Search

Several intense energy irradiation testing facilities are functional to study the effects of radiation damage in fusion materials. Tungsten is a potential candidate material to build the plasma-facing armour components of fusion reactors. Therefore, it is vital to estimate the radiation-induced primary damage in the material, particularly due to the large amount of transmutation nuclides produced in the high-energy neutron sources. The basic evaluated nuclear data libraries provide the essential neutron reaction data to estimate the transmutations and their effects during irradiation. However, much of the important high-energy neutron reaction data are often not explicitly provided to directly study their impact on radiation damage. Such inexplicit nuclear reaction data, especially for the transmutation nuclides, can have a significant effect on the estimation of primary radiation damage. A new specific methodology has been developed to unravel information from these inexplicit nuclear reaction data from the basic ENDF-6 libraries and use them with explicit data for estimating the primary radiation damage. The energy distributions of transmuted nuclides (the recoil nuclei in general) and light charged particles produced as a result of several types of neutron–nucleus interactions in tungsten, are quantified using the cross-sections from ENDF / B-VIII.0 and TENDL-2019 basic evaluated nuclear data libraries and their inter-comparisons are carried out. The neutron-induced atomic displacements and heating in tungsten are estimated for irradiations under the fusion and IFMIF-DLi neutron spectra. It shows that the transmutation reactions under these neutron spectra can sometimes contribute more than 50% of the total primary damage, most of which can be from inexplicit nuclear reaction data. Significant differences between the two sources of basic evaluated nuclear data libraries, especially in the production of transmutation nuclides and consequently, in their contributions to damage are observed. The present study helps to illustrate the importance of inexplicit nuclear reaction data and the usefulness of the ENDF-6 libraries for estimating primary radiation damage due to transmutations in plasma-facing materials. [ABSTRACT FROM AUTHOR]

Hadron therapy (HT) is a technique that uses accelerated ions, notably protons, and carbon ions, to destroy tumors. It has demonstrated high success rates in the treatment of certain cancers resistant to irradiation. In HT, the majority of the dose is delivered to the tumor volume by electromagnetic interactions with atomic electrons. However, it is important to note that the primary particles used in HT can induce nuclear reactions, generating undesirable secondary radiation, certainly neutrons. As a result, a significant portion of the patients' body may be exposed to the secondary background radiation field. Thus, these unwanted secondary neutrons should be evaluated. The purpose of this research in first is to characterize the secondary neutron production (SNP) during the administration of Hadrons with energies 140 MeV proton, and 264 MeV/u carbon ions ( C) in a soft tissue phantom. Second, comparing the neutron spectrum for different angles for protons and carbon ions for carbon targets to experimental data. Our results show that the neutron spectra (thick target neutron yields TTNYs) at energies less than 20 MeV, using the INCL4.6/GEM models and JQMD/GEM across a wide range would be suitable for the Monte Carlo transport simulation. [ABSTRACT FROM AUTHOR]

In the present work, the neutron yield of a Pu-Be neutron source has been calculated by using two methods, calculation with the SOURCES4C code developed at LANL and via an equation, respectively. The present work aims to determine and compare the two methods in obtaining the yields and spectra of an (a,n) type neutron source. The source used for this work is being held at IFIN-HH, and was well experimentally characterized in previous work. The results of the present work compare well with the experimental data. [ABSTRACT FROM AUTHOR]

Featured Application: Numerical modelling of yields and distributions of nuclear fusion products in magnetically confined plasma. Velocity distributions of fuel nuclei enter the formulae for distributions of products of fusion reactions in plasma. The formulae contain multiple integration, which is a computationally heavy task. Therefore, simplifications of the integrand are advantageous. One of possible simplifications is the use of closed-form analytical distributions of fast deuterons and tritons, accounting for slowing down and pitch-angle scattering and neglecting the speed diffusion. The plausibility of such a model has been studied from the viewpoint of its influence on the calculated spectra of fusion neutrons. Calculations have shown that the speed diffusion effect on suprathermal ion distribution tails does not significantly alter the qualitative behaviour of energy and angle distributions of fusion products in a beam-heated plasma. [ABSTRACT FROM AUTHOR]

The neutron spectrum was measured at two locations in the spent fuel storage facility of the Temelín nuclear power plant. The measurement had two primary objectives: to map the neutron -γ field by quantifying the ambient dose equivalent H∗(10) and to identify methods that could improve the quality of the adjusted neutron spectrum using a Bonner Sphere Spectrometer (BSS). Three spectrometers were used: a BSS and two proton recoil spectrometers. Hydrogen-filled proportional counters and an EJ309 scintillator were used to construct the a priori spectrum for BSS adjustment. The details of this process and its results are discussed. The a posteriori spectrum was used to calculate the ambient dose equivalent H∗(10). The resulting spectrum is highly thermalised, but the predominant contribution to H∗(10) was in the 100 keV-1.3 MeV range. The use of hydrogen-proportional counters in combination with the BSS is recommended. • Neutron spectra at spent fuel storage were measured by three spectrometers. • Ambient dose equivalent was calculated from the neutron spectra. • The neutron spectrum is highly thermalised, but most of the ambient dose equivalent is from the 100 keV-1.3 MeV energy range. • Due to the prevalence of contribution H∗(10) in range 100 keV-1.3 MeV use of hydrogen proportional counter spectrometer is recommended. [ABSTRACT FROM AUTHOR]

The nuclear physical properties of austenitic chromium-nickel steel EK-164 (Fe–16Cr–19Ni–2Mo–2Mn–Nb–Ti–B) and its manganese-based modifications EK-164Mn (Fe–16Cr–20Mn–2Mo–Nb–Ti–B) and EK-164MnW (Fe–16Cr–20Mn–2W–Nb–Ti–B) under irradiation (up to 5 effective years) in the neutron spectra of fast power (BN-600) and thermonuclear (DEMO-C) reactors and after irradiation (nuclear cooling up to 1000 years) were determined. The investigations were based on using the ACDAM-2.0 calculation complex with consideration of the elemental (alloying and impurity) compositions of steels and typical neutron spectra of fission (BN-600) and fusion (DEMO-C) reactors. The entire set of nuclear physical characteristics of steels (primary radiation damage; transmutation of elements; radiogenic accumulation of elements, including hydrogen, helium, and boron; γ-dose rate; radioactivity; nuclear energy release; isotope contributions) is obtained and their comparative analysis is performed. The time intervals for fulfilling the criteria of low activation of steels after irradiation (γ-dose rate less than 10 mSv/h, radioactivity less than 1010 Bq/kg, energy release less than 10 W/m3) were determined. [ABSTRACT FROM AUTHOR]

Velocity distributions of fuel nuclei enter the formulae for distributions of products of fusion reactions in plasma. The formulae contain multiple integration, which is a computationally heavy task. Therefore, simplifications of the integrand are advantageous. One of possible simplifications is the use of closed-form analytical distributions of fast deuterons and tritons, accounting for slowing down and pitch-angle scattering and neglecting the speed diffusion. The plausibility of such a model has been studied from the viewpoint of its influence on the calculated spectra of fusion neutrons. Calculations have shown that the speed diffusion effect on suprathermal ion distribution tails does not significantly alter the qualitative behaviour of energy and angle distributions of fusion products in a beam-heated plasma.

Two organic scintillators (EJ309 and stilbene) with their different detection chains and elaborate software were used to measure neutron spectra of AmBe and PuC sources. Method of shadow cone ensures minimization of scattered neutron influence. To separate neutrons and γ, a modern method of pulse shape discrimination was employed using fast digitizers. Calibration was performed by γ sources and three different methods for the identification of the Compton edge. Spectrometric systems also used different light output functions and unfolding methods. The same spectra were calculated using the MCNP6 code and the JENDL-5 library. The measured and calculated results were mutually compared, and also with the published results available in the literature. Generally, the results agree on the main peaks of neutron spectra but differ in their abundance. The lower limit for the ability to distinguish between neutrons and γ in spectrometry with organic scintillators was found at about 1.5 MeV. • Neutron spectra of AmBe and PuC sources were measured and calculated. • The measured/unfolded spectra and calculated ones agree in shape. • Methods of energy calibration and neutron/gamma separation are described. [ABSTRACT FROM AUTHOR]

FANT (Fuente Ampliada de Neutrones Térmicos ; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, which has been developed by the Neutron Measurement Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid (LMN-UPM). In FANT, an isotopic neutron source (241Am/9Be) produces the primary neutrons. The design and facility optimization were carried out by extensive Monte Carlo calculations. In addition, Monte Carlo methods were used to evaluate the facility's performance to produce a constant and uniform thermal neutron field; these results were validated through experimental methods. FANT is designed to have two neutron sources; the objective of this work is to estimate the ambient dose equivalent due to neutrons and gamma-rays by Monte Carlo methods, and to compare these values with measured experimental doses. Thus, the performance of FANT with the two 241Am/9Be sources of LMN-UPM, with regard to the ambient dose equivalent H*(10) produced by both neutrons and photons around the facility, is analyzed in this work. The results are compared with those previously obtained in the framework of the results obtained with the LB6411 device around FANT. • MCNP benchmarking with experimental H*(10) measurements: γ-rays and neutrons. • H*(10) around thermal neutron irradiation facilities: total, γ-rays and neutrons. • Berthold LB6411 and LB1236 area monitors. • Radiological protection assurance around thermal neutron irradiation facilities. [ABSTRACT FROM AUTHOR]

A methodology for designing a Bonner sphere spectrometer was proposed based on the singular value decomposition (SVD) and information entropy theory. First, features of Bonner spheres were classified based on SVD, and then, the rationality of the selected spheres was analyzed. Second, the information entropy is used as the quantitative evaluation standard; combined with the unfolding method, the contribution of each Bonner sphere on the information entropy of the unfolded spectra was quantitatively analyzed. The spheres that have high influence on the information entropy of solution spectra can be identified, and then, the information entropy variation of the solution spectra can be suppressed by reducing the count fluctuation of these spheres. According to the proposed method, an optimized spectrometer was developed and applied to measuring 252Cf spectra. [ABSTRACT FROM AUTHOR]

Results of experimental verification of results on neutron inelastic scattering cross section for iron obtained in measurements of spectra of neutrons undergoing inelastic scattering at incident neutron energies of 6.0, 7.0 and 8.0 MeV and their calculations within the framework of statistical theory of nuclear reactions and direct interactions are presented. Description is given of the methodology of measurements and simulation calculations. Spectra of neutrons undergoing inelastic scattering obtained for iron in the process of experiments are analyzed in comparison with calculated data. New measurements of neutron inelastic scattering spectra and their analysis within the framework of contemporary modeling understanding allowed formulating proposal on introduction of corrections in the indigenous BROND-2.2 library of recommended evaluated neutron data and insignificant corrections in the most recent version of the library BROND-3.

A new approach based on Prompt Gamma-ray Neutron Activation Analysis (PGNAA) was proposed for measuring neutron energy spectra. A system, which consisted of a high purity germanium (HPGe) detector and a multi-layer sample, was used to validate the method. The composite layers of the sample comprised of lead, polymethyl methacrylate, sodium chloride powder and borated polyethylene. Prompt gamma-rays emitted from Pb, H, O, C, Cl and B were selected to determine the incident neutron energy spectrum. The response function for the production of gamma rays from these elements by monoenergetic neutrons was obtained through Monte Carlo simulations. Then the neutron fields of Cf-252 neutron source and D-T neutron generator were investigated with this approach. With the measured prompt gamma-rays and the response matrix, the neutron spectra were obtained through the PSO algorithm. The unfolded results have a good agreement with the simulation spectrum. [ABSTRACT FROM AUTHOR]

Simulated workplace neutron fields of a 241 Am–Be source moderated by polyethylene spheres have been developed. The characterization in terms of neutron fluence rate and neutron ambient dose equivalent rate, H ∗ (10) , was performed using Bonner sphere spectrometer and the MAXED and FRUIT unfolding codes. The neutron fluence rates and the H ∗ (10) rates obtained from the two codes are in good agreement with the discrepancies of 8% and 6%, respectively. The neutron ambient dose equivalent-averaged energies and the neutron fluence-to-dose equivalent conversion coefficients were also evaluated in comparison between the two codes. The discrepancies are < 10% and 12%, respectively. [ABSTRACT FROM AUTHOR]

At the Nuclear Physics Institute (NPI) of Czech Academy of Sciences, the d+Be source reaction was investigated for thick beryllium target (thickness of 8 mm) and 15 MeV deuteron beam extracted from the U-120M isochronous cyclotron. For neutron field measurement of the d(15)+Be source reaction in close source-to-sample distance (non-point-like geometry), the multi-foil activation technique was utilized. Sets of nine activation materials (Au, Co, Ti, In, Al, Y, Fe, Ni, Nb) were irradiated by neutrons from the d(15)+Be source, and activated dosimetry foils were analysed by means of the gamma-ray spectrometry method using HPGe detector. From obtained reaction rates, the broad neutron spectra were reconstructed employing the modified version of SAND-II unfolding code. New intensive neutron field of broad spectrum up to 20 MeV and intensity of 3 × 10 10 cm − 2 s − 1 (close to the target) was successfully developed. The study of d(15)+Be source reaction provided new spectral data, and obtained neutron field extends the utilization of cyclotron-based fast neutron generators at the NPI and provides new experimental opportunities for future intensive irradiation experiments such as fast neutron activation analysis and radiation hardness tests of electronics and materials for nuclear energetics and aerospace industry. • Development of Accelerator-driven neutron sources and neutron fields. • Fast neutron spectrometry by multi-foil activation technique. • Nuclear data measurement and integral validation. [ABSTRACT FROM AUTHOR]

Self-Powered Neutron Detectors (SPNDs) are currently used in reactors’ environment to sense the magnitude of neutron-fluxes, usually for spatial-distribution mapping of the fuel region as to optimize burn-up strategies. During the development of tailored instrumentation for Lead-cooled Fast Reactors, the possibility to perform online spectral deconvolution of fast neutron-fluxes was recognized. Seven geometrically similar SPNDs with different neutron-sensitive materials, have been characterized by the Monte Carlo transport code MCNPX. Thanks to a database of spectral sensitivities vs. neutrons’ energy, a mathematical deconvolution process from 7 electric current values virtually measured by SPNDs was started, retrieving spectral information, in terms of 7 energy windows, of the neutron-flux the detectors are subjected by. This paper describes the procedure that led to those results, prefiguring future development to improve proposed method. [ABSTRACT FROM AUTHOR]

A well-defined neutron spectrum is essential for many types of experimental topics and is also important for both calibration and testing of spectrometric and dosimetric detectors. Provided it is well described, such a spectrum can also be employed as a reference neutron field that is suitable for validating selected cross sections. The present paper aims to compare calculations and measurements of such a well-defined spectra in geometrically similar cores of the LR-0 reactor with fuel containing slightly different enrichments (2%, 3.3% and 3.6%). The common feature to all cores is a centrally located dry channel which can be used for the insertion of studied materials. The calculation of neutron and gamma spectra was realized with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Only minor differences in neutron and gamma spectra were found in the comparison of the presented reactor cores with different fuel enrichments. One exception is the gamma spectrum in the higher energy region (above 8 MeV), where more pronounced variations could be observed. [ABSTRACT FROM AUTHOR]

The thermal neutron irradiation device (FANT), developed at the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid, is a high-density polyethylene regular parallelepiped, with a rather uniform neutron fluence inside its irradiation chamber. It uses a A m 95 241 / B e 4 9 neutron source aiming to provide thermal neutron fluence rates. Neutron spectra and neutron fluences were estimated with Monte Carlo methods in the FANT irradiation chamber when a C f 98 252 neutron source is used and were compared with the results obtained with the A m 95 241 / B e 4 9 source. Regardless of the neutron source, the largest contribution is due to thermal neutrons, producing also epithermal and fast neutrons. Per neutron emitted by the source, the use of C f 98 252 produces a larger amount of neutrons. • 241Am/9Be neutron source. • Neutrons sources 241Am/9Be and 252Cf are moderated and scattered in the system. • 252Cf neutron source. • MCNP6.2 [ABSTRACT FROM AUTHOR]

An external neutron source is necessary to start up and run a subcritical thermal nuclear reactor. The external neutron source can be a neutron generator, a radionuclide having spontaneous fission or an isotopic neutrons source. Turning off the neutron generator or removing the neutron source shuts down the reactor. In subcritical reactors for teaching, fuel tubes and items for irradiation can be inserted or removed even when the reactor is running because are radiological safe. The aim of this work was to estimate, using Monte Carlo methods, the neutron and γ-ray spectra at five sites outside a Subcritical thermal neutron reactor Nuclear Chicago model 9000 using water and polyethylene moderators. In addition, the ambient dose equivalent and the ambient dose, for neutrons and γ-rays were calculated. For neutrons, the ambient dose equivalent were calculated with the neutron fluence-to-ambient dose equivalent coefficients from the ICRP 74 and the neutron fluence-to-ambient dose response of the neutron area monitor LB 6411; while, the ambient dose was calculated using the ICRU 95 neutron fluence-to-ambient dose coefficients. For γ-rays, was calculated the ambient dose equivalent using the ICRP 74 photon fluence-to-ambient dose equivalent coefficients; while, the ambient dose was calculated with the photon fluence-to-ambient dose coefficients from the ICRU 95. Outside the reactor, neutron spectra have fast, epithermal and thermal neutrons, and gamma ray spectra have γ rays induced in the inelastic scattering and capture of neutrons with the reactor materials. Neutron and γ-ray fluence rates and dose rates are larger in the water-moderated reactor. The ambient dose rates, due to neutrons and γ-rays, are smaller than the ambient dose equivalent, because the ambient dose resembles the maximum effective dose. • Using Monte Carlo methods were estimated the neutron and γ-ray spectra outside a subcritical nuclear reactor. • The H*(10) and the new H* dose rates were calculated. • For neutrons, the H*(10) dose rates were estimated using the dose response of the LB 6411 neutron area monitor. • Previous results based on, still in use, H*(10) are larger than H*, aimed to evaluate the effective dose. [ABSTRACT FROM AUTHOR]

Evaluation of neutron fluence in a reactor pressure vessel (RPV) together with surveillance specimen programs for RPV materials are one of the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of RPV residual lifetime. This paper covers transport of neutrons through the RPV of a VVER-1000 nuclear reactor. This problem is of increased importance as it concerns issues around VVER NPP life extension. With regards to the construction (reduced thickness of the lateral reflector), this issue plays greater role in VVER reactors than in Western types of PWR reactors. RPV material degradation depends mainly on neutron flux and spectra. Both quantities can be calculated or measured. This paper compares MCNP calculations and measurements on zero-power experimental reactor LR-0 with TORT calculations for VVER-1000. The goal is to find a reasonable method for precise estimation of neutron fluence and attenuation factor through the RPV. The calculations were performed with MCNP stochastic code and TORT deterministic transport code. The measurements were performed in a VVER-1000 mock-up placed in reactor LR-0 (Research Center Řež). [ABSTRACT FROM AUTHOR]

Neutron fluxes and spectra in the horizontal experimental channels of reactor WWR-M of the Institute for Nuclear Research, National Academy of Sciences of Ukraine (Kyiv) have been calculated using neutron transport Monte Carlo model with fuel enriched in 235U both to 36 and 19.7 %. It is shown that at the very beginning operating with low-enriched fuel, when the reactor core is 28 % filled with "fresh" fuel assemblies, and the remaining cells are filled with beryllium displacers, there is a significant change in the parameters of neutron beams. However, after the reactor will begin to operate at its usual mode, that will be after completing all or most part of the core with fuel assemblies partially burnt out, spectra and fluxes in channels will restore most of their previous values. Some differences are mainly due to changes in composition of the core – the removal of two voluminous vertical water channels located within the core. The work can be helpful to experimenters