) for systems modeled by the neutron transport equation. The 4th-Order-SENS module implements the nth-Order Comprehensive Adjoint Sensitivity Analysis Methodology for Linear Systems (nth-CASAM-L), which is the only practically implementable methodology for obtaining the exact expressions of arbitrarily high-order sensitivities of model responses to model parameters, for response-coupled forward/adjoint large-scale linear systems. In addition to presenting the equations that are solved to obtain the 1st-order through 4th-order sensitivities, this work also describes the components of the module 4th-Order-SENS, including the user interface, input file, output files, and several independent code verification capabilities using symmetries and/or finite-difference formulas. The 4th-Order-SENS module is written in Python and Fortran and runs on Linux platforms. Several illustrative applications involving fixed-source problems in one-dimensional spherical and slab geometries are also presented.

Dosimetric effects of inserted non-radioactive elements in a tumor area in proton therapy

Introduction: Various prompt gamma (PG) ray-based techniques have been proposed to monitor the proton range during treatment, but the poor PG statistics produced entangle their clinical application. Recently, we developed a Monte Carlo (MC) simulation model for the enhancement of PG production based on the nuclear cross section of protons with the material transversed, via inserting the non-radioactive elements 19F, 17O, and 127I in a hypothetical tumor area. This work aimed to study the dose distribution changes caused by the addition of specific % weight fractions of PG enhancers in a water medium using MC simulations.Methods: Our MC model was created using the TOPAS MC package and retaining identical geometries, mixture compositions, and incident proton energies (75, 100, and 200 MeV, respectively) previously tested for the PG statistics enhancement. The total dose deposition in water and in mixture compositions that have been found to maximally increase the PG production was scored and evaluated based on the important dosimetric metrics R90, Bragg peak (BP) width, and full width at half maximum (FWHM). Furthermore, the spatial correlation of PG emissions relative to BPs was also studied and compared on the basis of the BP decrease at R90.Results and Discussion: There is no significant change in total dose deposition except for 127I. However, dose curve shifts in R90 toward shallower depth, followed by steeper BP and reduced FWHM, were observed in all cases. The percentage changes vary with incident proton energies and mixture compositions. The addition of the stable elements had no effect on the PG spatial emission. The dosimetric study reveals that the addition of the tested stable elements did not change the dose distribution and did not alter the dose deposited by secondary particles. Since their addition increases the electron density of the medium, the shift of BP to shallower depths is linked with the mixture composition changes. Furthermore, a steeper BP value is observed that could be beneficial for OAR avoidance. Since this theoretical study of using 19F, 17O, and 127I as PG enhancers is promising from a dosimetric point of view, experimental studies are necessary to determine their clinical application feasibility.

Sensitivity and uncertainty analysis of effective multiplication factor of KRITZ-1 benchmark using ENDF/B-VIII.0 and JEFF3-3 nuclear data libraries

Sensitivity and uncertainty analysis have been performed to estimate the uncertainty in the effective neutron multiplication factor, keff, of the KRITZ-1 benchmark and to identify which nuclear reactions and isotopes of interest could be significantly influencing it. First, the adjoint-weighted perturbation method using the KSEN card of the MCNP6 code is applied to compute the sensitivity vectors for 235U, 238U, 1H, 16O, and 10B isotopes with the JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries. Then, the sensitivity vectors and covariance matrices were combined to calculate the uncertainty in keff due to uncertainties in nuclear data. The 44 group covariance matrices have been processed with the NJOY21 system.It was found that the uncertainty in keff is mostly due to uncertainties on ν‾p , fission cross section of 235U, the capture cross section of the two isotopes 238U and 1H in the aforementioned data libraries

Phenomenological Description and Parameterization of Nucleon-Induced Fission Cross Sections of Heavy Nuclei in Intermediate and High Energy Region

Phenomenological Description and Parameterization of Nucleon-Induced Fission Cross Sections of Heavy Nuclei in Intermediate and High Energy Region

Empirical formula for (n, f) reaction cross sections of Uranium isotopes at 1–20 MeV neutrons

This article describes how to calculate neutron-induced fission reaction cross sections using a proposed empirical formula and the EMPIRE 3.2.3 and TALYS 1.95 computer codes for Uranium isotopes up to the third fission plateau. In this study, the excitation functions of 232U (n, f), 233U (n, f), 234U (n, f), 235U (n, f), 236U (n, f), 237U (n, f) and 238U (n, f) nuclear reactions were calculated at 1–20 MeV neutron energies. The results were contrasted to measured values from the Experimental Nuclear Reaction Data (EXFOR) as well as the evaluated data from Evaluated Nuclear Data File (ENDF) such as (EAF-2010, JEFF-3.3, ENDF/B-VIII.0 and TENDL-2019). Overall, the calculated, experimental, and evaluated fission cross-sections are in concordance.