Search

The fast neutron fluence and the activation monitor activities are routinely calculated with TORT deterministic code and BUGLE-B7 nuclear data library with 47 broad energy groups. The objective of the paper is to analyse options to improve reactor dosimetry transport calculations. There are two paths to improve reactor dosimetry calculations. Increasing geometry, angular and energy mesh size is applicable for TORT code while using newer nuclear data libraries is relevant for both deterministic and Monte Carlo codes. Two new calculation options (improved TORT and Monte Carlo MCNP6) were compared with the standard TORT calculation for VVER-440 Dukovany Unit 3 Cycle 31. The fast neutron fluence with 0.5 MeV threshold as well as activity of Fe. Ni, Ti, Cu. Mn and Nb monitors were evaluated. Standard TORT calculations were improved from S16P3 to S30P3 with three times finer axial mesh size. 120° core symmetry r-ϑ mesh size with 0.5° step and fine multigroup libraries VUAMIN-B7 with 199 neutron energy groups and ENDF/B-VH.1 with 200 neutron energy groups. Both ENDFB and IRDFF activation cross sections were used. The drawback of expanded mesh size is raised calculation runtime since TORT deterministic code is not parallelized and one calculation can require multiple weeks of CPU time. An alternative option of using MCNP6 Monte Carlo code with continuous ENDF/B-VH.1 nuclear data with detailed 3-D geometry and pin-wise effective neutron source prepared by MOBY-DICK diffusion code reactor analysis wras explored. It was found that using finer mesh size affects reactor dosimetry' tallies less than the choice of nuclear data library. BUGLE-B7 and VTTAMIN-B7 produce results typically within 1% difference. ENDF/B-VH.1 calculations with 200 neutron energy' groups with TORT code are even in better agreement with MCNP6 calculations with continuous nuclear data libraries. The largest differences of around 2% were observed between VTTAMIN-B7 library based onENDFB-VH.O nuclear data and ENDF/B-VH.1 library. Nuclear data library' has larger impact on the results with up to 7 % difference between all 0.5 MeV fast neutron fluence calculations. The largest intact of nuclear data was observed for Mn(n.2n) monitor. [ABSTRACT FROM AUTHOR]

In structural fusion material research, double differential sections are necessary to determine heating and damage due to secondary particles that may occur in structural materials. Therefore, in this study, the double differential proton emission cross sections of 27Al and 24Mg target nuclei were calculated theoretically with the TALYS nuclear reaction code at 14 MeV neutron energy. The calculated values were compared with existing experimental data in the EXFOR library. Additionally, the contribution of direct, compound and preequilibrium reactions in theoretical calculations was investigated. [ABSTRACT FROM AUTHOR]

This paper deals with the first radiation dose modelling of the Accelerator laboratory of the Institute of Slovak Academy of Sciences in Piešťany. The need arises from the planned neutron production in this facility. Neutrons will be produced via the D(d,n)3He nuclear reaction (Q = 3.269 MeV), where depending on the beam energy, quasi-monoenergetic neutrons with energies in the range between 2.5 – 6 MeV will be available for planned studies focused on fast neutron radiography. To allow the safe operation of the facility and to get the permission of the Public health authority of the Slovak Republic, a complex 3D model needs to be created to estimate the radiation doses of interest. The paper describes the model developed using the ORNL SCALE package, and the first results of neutron coupled simulations are shown and discussed. [ABSTRACT FROM AUTHOR]

The neutron Time-of-Flight (TOF) research facility at CERN, n_TOF, has been a pioneering platform for neutron cross-section measurements since its inception in 2001. It boasts three distinct experimental areas, each tailored to address a specific range of neutron energies. This paper delves into the intricacies of the n_TOF facility, including its recent upgrade during the Long Shutdown 2 (LS2) at CERN. Additionally, it highlights the key characteristics of the detectors employed for capture and fission cross-section measurements, paving the way for future research endeavors. [ABSTRACT FROM AUTHOR]

We have examined the structural properties of Pt-isotopes within the axially deformed Relativistic-Mean Field theory (RMF) with NL-3 and NL-3*as force parameters. We determined the binding energy (B.E.), the two neutron separation energy (S2n), the differential-neutron separation energies (dS2n), and some other observables of various isotopes of the Pt-element. The Finite-Range Droplet Model (FRDM) is extremely well compatible with both the force parameters, NL-3 and NL-3*. In addition to probable shell and/or sub shell closure at N=104 (A=182, Z=78), we also found stability at N=108 (A=186, Z=78). [ABSTRACT FROM AUTHOR]

The design of new generation fast nuclear reactors requires highly accurate cross-section measurements in the MeV energy region. The 242Pu fission cross section is of particular interest for Pu incineration and nuclear waste production. There are discrepancies around 1 MeV incident neutron energy between libraries and among experimental data. Some data suggest the presence of a strong structure between 1 and 1.2 MeV whereas it is barely visible on some other data and its shape is very different among evaluations. The large majority of the 242Pu(n,f) measurements have been carried out with respect to the 235U(n,f) secondary-standard cross section. This introduces a strong correlation between measurements from different research teams. Moreover, this reference cross section exhibits structures, in particular a steep increase of +10% at 1 MeV. Therefore, we aim to re-measure the 242Pu(n,f) cross section relative to the primary-standard 1H(n,n)p cross section, by using a proton recoil detector. This standard has a very high accuracy (0.4%), is not used for other 242Pu measurements, and is structureless. An experiment has been carried out in October 2022 at the MONNET facility in JRC Geel, with incident neutron energies from 0.9 MeV to 2.0 MeV. The experimental setup will be presented, and the analysis procedure will be detailed. [ABSTRACT FROM AUTHOR]

The total and double differential scattering cross sections (DDSCS) of isotropic graphite have been measured in the J-PARC Materials and Life Science Experimental Facility using five samples with different porosities and grain sizes. The DDSCS measurements were performed in Beam Line No.14 (AMATERAS) with nine incident neutron energies ranging from 1.2 to 94 meV. In the derived cross sections, there are no clear sample dependence. The neutron total cross sections in the energy region from 1 to 100 meV were measured using Beam Line No. 04 (ANNRI). All samples in the epithermal range (over 40 meV) tended to have values close to the free atom cross-section of graphite. However, at the first Bragg edge, the deduced total cross sections by these measurements started to vary. The difference became more significant with decreasing the neutron energy. [ABSTRACT FROM AUTHOR]

The paper aims to assess the usability of vacant areas within the Slovak University of Technology in Bratislava, to procure and install a new deuterium-deuterium neutron generator. The complex 3D model of the vacant areas either with their close vicinity is modelled in the SCALE system and the first simulations of radiation doses at the 3 principal positions are calculated utilizing the MAVRIC sequence and variance reduction methods. The assumed energy of neutrons emitted from the generator is 2.5 MeV, whereas a first approximation, the isotropic distribution is assumed. A discussion about the model simplifications and the first radiation dose results are provided. [ABSTRACT FROM AUTHOR]

The use of extensive graphite material as moderator, reflector, and regulating rod in thermal MSR often limits the operational life of the reactor due to graphite material integrity. The graphite material is challenged by fast neutron exposure during its service. Therefore, neutron fluence analysis on graphite material of TMSR-500 (ThorCon) was simulated with MCNP6 code. Neutron fluence calculations on the TMSR-500 were carried out with a nominal power of 557 MWt. The reactor operates at a temperature of around 977 K. The analysis is intended to identify the spot and estimate the maximum exposure to high-energy neutron fluence (>0.1 MeV) on the graphite material of TMSR-500 during four years of operation. The maximum fast neutron fluence (E> 0.1 MeV) in the graphite moderator of TMSR-500 is found to be 1.506×1022 n.cm−2 during four years of operation. Damage to graphite material could occur due to radiation exposure caused by high energy neutrons (E> 0.1 MeV) when it exceeds the recommended limit of neutron fluence 4.0×1022 n.cm−2. Thus, the fast neutron fluence exposure to the graphite moderator of TMSR-500 during 4 years of operation is well below the maximum acceptable limit. [ABSTRACT FROM AUTHOR]

Neutrons can induce background events in underground experiments looking for rare processes. Neutrons in a MeV range in deep underground laboratories are produced in spontaneous fission processes and (α, n) reactions. A number of computer codes are available to calculate cross-sections of (α, n) reactions, branching ratios to various states and neutron yields. We have used the SOURCES4 code to calculate neutron yields and energy spectra with input cross-sections and branching ratios taken from experimental data and calculations based on EMPIRE2.19/3.2.3 and TALYS1.9. We report here a comparison of SOURCES4 calculations with experimental data obtained with alpha beams with fixed energies. [ABSTRACT FROM AUTHOR]

The results of model calculations for (n, t) and (n, f) reaction cross section using a standard codes EMPIRE 3.2 for Neptunium (Np), Americium(Am) and Curium (Cm) element in neutron energy range En = 1-20 MeV are presented. Different nuclear model presents in the codes have been considered for the calculation of cross sections of desired nuclear reactions. The contribution of various model to the cross section within the compound nucleus CN pre-equilibrium emission PE and direct reaction DR is also discussed. Our data are compared with the existing experimental data and current evaluated nuclear data from ENDF/B-VIII.0, CENDL 3.2 and JEFF 3.3 and accurate results predicted is therefore crucially important to study radiative waste transmutation and design of various nuclear system [ABSTRACT FROM AUTHOR]

There has been a continued effort since 2019 within the IAEA INDEN collaboration to improve the evaluation of neutron induced reactions on iron isotopes. The reason for the 30% underestimation of the neutron leakage spectrum from a thick iron sphere was found primarily to be due to the overestimation of the inelastic cross sections in the 56Fe evaluated data file produced within the CIELO project of the OECD/NEA Data Bank. The over-estimation of the neutron flux between the resonances near 300 keV was traced to neglecting the fluctuating nature of the total cross section of 57Fe in the fast neutron energy range, since the evaluated resolved resonance range of 57Fe extended only up to 190 keV. The added 1/v background in the "iron window" below 28 keV is in excellent agreement with the independently evaluated one in the JENDL-5.0 library that included the direct capture component in the evaluation. Performance of the updated 56,57Fe evaluations was tested on a set of criticality benchmarks from the ICSBEP Handbook, including the dependence on reflector thickness and on new deep penetration shielding benchmark using a 252Cf(sf) neutron source undertaken at Rez, Czech Republic. Neutron leakage for 43 MeV incident neutrons was also validated. [ABSTRACT FROM AUTHOR]

The n_TOF facility hosts CERN's pulsed neutron source, comprising two beam lines of different flight paths and one activation station. It is based on a proton beam delivered by the PS accelerator impinging on a lead spallation target. During Long Shutdown 2 (LS2) at CERN (2019-2021), a major upgrade of the spallation target was carried out in order to optimize the performances of the neutron beam. Therefore, the characteristics of n_TOF two experimental areas were investigated in detail. In this work, the focus is on the second experimental area (EAR2), located 20 m above the spallation target. Preliminary results of the neutron energy distribution and beam line energy resolution are presented, compared to previous experimental campaigns and Monte Carlo simulations with the FLUKA code. Moreover, preliminary results of the spatial beam profile measurements are shown. [ABSTRACT FROM AUTHOR]

To probe the neutron inelastic scattering off 232Th, an experiment took place at the EC-JRC Geel conducted with the experimental setup GRAPhEME to detect emitted γ-rays. The prompt γ-ray spectroscopy method was used and 70 experimental 232Th(n, n'γ) cross sections were obtained from the experimental data. Combining these cross sections, nuclear-structure data available in databases and hypotheses to complete the latter, neutron inelastic level production cross sections in 232Th and the total inelastic cross section were calculated. For the first time, the total inelastic cross section of an actinide nucleus was derived on the total neutron energy range from experimental data only. Comparisons of (n, n') cross section data with evaluated data reveal a good agreement between them all above 300 keV of neutron energy. TALYS calculations are compatible but lower than the evaluated data. [ABSTRACT FROM AUTHOR]

miniBELEN is a modular and transportable neutron moderated counter with a nearly flat neutron detection efficiency up to 10 MeV. Modularity implies that the moderator can be reassembled in different ways in order to obtain different types of response. The detector has been developed in the context of the Measurement of Alpha Neutron Yields (MANY) collaboration, which is a scientific effort aiming to carry out measurements of (훼, 푛) production yields, reaction cross-sections and neutron energy spectra. In this work we present and discuss several configurations of the miniBELEN detector. The experimental validation of the efficiency calculations using 252Cf sources and the measurement of the 27Al(훼, 푛) 30P reaction is also presented. [ABSTRACT FROM AUTHOR]

A new measurement of the 235U(n,f) fission cross section was carried out at n_TOF. The experiment covered the neutron energy range from 10 MeV up to 500 MeV, and it used the 1H(n,n) cross section as normalization for the neutron fluence measurement. In this contribution, the measurements and the characterization of the detectors covering the incident energy range up to 150 MeV are discussed. [ABSTRACT FROM AUTHOR]

The measurement of the 235U(n,f) reaction cross section in the neutron energy region 10 MeV to 500 MeV was carried out at the CERN n_TOF facility. The experimental campaign, completed in 2018, provided precise and accurate data on the fission reaction relative to neutron-proton elastic scattering. A description and characterization of the used setup for the simultaneous measurement of fission fragments and neutron flux is reported here. [ABSTRACT FROM AUTHOR]

The Hauser-Feshbach fission fragment decay model, HF3D, calculates the statistical decay of fission fragments through both prompt and delayed neutron and γ-ray emissions in a deterministic manner. While previously limited to the calculation of only first-chance fission, the model has recently been extended to include multi-chance fission, up to neutron incident energies of 20 MeV. The deterministic decay takes as input prescission quantities–fission probabilities, pre-fission neutron energies, and the average energy causing fission– and post-scission quantities–yields in mass, charge, total kinetic energy, spin, and parity. From those fission fragment initial conditions, the full decay is followed through both prompt and delayed particle emissions. The evaporation of the prompt neutrons and γ rays is calculated through the Hauser-Feshbach statistical theory, taking into account the competition between neutron and γ-ray emission, conserving energy, spin, and parity. The delayed emission is taken into account using time-independent calculation using decay data. This whole formulation allows for the calculation of prompt neutron and γ-ray properties, such as multiplicities and energy distributions, both independent and cumulative fission yields, and delayed neutron observables, in a consistent framework. Here, we describe the implementation of multi-chance fission into the HF3D model, and show an example of prompt and delayed quantities beyond first-chance fission, using the example of neutron-induced fission on 239Pu. This expansion represents significant progress in consistently modeling the emission of prompt and delayed particles from fissile systems. [ABSTRACT FROM AUTHOR]

Measurements of independent fission yield distributions in neutron-induced fission at high neutron energies are important for our fundamental understanding of the fission process, and are also relevant for reactor physics applications. So far, measurements of independent fission yields in proton-induced fission have been performed at the IGISOL facility at the University of Jyväskylä, using the Penning trap as a high resolving-power mass-filter. In order to also facilitate measurements of neutron-induced fission, a dedicated ion guide and a proton-to-neutron converter was developed. However, the first measurement indicates that fewer fission products than expected reach the Penning trap. To explore potential reasons and possible improvements, a simulation model was also developed and benchmarked. The benchmark showed that the model is able to reproduce the performance of the ion guide remarkably well and that the main reason for the low yield of fission products is the low collection efficiency of the ion guide. Based on the benchmark, a new ion guide is being designed. In the new design, the positions of the uranium targets and volume of the ion guide have been changed to increase the collection of fission products. This results in a five-fold increase of the yield. However, the collection efficiency of the new ion guide still needs to be improved in order to achieve intensities of the extracted fission products that are large enough to allow for reasonable measurement times. Because the volume of the ion guide is increased significantly, the extraction time of the ions is expected to be longer than that from the previous ion guide. Therefore, an electric field guidance system that consists of a combination of a stationary electric field and an RF-carpet is considered to be deployed. The stationary field, produced from a set of DC-ring electrodes, accelerates the ions towards the RF-carpet at end plate of the ion guide. The RF-carpet consists of a time-dependent field, produced from a radio-frequent structure of concentric rings, with a DC-component that guides the ions towards the exit hole in the center of the end plate. In this paper we present the current status of the simulations and design of the new ion guide. [ABSTRACT FROM AUTHOR]

At low incident neutron energies nuclear reaction cross sections exhibit a distinct resonance structure which cannot properly be described by (semi-)microscopic models. Usually R-matrix theory is applied which provides a suffi ciently accurate but phenomenological description of the resonance region. However, standard R-matrix theory is only suited for two-particle channels. Three- and many-particle channels which may occur in light nuclear systems even at rather low incident energies are usually treated in an approximative or effective way. In this contribution a novel R-matrix formalism for three-body channels and its application to nuclear systems will be presented. It is a generalized and significantly modified form of a proposal by W. Glöckle based on the Faddeev equations. [ABSTRACT FROM AUTHOR]

A survey has been conducted of the resonance data available in the major evaluated libraries for a range of mid-mass isotopes. Although many isotopes have data derived from experimental measurements and rigorous analyses, the data for some others, particularly for isotopes that are unstable, are purely theoretically generated, e.g. via the use of the TARES code. A subsequent project has developed a suite of automated tools to rapidly inspect the resonance data available for a specified isotope. These tools calculate a range of values and distributions for the available data (e.g. cumulative number of resonances versus neutron energy) and use a comparison against expected theoretical values (e.g. a Porter-Thomas distribution) to suggest their physical validity or otherwise. The motivation for this study, the tools developed, and results for a number of isotopes, are presented in this paper. [ABSTRACT FROM AUTHOR]

The β-delayed neutron emission in the 85As β-decay has been measured at the IGISOL facility of the Accelerator Laboratory of the University of Jyväskylä (JYFL). The complete β-decay has been studied with a setup which consists of a plastic scintillator, the MOdular Neutron time-of-flight SpectromeTER (MONSTER), and two types of γ-rays detectors (HPGe and LaBr3). The, β-delayed neutron energy distribution has been determined by unfolding the TOF spectrum with the iterative Bayesian unfolding method. [ABSTRACT FROM AUTHOR]

The neutron time-of-flight facility n_TOF is characterised by its high instantaneous neutron intensity, high-resolution and broad neutron energy spectra, specially conceived for neutron-induced reaction cross section measurements. Two Time-Of-Flight (TOF) experimental areas are available at the facility: experimental area 1 (EAR1), located at the end of the 185 m horizontal flight path from the spallation target, and experimental area 2 (EAR2), placed at 20 m from the target in the vertical direction. The neutron fluence in EAR2 is ˜ 300 times more intense than in EAR1 in the relevant time-of-flight window. EAR2 was designed to carry out challenging cross-section measurements with low mass samples (approximately 1 mg), reactions with small cross-sections or/and highly radioactive samples. The high instantaneous fluence of EAR2 results in high counting rates that challenge the existing capture systems. Therefore, the sTED detector has been designed to mitigate these effects. In 2021, a dedicated campaign was done validating the performance of the detector up to at least 300 keV neutron energy. After this campaign, the detector has been used to perform various capture cross section measurements at n_TOF EAR2. [ABSTRACT FROM AUTHOR]

The slow neutron capture process (s-process) is responsible for producing about half of the elemental abundances heavier than iron in the universe. Neutron capture cross sections on stable isotopes are a key nuclear physics input for s-process studies. The 72Ge(n, γ) Maxwellian-Averaged Cross Section (MACS) has an important influence on the production of isotopes between Ge and Zr in the weak s-process in massive stars and so far only theoretical estimations are available. An experiment was carried out at the neutron time-of-flight facility n_TOF at CERN to measure the 72Ge(n, γ) reaction for the first time at stellar neutron energies. The capture measurement was performed using an enriched 72GeO2 sample at a flight path length of 184 m, which provided high neutron energy resolution. The prompt gamma rays produced after neutron capture were detected with a set of liquid scintillation detectors (C6D6). The neutron capture yield is derived from the counting spectra taking into account the neutron flux and the gamma-ray detection efficiency using the Pulse Height Weighting Technique. Over 70 new neutron resonances were identified, providing an improved resolved reaction cross section to calculate experimental MACS values for the first time. The experiment, data analysis and the new MACS results will be presented including their impact on stellar nucleosynthesis, which was investigated using the post-processing nucleosynthesis code mppnp for a 25 solar mass model. [ABSTRACT FROM AUTHOR]

The prompt fission neutron spectrum (PFNS) from neutron-induced fission is a fundamental quantity for the behavior of nuclear reactors, and has been measured many times on a wide variety of nuclei and covering different ranges of incident and emitted neutron energies. However, results from past measurements are frequently called into question in modern nuclear data evaluations because of a lack of thorough experimental documentation and incomplete uncertainty analyses. The Chi-Nu experiment at Los Alamos National Laboratory was designed to produce high-precision measurements of the PFNS of major actinides over a wide range of incident and emitted neutron energies, and with the documentation and covariance analysis required to ensure that the results of this experiment maintain their impact long into the future, thereby avoiding this pitfall of past measurements. In this work we describe the Chi-Nu experiment along with summaries of the treatment of and methods developed to address two important components of the analysis of Chi-Nu data: random-coincidence backgrounds and MCNP simulations. Furthermore, we describe the first results for correlations not just between all data points collected on a single target nucleus, but also between all data points from separate Chi-Nu measurements on 235U and 239Pu. These correlations are important for accurately calculating ratios of the PFNS from one actinide to another, which are rare and can be informative for nuclear data evaluation efforts. [ABSTRACT FROM AUTHOR]

A new version of the XSUN-2022 computer code package is under preparation to be released through NEA Data Bank which includes the latest updates of the SUSD3D sensitivity/uncertainty (S/U) code and the corresponding cross section and covariance matrix libraries based on the JEFF-3.3, ENDF/B-VIII.0 and FENDL-3.2 evaluations. The XSUN code package is a user–computer interface environment, written in Xbase++ ® and using Compiler 1.90.331 and Alaska 32-Bit Linker, for user-friendly pre- and post-processing of the input and output data for a complete and self-consistent set of deterministic codes. The flowchart integrates the codes TRANSX, PARTISN, and SUSD3D, all available from the OECD/NEA Data Bank and RSICC. The codes are used for the preparation of multigroup nuclear cross-section, neutron and gamma transport calculations for criticality and shielding calculations and nuclear data sensitivity and uncertainty calculations, respectively. XSUN-2022 is the 3rd released version, after XSUN-2013 and -2017. [ABSTRACT FROM AUTHOR]

The present work focuses on the complexity of the neutron transport phenomenon, which throws its shadows on every physical system wherever neutron is produced or used. The Monte Carlo N-Particle Transport Code (MCNP) investigated flux perturbations in the neutron field caused by an absorber. The geometry of the present experiment was designed to reach an isotropic neutron field. The neutron source was a 241AmBe with physics dependent only on alpha-beryllium interaction as a physical constraint. The neutron production rate is independent of what happens after it is generated. The geometries have been designed to get a volume of uniform neutron densities within a spherical volume of radius 15 cm in every neutron energy group up to 1 MeV, with a slight central tendency of higher energy neutrons making the field isotropic. Absorbers of different dimensions were placed within the volume to investigate the field perturbation. Two types of neutron absorbers were used, indium and gold. Flux density inside and outside the absorber samples was determined, while the spatial neutron flux distribution produced by the AmBe source was taken as a reference. This study displayed that absorbers of various dimensions perturb the neutron field in a way that is dependent on the absorption and scattering cross-sections, particularly in the neutron resonance. Unlike the simple picture of reducing the number density of neutrons, the perturbation was found to influence the moderation of neutrons in the medium, significantly above 1 MeV. [ABSTRACT FROM AUTHOR]

The use of the Average Functional Group Approximation for self-shielding corrections at inelastic neutron spectrometers is discussed. By taking triptindane as a case study, we use the above-mentioned approximation to simulate a synthetic dynamic structure factor as measured on an indirect-geometry spectrometer, as well as the related total scattering cross section as a function of incident neutron energy and sample temperature, and the transmission spectra depending on the sample thickness. These quantities, obtained in a consistent way from the Average Functional Group Approximation, are used to calculate the energy-dependent self-shielding correction affecting the sample under investigation. The impact on the intensities of low-energy vibrational modes is discussed, showing that at typical experimental conditions the sample-dependent attenuation factor is about 15% higher compared to the correction at higher energies. [ABSTRACT FROM AUTHOR]

In this research, the effects of the constituent elements relative atomic mass of liver tissue and variations of incident neutron energy were considered. The liver tissue together with soft tissues, which dwell in front of it, were decomposed into their constituent elements and so were emitted by neutrons in radiotherapy. This research might be implemented for other tissues with respect to the therapy of cancer. The amounts of KERMA (KINETIC ENERGY RELEASED PER UNIT MASS) of the neutron were obtained in the tissue because of the collision between neutron and nucleus of fundamental elements of the body's tissue, and then some interesting results were concluded. The results showed that the foremost amount of KERMA (because of neutron emission) occurred within the fast neutron energy range 4-10 MeV and neutron energy 16 MeV. Besides, inclinations of the KERMA graphs were very alike and bore a resemblance for different elements, and the utmost KERMA arose for the elements that have a low atomic mass including hydrogen. Also, most values of KERMA get related to hydrogen thanks to the abundance of it. [ABSTRACT FROM AUTHOR]

The neutron capture cross section of 128Te was measured by the TOF method in the energy regions of 15 – 91 keV and 550 keV. The cross section was determined with accuracies of 4-7%. The results are compared with past experimental data and evaluations of JENDL-4. A large disagreement with the recent activation measurement at around 550 keV was observed. [ABSTRACT FROM AUTHOR]

The isotopes of Zr with A = [90, 91, 92, 94] make up more than 97% of naturally occurring Zr and are important to many nuclear applications such as nuclear reactors. One of the attractive qualities of naturally occurring Zr isotopes is that they have a low σγ/σt ratio at most neutron energies. Thus, they improve the neutron economy in reactors by preferentially scattering neutrons rather than absorbing them. This same quality also presents a challenge to measuring the capture cross section, σγ, of Zr isotopes. The ENDF/B-VIII.0 library has a relative uncertainty of approximately 10–20% for incident neutron energies < 0.1 MeV and an uncertainty greater than 20% for energies > 0.1 MeV for the majority of natural Zr isotopes. This motivated the Nuclear Criticality Safety Program to embark on a campaign to accurately measure and evaluate these Zr isotopes. In this work, we demonstrate energy-dependent neutron capture cross section measurements for the first enriched sample to be measured: 90Zr. [ABSTRACT FROM AUTHOR]

We report on the development of neutron-based techniques to non-destructively measure the composition of concrete. Previous experimental studies demonstrated the viability of the unfolding technique to determine the ratio of water, sand and cement in well-characterised concrete samples from the transmitted neutron energy spectrum. In this work, we used MCNP6 simulations to demonstrate the extension of the technique to determine elemental compositions from transmitted, or scattered neutron energy spectra. In both cases, the simulated energy spectra provided a reliable method to unfold the composition of samples with known elemental ratios. The precision of the technique was limited by the statistical uncertainties of the simulated spectra, particularly for the case of scattered neutrons. The accuracy of the technique was heavily dependent on the uniqueness of each of the elemental responses, and reasonable prior knowledge of the composition. Given the promising results at this stage, future developments will include the addition of further elements to the response matrix, and experimental verification. [ABSTRACT FROM AUTHOR]

SAMOP is a subcritical assembly for the Mo-99 isotope production. This isotope plays important role as a generator of Tc-99m isotope which is widely used for diagnostic in nuclear medicine. The SAMOP which is developed at PSTA BATAN uses uranyl nitrate solution UO2(NO3)2 as fuel material and also as target material in the process of Mo-99 production. This process is based on the fission reaction of uranium U-235 contained in the uranyl nitrate solution. The U-235 fission reactions occur as long as there are still neutrons from external neutron source and it will stop whenever the external neutron source is removed. SAMOP needs an external neutron source which can provide neutrons with an average neutron flux of 108 n/cm2s for maintaining the sustainability of U-235 fission reactions. Presently, SAMOP external neutrons need is provided by the neutrons from the radial beam port of the Kartini research reactor. For further development, SAMOP will be equipped with a compact neutron generator as its external neutron source. Compact neutron generator is a kind of neutron generator which has small size compared to conventional neutron generator. Neutron generator produces neutrons by the D-D or D-T fusion reaction. The neutron energy produced by the D-D and D-T fusion reactions are 2.5 MeV and 14.1 MeV respectively. Therefore, the feasibility study of axial compact neutron generator as SAMOP neutron source has been done. Axial compact neutron generator is a kind of compact neutron generator where the position of ion source and target in compact neutron generator is along the axis of the compact neutron generator. The results of the study show that the axial compact neutron generator has a chance to substitute a nuclear reactor as external neutron source of SAMOP. It can produce neutrons which neutron yield is sufficient to meet the neutron need of SAMOP. It depends on the type of the axial compact neutron generator, one of which is the product of The Lawrence Berkeley National Laboratory i.e. the axial compact neutron generator with fully shielded high voltage which can produce neutron yield up to 109 n/s from D-D fusion reaction. The benefit of using axial compact neutron generator as neutron source of SAMOP is that the operational cost will reduce since the neutron production cost using axial compact neutron generator is cheaper than a nuclear reactor. In addition, the operational procedure of axial compact neutron generator is simpler than a nuclear reactor. [ABSTRACT FROM AUTHOR]

The total kinetic energy (TKE) release in the fast neutron-induced fission of various actinide nuclei was measured for neutron energies for En =2.6-100 MeV at the Weapons Neutron Research facility of the Los Alamos National Laboratory. The data are compared to the GEF model of the fission process. The variances of the TKE distributions appear to decrease with increasing Z and A of the fissioning systems. [ABSTRACT FROM AUTHOR]

Since neutrons especially fast neutron has a high weighting factor in radiation dosimetry, exposing the human's body to a neutron source in a laboratory or clinic may put body tissues at risk and dangerous situations. In doing so, in this investigation, body tissue together with soft tissues (including adipose and skin) were decomposed into their fundamental elements, and then got emitted by neutrons. This research can be put into operation all tissues about cancer therapy. The amounts of neutron energy because of the collision between neutrons and nuclei of the basic elements got calculated in the tissue Next, some results got concluded. The results imply that the highest transferred energy happens within the energy range 3.5-10.5 MeV and also energy ranges 15-16.5 MeV. As well, the tendencies of the transferred-energy graphs were very in a similar way and are the same as other fundamental elements, and the maximum transferred energy happened for the elements that had a small atomic weight. [ABSTRACT FROM AUTHOR]

Recently Molten Salt Reactor (MSR) gains more interest among nuclear researchers due to its promising competitiveness and safety features. ThorCon MSR (TMSR500) is this kind of reactor projected to be built in Indonesia. Due to a limited number of proven operating reactors and limited knowledge of reactor characteristics, several aspects have to be studied, such as neutronic features, material, fuel circulation, fuel management, etc. This paper reveals some important neutronic safety features of MSR, referring to ThorCon design, such as temperature reactivity feedback of fuel, moderator, and reflector. Void reactivity feedback and neutron energy spectrum behavior during voiding were also examined by performing MCNP6 calculations and nuclear data ENDF/B-VII. Full core geometry was modeled, and the drift effect was not taken into account. Neutron flux and power density distribution were also addressed. The purpose of this paper is to review the aforementioned important safety parameters by doing the Monte Carlo code calculations. The results indicated that the core induces negative reactivity feedback due to temperature increase in fuel and moderator, but positive for temperature increase in the reflector. The temperature reactivity coefficient of fuel, moderator, and reflector exhibited −5.82×10−3 (%dk/k)/°C, −9.68×10−4 (%dk/k)/°C, and +1.84×10−4 (%dk/k)/°C, respectively. The introduction of void in the core either as a result of boiling or pump cavitation would induce positive reactivity of +0.115 (%dk/k)/(%void). Overall, the temperature reactivity coefficients exhibited large negative, and there is a comfortable margin between the operating temperature to the boiling point that might prevent void generation due to fuel boiling. [ABSTRACT FROM AUTHOR]

The variance reduction techniques are necessary for Monte Carlo calculations in which obtaining a detailed calculation result for a large and complex model is required. The GVR method named as global weight window generator (GWWG) was proposed by the FDS team. In this paper, two typical calculation examples, ISPRA-Fe benchmark in SINBAD (Shielding Integral Benchmark Archive Database) and TF Coils (Toroidal Field coils) of European HCPB DEMO (Helium Cooled Pebble Bed demonstration fusion plant), are used to study the performance of GWWG method. It can be seen from the calculation results that the GWWG method has a significant effect in accelerating the Monte Carlo calculation. Especially when the global convergence calculation results are needed, the acceleration effect (FOMG) can reach 105 or more. It proves that the GWWG method is an effective tool for deep-penetration simulations using Monte Carlo method. [ABSTRACT FROM AUTHOR]

This article focuses on description of two different neutron fields from linear and cylindrical Inertial Electrostatic Confinement (IEC) neutron generators. Both of these generators are well defined and commonly used. They use a deuterium-tritium reaction that produces neutrons with energies in the range 13 – 16 MeV, depending on the direction and the energy of the incoming deuterium nucleus. Two-parametric spectrometric system for neutron/gamma mixed fields NGA-01 was used to characterize neutron spectra in the proximity of generators. The cylindrical 45x45 mm stilbene scintillator was connected to this device using an active voltage divider. This way, we were able to measure neutron energies in the range 1 - 15 MeV while filtering out gamma radiation, even when counts per second is high. For the neutron spectrum calculation recoil spectra using deconvolution through maximum likelihood estimation was used. Measured neutron spectra have been compared with simulations realized via MCNP6. According to the theoretical prediction, these two types of generators produce different neutron fields. In case of the linear generator the target is very close to point located tritium bombarded by deuterons. Thus the neutron spectrum varies depending on the angle between the detector axis and the axis of the generator. Both experimental results and simulation show a shift of the neutron energy peak in pulse height histogram. For IEC type generators the neutron spectrum is more complicated. The shape and the position of the neutron energy peak depend heavily on the position of the detector. The most prominent effect is in the position in the plane perpendicular to the generator axis. In this case, the peak splits into two peaks that can be measured and distinguished. These results were verified by the diamond detector which was also used for characterization of the IEC type generator. [ABSTRACT FROM AUTHOR]

Thermal neutron scattering data have an important influence on the high-fidelity neutronics calculation of thermal reactors. Due to the limited storage capabilities of computers, a discrete ACE representation of the secondary neutron energy and angular distribution has been used for Monte Carlo calculation since the early 1980s. The use of this discrete representation does not produce noticeable effects in the integral calculations such as keff eigenvalues, but can produce noticeable deficiencies for differential calculations. A new continuous representation of the thermal neutron scattering data was created in 2006, but was not widely known. Recently, the continuous representation of the thermal neutron scattering ACE data based on ENDF/B-Ⅷ.0 library was officially released and was available for all users. The new representation shows great difference compared with the discrete one. In order to utilize the more physical and rigorous representation data for high fidelity neutronic-thermohydraulic coupling calculation, the on-the-fly treatment capability was proposed and implemented in RMC code. The two-dimensional linear-linear interpolation method was used to calculate the inelastic scattering cross sections and the secondary neutron energies and angles. The on-the-fly treatment capability was tested by a pressurized water reactor assembly. Results show that the on-the-fly treatment capability has high accuracy, and can be used to consider the temperature feedback in the neutronic-thermohydraulic coupling calculations. However, the efficiency of the on-the-fly treatment still need to be improved in the near future. [ABSTRACT FROM AUTHOR]

The simulation of independent and cumulative yields requires precise knowledge of the initial conditions of the fission fragments immediately after scission. In this paper, we use a quantum-mechanical description of fission dynamics to extract the initial mass distribution of fission fragments for the neutron-induced fission of the two major actinides 239 Pu and 235 U, both for thermal fission and as a function of the incident neutron energy. [ABSTRACT FROM AUTHOR]

Subcritical assembly for Mo-99 production (SAMOP) has been developed at The Center for Accelerator Science and Technology (CAST), National Nuclear Energy Agency (SATAN). The objective of the SAMOP development is to provide a prototype of a subcritical assembly which produces Mo-99 isotope. In the nuclear medical application, Mo-99 plays an important role as a generator of isotope Tc-99m which is the most widely used radioisotope for diagnostic purposes. SAMOP use uranyl nitrate solution UO2(NO3)2 as fuel as well as target material for the process of Mo-99 production. This process is based on the fission reaction of uranium U-235 contained in the uranyl nitrate solution. The U-235 fission reactions occur as long as there are still neutrons from external neutron source and it will stop whenever the external neutron source is removed. Presently SAMOP uses external neutron source from the radial beam port of the Kartini research reactor which emerges thermal neutrons in the order of 108 n/cm2s. For further development, SAMOP will be equipped with a compact neutron generator as an external neutron source. Compact neutron generator is a neutron source device that contains a compact ion accelerator and that produces neutrons by fusing isotopes of hydrogen. The fusion reactions take place in the device by accelerating either deuteron, triton, or a mixture of these two isotopes into a metal hydride target which also contains deuterium, tritium or a mixture of these isotopes. The fusion of deuterium atoms (D + D) results in the formation of a He-3 ion and a neutron with a kinetic energy of approximately 2.5 MeV. The fusion of a deuterium and a tritium atom (D + T) results in the formation of a He-4 ion and a neutron with a kinetic energy of approximately 14.1 MeV. SAMOP needs an external neutron source which can provide neutrons with an average neutron flux of 108 n/cm2s for maintaining the sustainability of U-235 fission reactions. The results of the study on compact neutron generator as an external neutron source of SAMOP show that a compact neutron generator has a chance to substitute a nuclear reactor as external neutron source of SAMOP. A compact neutron generator can produce fast neutrons which neutron yield is sufficient to meet the need of neutron for SAMOP. It depends on the type of the compact neutron generator, one of which is the compact neutron generator produced by LSNL i.e. the axial compact neutron generator which can produce neutron yield up to 109 n/s. The benefit of using a compact neutron generator as an external neutron source of SAMOP is that the operational cost will reduce since the neutron production cost using a compact neutron generator is cheaper than a nuclear reactor. In addition, the operational procedure of a compact neutron generator is simpler than a nuclear reactor. [ABSTRACT FROM AUTHOR]

The spinel compounds of CuMn2-xTixO4 with x = 0.0, 0.25 and 0.50 were prepared in polycrystalline form using solid state reaction method at 950°C. Their phase purity was confirmed as tetragonal distorted spinel with space group I41/amd by fitting the room temperature neutron diffraction (ND) patterns using Rietveld software. However, the lattice parameters of the unit cell in x = 0.25 and 0.50 changes significantly in comparison to CuMn2O4 upon Ti substitution. In addition, the magnetic field dependent magnetization behavior significantly altered particularly below the magnetic ordering temperature in all the derivatives. [ABSTRACT FROM AUTHOR]

Fission product yields (FPY) are one of the most fundamental quantities that can be measured for a fissioning nucleus and are important for basic and applied nuclear physics. Recent measurements using mono-energetic and pulsed neutron beams generated using Triangle Universities Nuclear Laboratory's tandem accelerator and employing a dual fission chamber setup have produced self-consistent, high-precision data critical for testing fission models for the neutron-induced fission of 235,238U and 239Pu between neutron energies of 0.5 to 15.0 MeV. These data have elucidated a low-energy dependence of FPY for several fission products using irradiations of varying lengths and neutron energies. This paper will discuss new measurements just beginning utilizing a RApid Belt-driven Irradiated Target Transfer System (RABITTS) to measure shorterlived fission products and the time dependence of fission yields, expanding the measurements from cumulative towards independent fission yields. The uniqueness of these FPY data and the impact on the development of fission theory will be discussed. [ABSTRACT FROM AUTHOR]

A hybrid position-sensitive neutron detector with active 10B-layer and proportional chamber filled by the gas mixture Ar+25%CO2 is being studied. The detector is used on the neutron source based at the linear accelerator. Recording two pairs of pulse heights which are produced in two gas gaps by 7Li and 4He nuclei between the anode and two cathodes gives a possibility to determine the X and Y neutron coordinates. This detector permits time-of-flight measurements. A significant difference in pulse height spectra for slow and fast (several MeV) neutrons was observed in experiment. In recent experiments with 10B-detector we found that (1) the pulse heights depend on the neutron energy and (2) the position of the maximum in the pulse height spectrum varies monotonically versus the maximum neutron energy. [ABSTRACT FROM AUTHOR]

Low-dimensional moderators were designed for the European Spallation Source (ESS), and the same concepts can be applied to compact sources. At ESS, quasi two-dimensional (2D) high-brightness moderators will serve all the instruments of the initial suite. The design of the moderators is influenced by several factors, such as the layout of the facility, the requirements for beam extraction, and the number of instruments; these constraints led to the choice of the 2D "butterfly" moderator for ESS. In an accelerator-based compact source, such moderators can be designed in an even more efficient way than for high-power sources, taking advantage of the lower heat loads and of the more compact arrangement of the target-moderator system, which gives more freedom in the optimization of the geometrical setup. Some promising design options have been explored. [ABSTRACT FROM AUTHOR]

In this contribution, we present the CoolGAL fast neutron production target system, to be used in the initial phase of the NEPIR irradiation beamline at the SPES facility, that will be operational in 2022. Initially, NEPIR will be used for shielding studies against fast neutrons for space applications and to investigate neutron-induced single event effects in microelectronic devices and systems. In CoolGAL, the neutron production component, a thick Be cylinder, is immersed in a static bath of Galinstan, a liquid alloy of Ga, In and Sn, contained by an outer water cooled copper cladding. MCNPX calculations indicate that, by using a 1 μA current of 70 MeV protons, it can produce a fast neutron energy spectrum that is somewhat flat the 30-65 MeV energy range and with a sharp cut-off at the beam energy. At the standard test point, located 2.6 m downstream from the source, the beam spot diameter, defined by the peculiar collimation scheme of the initial phase of NEPIR, is 10 cm and the integral fast neutron flux is Φn(1< En<65 MeV) ~ 3×106 n cm-2s-1. Using proton beams with two different energies, one can calculate, by subtraction, the effects due to the neutrons in the energy interval defined by the two cut-off values. Preliminary results of ANSYS calculations, for a 1 μA proton current of 70 MeV protons (70 W), show a limited regime temperature (28 °C) of the Be component, capable of ensuring the exceptional safety level required for the operation at SPES. The thermal reliability of CoolGAL is very promising for future developments that require higher proton currents. [ABSTRACT FROM AUTHOR]

Ad-hoc parametrized Tolman-Oppenheimer-Volkoff (PTOV) equations proposed by Velten, Oliveira, and Wojnar can be correctly derived from the Einstein field equation with a modified matter with anisotropic pressure and effective energy density. The advantage of the latter form of PTOV is we can derive other properties of neutron star beyond static and spherical symmetric properties such as moment inertia and tidal deformation. Here we show that the corresponding pressures, energy density, and speed of sound profiles of modified energy-momentum tensor obey the energy conditions. [ABSTRACT FROM AUTHOR]

The results of experiments performed with a Ø25 x 25 mm solution-grown stilbene crystal in mono-energetic neutron fields in the 80-to-230 keV energy range are presented. The goal of the measurements, performed at the AMANDE facility, was to explore the capabilities of this organic scintillator to measure neutrons at the lowest possible energy with good pulse shape discrimination (PSD). The time of flight (TOF) technique was used in order to help with the neutron-gamma discrimination. The data are collected via a programmable digital acquisition (DAQ) system CAEN DT7530 using the software CoMPASS with the charge comparison method (CCM). The data are analysed using post-processing codes developed in the ROOT environment. The results show that the stilbene detector has discrimination capabilities for energies as low as 80 keV. [ABSTRACT FROM AUTHOR]

A standard theoretical neutron energy flux distribution is achieved for the triton-triton nuclear fusion reaction in the range of triton energy about ≤10 MeV. This distribution give raises an evidence to provide the global calculations including the characteristics fusion parameters governing the T-T fusion reaction. [ABSTRACT FROM AUTHOR]

The structural materials of Indonesian Experimental Reactor (RDE) is made from graphite that dominate material used on core structure. So that graphite material is very important role, both as core structure material, reflector and also as a moderator and fuel layer and fuel matrix. In thermal neutron energy range, the neutron scattering collision in moderator material such as graphite influences the neutron cross-section and the resulting energy distribution, so that neutrons will get an increase in energy for excitation in the material. Due to high neutron absorption cross section, boron and its compounds find extensive application in the nuclear industry. In actually it is very difficult to obtain pure uranium or thorium dioxide without any other substance. Usually uranium dioxide or thorium kernel always has impurity material like boron. Boron is one of the materials that has strong neutron absorber, specially for Boron-10. The research starting from modeling of kernel TRISO coated fuel particle, spherical pebble fuel and full core modeling by involving multiple heterogeneity calculations. Boron impurities in the TRISO kernel coated fuel particles was carried out with 27 data varied concentration of boron are 0ppm, 1ppm, 2ppm, 3ppm, 4ppm, 5ppm, 6ppm, 7ppm, 8ppm, 9ppm, 10ppm, 15ppm, 20ppm, 25ppm, 30ppm, 35ppm, 30ppm, 35ppm, 40ppm, 45ppm, 50ppm, 60ppm, 70ppm, 80ppm, 80ppm, 90ppm and 100ppm. All calculation analysis will be done using Monte Carlo MCNP6 with continuous neutron energy cross section taken from ENDF/B-VII file. Investigation of multiplication factor effect due to thermal neutron scattering crossing data S(α,β) for graphite and boron impurities on TRISO UO2 or ThO2 kernel coated fuel particle, spherical pebble fuel and full core calculation will be conducted. The all calculation results of the criticality calculation due to effect of boron impurity for both for UO2 and ThO2 kernel coated fuel particles are clearly showed that there are no significant influences effect on multiplication factor value. While criticality calculations using the S(α,β) option for UO2 and ThO2 kernel fuels give the results of a slightly lower multiplication factor with a maximum percentage difference is below than 1,3% for the calculation of the effective multiplication factor on the full core calculation. [ABSTRACT FROM AUTHOR]