Fission Cross Sections Research Articles

Research on Emerging Medical radionuclides from the X-sections (REMIX): The Accelerator-based Production of 47Sc, 149Tb, 152Tb, 155Tb and 161Tb

The REMIX project is focused on the cyclotron-based production of 47Sc, 149Tb, 152Tb, 155Tb and 161Tb radionuclides, whose decay characteristics make them suitable for medical applications. This work will outline the main results achieved withing the REMIX collaboration, that is organized in the following Work Packages (WP): WP1. Target manufacturing (49Ti, 50Ti and 155Gd2O3) and characterization; WP2. Nuclear cross section (XS) measurements with 49Ti and 50Ti targets for 47Sc production; WP3. Nuclear XS measurements with natDy, 159Tb and natEu targets for xxTb production; WP4. Nuclear XS modeling for 47Sc and 155Tb production; WP5. Dosimetric calculations for 47Sc- and xxTb-labelled radiopharmaceuticals; WP6. 155Tb Thick Target Yield (TTY) measurements; WP7. Apparatus design and realization for irradiation tests with the LARAMED beamline. Since the LARAMED bunkers and ancillary laboratories are currently under completion at the INFN-LNL, the nuclear XS experiments are carried out in collaboration with the GIP ARRONAX facility (Saint-Herblain, France) and the Thick Target Yield (TTY) measurements are performed at the Sacro Cuore Don Calabria hospital (SCDCh, Negrar, Verona, Italy).

Double folding analysis of the α+28Si elastic scattering system

Using the double folding model (DFM), we reanalyzed 20 sets of experimental angular distributions for the elastic scattering [Formula: see text]+[Formula: see text]Si system in a wide energy range of [Formula: see text]-particles from 12.7[Formula: see text]MeV to 240[Formula: see text]MeV. We used three effective nucleon–nucleon (NN) interactions that are CDM3Y6, São Paulo Potential (SPP2), and Brazilian Nuclear Potential (BNP) to generate the DF potentials. The data analysis was performed using semi-microscopic (SM) and fully microscopic (FM) methods. In the SM approach, the real part of the optical potential (OP) is created with the DFM, and the Woods–Saxon (WS) potential is considered for the imaginary part of the OP. In the FM approach, the DF potential renormalization factors [Formula: see text] and [Formula: see text] for the real and imaginary parts of the OP were extracted. We investigated the energy dependencies of the [Formula: see text] and [Formula: see text] factors, as well as of nuclear potential volume integrals and reaction cross-sections. We reasonably reproduced the data using [Formula: see text] and [Formula: see text] factors obtained from their energy-dependent functions.

Measurement of the neutron energy response curve of 4H-SiC detector based fission target detection system at the CSNS Back-n white neutron source

The fission target detection system, which is based on U-235(or U-238)fission target and large area 4H-SiC detector, has been developed for pulsed neutron detection. However, calibration of its neutron energy response curve is challenging due to its very low detection efficiency to fast neutrons. In the past, the calibration of the neutron sensitivity for the fission target detection system was limited to using ∼ 14 MeV neutrons generated by the Cockcroft-Walton accelerator. To obtain the response function for the complete range of neutron energies, extrapolation based on neutron fission cross sections was required. In this paper, we present the calibration of its energy response at the China Spallation Neutron Source (CSNS) Back-n beam line using the time of flight method, both with double-bunch and single-bunch mode in end-station 1(with a flight path of ∼ 55 m). The SiC detector with a thin sensitive thickness of 30 μm has lower scattering background than the traditional Si-PIN detector . As a result, the energy response curve of the fission target detection system above 0.26 MeV has been successfully acquired. The spectrum of the double-bunch mode was unfolded using the Bayesian algorithm, and the results indicate that, under conditions of sufficient counts, this method can yield results similar to those obtained from the single-bunch mode. The issue of pulse height defect of SiC detector to fission fragments is discussed. This experiment confirms that our detection system of SiC+fission target has a strong signal-to-noise ratio and anti-interference capability. At the same time, it demonstrates that this device has its own advantages in measuring the energy response of low sensitivity detectors.

Separation of 44mSc/44gSc Nuclear Isomers Based on After-Effects.

44gSc presents a particular interest for application in nuclear medicine for positron emission tomography (PET) due to its favorable nuclear decay properties (t1/2 = 3.97 h, Emax = 1.47 MeV, branching ratio 94.3% β+). Its nuclear isomer 44mSc (t1/2 = 58.61 h) decays by isomeric transition (IT) into 44gSc, accompanied by ≈12% of conversion electron emission, which can cause a partial release of the daughter 44gSc from the chelate complex. A 13 MeV cyclotron at TRIUMF was used to produce both 44mSc and 44gSc via the natCa(p,n)44m,gSc reaction. A 44mSc/44gSc generator was designed by using a Strata C-18E cartridge. After several tested systems, a successful separation method was developed using DOTATOC as a chelator, a Strata C-18E cartridge as a generator column, and an elution solution of 0.1 M NH4-α-HIB. The yield of the generator with the daughter 44gSc release was equal to 9.8 ± 1.0% (or ≈80% per portion of conversion). This result shows the important role of after-effects in the design of radionuclide generators. Nuclear cross-section calculations were applied using the TALYS code to allow for the determination of the most promising alternative routes for 44mSc production, which will enable the development of a full-scale 44mSc/44gSc radionuclide generator based on after-effects.

Evaluation of the production amount of 225Ac and its uncertainty through the 226Ra(n,2n) reaction in the experimental fast reactor Joyo

ABSTRACT 225Ac is a promising nuclide for targeted alpha therapy. At present, the nuclide is mainly extracted from 229Th generators, but its total radioactivity of the world is only 63 GBq/y. Therefore, it is unclear whether this method alone can meet future global medical demand. In this study, we analyzed the production amount of 225Ac and its uncertainty due to the neutron flux and nuclear reaction cross-sections while analyzing the generation process of 225Ac using the experimental fast reactor Joyo operated by the Japan Atomic Energy Agency (JAEA). The analysis showed that irradiating 1 g of 226Ra for 45 d at the core center of Joyo can produce 15.4 ± 6.2 GBq of 225Ac. Therefore, with six irradiations per year (45 d × 6 = 270 d/year), Joyo can produce 55.3 to 129.1 GBq of 225Ac, equivalent to 86 to 200% of the current world supply.

Experimental assessment of nuclear cross sections for the production of Tb radioisotopes with a medical cyclotron

155Tb is one of the most interesting radionuclides for theranostic applications. It is suitable for SPECT imaging and it can be used as a true diagnostic partner of the therapeutic 149Tb and 161Tb. Its production by proton irradiation using enriched 155Gd and 156Gd oxide targets is currently being investigated and represents a promising solution. To achieve the level of radionuclidic purity required in the clinical setting, the co-production of Tb impurities has to be minimized. For this purpose, an accurate knowledge of the cross sections of the nuclear reactions involved is of paramount importance. In this paper, we report on the assessment of cross sections of the reactions 154Gd(p,xn)153,154,154m1,154m2Tb, 155Gd(p,xn)154,154m1,154m2,155Tb, 156Gd(p,xn)155,156Tb and 157Gd(p,2n)156Tb derived with a specific data analysis procedure developed by our group. This method allows to disentangle the nuclear contributions from the production cross section by inverting linear systems of equations and it requires the measurement of the cross sections from as many materials as the reactions involved in the production of the radionuclide under study. For this purpose, the experimental data previously measured by our group at the Bern medical cyclotron by irradiating natural Gd2O3, enriched 155Gd2O3 and enriched 156Gd2O3 targets were used. For some of these nuclear reactions, cross sections were assessed for the first time. On the basis of our findings, production yield and purity can be calculated for any kind of isotopic composition of the enriched material.

Cross-section measurement of thulium radioisotopes with an 18 MeV medical PET cyclotron for an optimized 165Er production

165Er is a pure Auger-electron emitter with promising characteristics for therapeutic applications in nuclear medicine. The short penetration path and high Linear Energy Transfer (LET) of the emitted Auger electrons make 165Er particularly suitable for treating small tumor metastases. Several production methods based on the irradiation with charged particles of Er and Ho targets can be found in the literature. In this paper, we report on the study of 165Er indirect production performed via the 166Er(p,2n)165Tm →165Er reaction at the 18 MeV Bern medical cyclotron. Despite the use of highly enriched 166Er2O3 targets, several Tm radioisotopes are produced during the irradiation, making the knowledge of the cross sections involved crucial. For this reason, a precise investigation of the cross sections of the relevant nuclear reactions in the energy range of interest was performed by irradiating Er2O3 targets with different isotopic enrichment levels and using a method based on the inversion of a linear system of equations. For the reactions 164Er(p, γ)165Tm, 166Er(p,n)166Tm, 166Er(p, γ)167Tm, 167Er(p,3n)165Tm, 167Er(p, γ)168Tm, 168Er(p,2n)167Tm and 170Er(p,3n)168Tm, the nuclear cross section was measured for the first time. From the results obtained, the production yield and purity of the parent radioisotope 165Tm were calculated to assess the optimal irradiation conditions. Several production tests with solid targets were performed to confirm these findings.

Strengths and weaknesses of the modal expansion method for perturbations calculations in nuclear reactor physics

Reactor design and uncertainties propagation request repeated execution of neutron diffusion calculations. To address this, one can rely on the Generalized Perturbation Theory which is restricted to single integral values such as reaction rate ratios. To overpass such limitations, the perturbed flux could be calculated with the modal expansion method. The difficulty to calculate more than one eigenvector has limited its use. Hence, modal expansion lacks a clear explanation of its strengths and weaknesses. Using recent advances in eigenvalues solvers, we have successfully applied modal expansion to the 2D-BIBLIS and IAEA-3D benchmarks. Our results are promising for fission and capture cross sections perturbations. Still, modal expansion is faulty for local perturbations or when it concerns the reflector. In terms of computational cost, modal expansion is more efficient than solving the direct problem with a calculation time divided by 8 for a basis of 20 eigenvectors.

Feasibility study of a compact and multi-gas supersonic plasma jet for nuclear astrophysics and space research

Supersonic plasma jet targets can improve the present knowledge on low energy nuclear reaction cross sections allowing to study electron screening effects. They could also be employed both in the preliminary tests of space vehicles thermal protection systems and in the validation of computational fluid dynamics (CFD) evaluations of hot exotic gases. This common interest between nuclear astrophysics and space research led to a joint effort between the European Recoil Separator for Nuclear Astrophysics collaboration of the Italian National Institute for Nuclear Physics and the Italian Aerospace Research Center (CIRA) to develop an innovative, compact and multi-gas supersonic plasma jet setup. In this work we present the characterization of a test supersonic cold jet target, performed with ion beam analysis techniques, as benchmark for the validation of the CFD code NExT developed by CIRA. Being NExT outputs in agreement within 5% with experimental data, we employed it together with the software CEA of the United States National Aeronautics and Space Administration to design setups to be employed in both nuclear astrophysics and space re-entry research fields. We finally provide an overview of the plasma technologies to be employed in the construction of this supersonic plasma jet, aiming for a good balance of costs and performances. A 200 kW plasma arch torch, being also commercially available, was found to be a good option for both nuclear astrophysics and space re-entry experiments in small laboratories: for the first, a 20% of ionization may be reached with still an adequate target thickness (∼1017nuclei/cm2), while for the latter the most energetic scenario could be reproduced.

Measurement of alpha-induced reaction cross-sections for [formula omitted]Zn with detailed covariance analysis

The production cross-section of 68Ge, 69Ge, 65Zn and 67Ga radioisotopes from alpha-induced nuclear reaction with natZn have been measured using the stacked foil activation technique followed by the off-line γ-ray spectroscopy in the incident alpha energy range 14-37 MeV. The obtained nuclear reaction cross-sections are compared with previous experimental data available in the EXFOR data library, evaluated nuclear data from TENDL-2019 and theoretical results, calculated using TALYS nuclear reaction code. We have also performed the detailed uncertainty analysis for these nuclear reactions and their respective correlation metrics are presented. Since α-induced reactions are important in nuclear medicine and developing the nuclear reaction codes so needful corrections related to the coincidence summing factor and the geometric factor have been considered during the data analysis in the present study.

CHARGED AND NEUTRAL PARTICLES BEAMS APPLICATION FOR RESEARCH OF ACCUMULATION ROUTES OF THE MEDICAL RADIOISOTOPE 103PD

The results of experimental data analysis about the accumulation routes of the important medical isotope 103Pd induced by charged particle beams (protons, deuterons, helium-3, helium-4) and neutral particles (gamma and neutrons) are presented. The main generalized data about the nuclear reaction cross-section of 103Pd production versus the irradiation techniques of natural and isotope-enriched targets and their content of unwanted accompanying radio nuclides were compared to theoretical predictions of too. The optimization procedure of the 103Pd accumulation is discussed in order to develop more promising ones, taking into account the requirements for the production of medical radioisotopes.

Progress in measurements of fission cross sections and total cross sections at CSNS Back-n white neutron source

The neutron-induced cross sections are of great significance for the design of nuclear devices and advanced reactors for the nuclear energy production. At CSNS Back-n white neutron source, new measurements of the fission cross sections and total cross sections are performed with two sets of Day-one spectrometers based on the multi-cell fast fission ionization chamber (FIC). The neutron-induced 236,238U fission cross sections relative to 235U from the fission threshold energy to 200 MeV were measured by using the TOF method and the Fast Ionization Chamber Spectrometer for Fission Cross Section Measurement (FIXM) in the single/double bunch mode of Back-n. The experimental uncertainties are analyzed in detail, and the results from the two modes are consistent. The measured 236,238U/235U fission cross section ratios are compared with previous experiments and evaluations. The 236,238U(n,f) cross sections are obtained based on the standard 235U fission cross section. The neutron total cross sections of carbon and aluminum in the energy region from 1 eV to 20 MeV have been measured by using the TOF method and transmission method based on the Neutron Total Cross Section Spectrometer (NTOX) in the double bunch mode. The total cross section results after unfolding are in good agreement with the previous measurements as well as the broadening of the ENDF/B-Ⅷ.0 evaluation with Gaussian function within the experimental uncertainty. The present results provide the experimental data for further measurements, relevant evaluations and the design of nuclear system.

Cryomilling of Isotope-Enriched Ti Powders for HIVIPP Deposition to Manufacture Targets for Nuclear Cross Section Measurement.

The realization of isotopically enriched Ti targets for nuclear cross-section measurements requires particular attention, from the starting material preparation up to the deposition technique. In this work, a cryomilling process was developed and optimized, aimed at reducing the size of 49,50Ti metal sponge as provided by the supplier (size up to 3 mm), to the optimal size of 10 µm, to fit the High Energy Vibrational Powder Plating technique used for target manufacturing. The optimization of the cryomilling protocol and the HIVIPP deposition using natTi material was thus performed. The scarce amount of the enriched material to be treated (about 150 mg), the need to obtain a non-contaminated final powder and a uniform target thickness of about 500 µg/cm2 were taken into account. The 49,50Ti materials were then processed and 20 targets of each isotope were manufactured. Both powders and the final Ti targets produced were characterized by SEM-EDS analysis. The amount of Ti deposited was measured by weighing, indicating reproducible and homogeneous targets, with an areal density of 468 ± 110 µg/cm2 for 49Ti (n = 20) and 638 ± 200 µg/cm2 for 50Ti (n = 20). The uniformity of the deposited layer was also confirmed by the metallurgical interface analysis. The final targets were used for the cross section measurements of the 49Ti(p,x)47Sc and 50Ti(p,x)47Sc nuclear reaction routes aimed at the production of the theranostic radionuclide 47Sc.

Fission cross-section measurement for [formula omitted] and [formula omitted] in [formula omitted] reaction induced by 14MeV D-T neutrons

In this work, the independent fission cross-sections of U(n,f)238Xe135g and U(n,f)238Xe135m reactions induced by neutrons with the energies of 14.1 MeV, 14.5 MeV, and 14.7 MeV were measured by activation methods. The neutrons from T(d,n)He4 were used in the experiments and their energies were determined by the ratio of Zr(n,2n)90Zr89 and Nb(n,2n)93Nb92m reaction cross-sections. Aluminum films were chosen as reference samples to measure the neutron fluence relative to the cross section of the Al(n,α)27Na24 reaction. The effect of self-absorption, geometry, and cascade coincidence were also considered during the data analysis. Besides, the increase in the daughter nuclide yield due to the decay of the parent nuclides in the same decay chain was deducted. As a result, the independent fission cross-sections of the U(n,f)238Xe135g reaction are 2.54 ± 0.14 mb, 3.05 ± 0.19 mb and 2.94 ± 0.19 mb, while the cross-sections of U(n,f)238Xe135m reaction are 2.11 ± 0.16 mb, 2.47 ± 0.18 mb and 2.34 ± 0.21 mb for 14.1 MeV, 14.5 MeV and 14.7 MeV neutrons, respectively. This work provides experimental data for the database of nuclear fission reactions.

Analyses of the bias and uncertainty of SNF decay heat calculations using Polaris and ORIGEN

The bias and uncertainty of calculated decay heat from spent nuclear fuel (SNF) are essential for code validation. Also, predicting these quantities is crucial for deriving decay heat safety margins, influencing the design and safety of facilities at the back end of the nuclear fuel cycle. This paper aims to analyze the calculated spent nuclear fuel decay heat biases, uncertainties, and correlations. The calculations are based on the Polaris and ORIGEN codes of the SCALE code system. Stochastically propagated uncertainties of inputs and nuclear data into calculated decay heats are compared. Uncertainty propagation using the former code is straightforward. In contrast, the counterpart of ORIGEN necessitated the pre-generation of perturbed nuclear cross-section libraries using TRITON, followed by coincident perturbations in the ORIGEN calculations. The decay heat uncertainties and correlations have shown that the observed validation biases are insignificant for both Polaris and ORIGEN. Also, similarities are noted between the calculated decay heat uncertainties and correlations of both codes. The fuel assembly burnup and cooling time significantly influence uncertainties and correlations, equivalently expressed in both Polaris and ORIGEN models. The analyzed decay heat data are highly correlated, particularly the fuel assemblies having either similar burnup or similar cooling time. The correlations were used in predicting the validation bias using machine learning models (ML). The predictive performance was analyzed for machine learning models weighting highly correlated benchmarks. The application of random forest models has resulted in promising variance reductions and predicted biases significantly similar to the validation ones. The machine learning results were verified using the MOCABA algorithm (a general Monte Carlo-Bayes procedure). The bias predictive performance of the Bayesian approach is examined on the same validation data. The study highlights the potential of neighborhood-based models, using correlations, in predicting the bias of spent nuclear fuel decay heat calculations and identifying influential and highly similar benchmarks.

Measurement of the Cross Section for the Neutron-Induced Fission of 238U Nuclei in the Energy Range of 0.3–500 MeV

The cross section for the neutron-induced fission of 238U nuclei has been measured using the time-of-flight spectrometer of the GNEIS neutron complex at the Petersburg Nuclear Physics Institute, National Research Center Kurchatov Institute, in the neutron energy range of 0.3–500 MeV. Fission fragments have been detected by low-pressure position-sensitive multiwire proportional counters. The cross section for 238U(n, f) fission has been measured with respect to the cross section for 235U(n, f) fission, which is an accepted international standard. Data on the energy dependence of the anisotropy of the angular distribution of fragments of neutron-induced 238U nuclei are also presented. The data obtained have been compared to previous experiments carried out using both similar and significantly different methods.

A Novel Scheme of the ARA Transform for Solving Systems of Partial Fractional Differential Equations

In this article, a new analytical scheme of the ARA transform is introduced to solve systems of fractional partial differential equations. The principle of the proposed technique is based on combining the ARA transform with the residual power series method to create an approximate series solution for a system of partial differential equations of fractional order on the form of a rapid convergent series. To illustrate the effectiveness, accuracy, and validity of the suggested technique, an Attractive physical system, the fractional neutron diffusion equation with one delayed neutrons group, is discussed and solved. Two different neutron flux initial conditions are presented numerically to clarify various cases in order to ensure the theoretical results. The necessary Mathematica codes are run using vital nuclear reactor cross-section data, and the results for various values of time are tabulated and graphically represented.

Computational simulation of minor actinide burning in a BN-600 reactor with fuel without uranium and plutonium

The paper presents the results of studies on the burning of minor actinides (MA) extracted from SNF of thermal reactors in a BN-600 reactor, which uses the complete set of MAs instead of traditional nuclear fuel types: uranium and/or plutonium. The advantages of such approach to MA burning are that long-lived waste is recycled and energy is produced that can be used, e.g., to generate electricity. Besides, where, e.g., a reactor with uranium or MOX fuel is used for transmutation, apart from burning “foreign” minor actinides, it will additionally generate “its own” MAs. The studies have shown that such reactor can be efficient only if based on fast neutrons, which is due to the specific properties of the minor actinide neutron capture and fission cross-sections as compared with traditional fuel nuclides. The calculation results have shown rather a high rate of MA transmutation and burning in a reactor fueled with minor actinides.

Measurement of neutron-induced fission cross sections of 232Th from 1 to 300 MeV at CSNS Back-n

232Th/233U fuel cycle is an alternate candidate for future nuclear energy. Neutron-induced fission cross section of 232Th is one of the important nuclear data if using thorium as the fuel in the future. 232Th(n, f) cross sections from 1 to 300 MeV were measured at China Spallation Neutron Source (CSNS) Back-streaming neutron facility (Back-n) relative to 235U(n, f) and n-p scattering. A fast ionization chamber for fission cross section measurement (FIXM) and a proton recoil telescope (PRT) were used to perform the measurement. The neutron energy in this measurement reaches up to 300 MeV which is supposed to be the highest among the current Experimental Nuclear Reaction Data (EXFOR) library. It stands as the only experimental data in the region of 200-300 MeV in EXFOR as well. The experimental and evaluated data of 232Th(n, f) cross section above 60 MeV are scarce, the measurement data of this work will be valuable references for future evaluation in high energy. For the data below 60 MeV, most of our data are more consistent with ENDF/B-VIII.0, excepting in the region of 1-2 MeV where our data are more in favor of CENDL-3.2.

Re-examination of radiative capture of deuteron 3He(d,γ)5Li at low energy

3He(d,γ)5Li is important in the Big Bang Nucleosynthesis (BBN), as it provides seeds for the 6Li formation. Within the framework of the potential model the 3He(d,γ)5Li is analyzed. In the present investigation the nuclear width , nuclear cross-sections, astrophysical S-factor, and nuclear reaction rates have been computed. The present study is in better agreement with the observed data.