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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

The neutron spectra measurement is important for larger fusion devices because it can provide ion temperature diagnostics independent of ion density. In this work, the neutron spectrum in the thermonuclear reaction is analyzed with a different approach, is approximately determined by the Maxwell distribution of the reaction particle, and is independent of the reaction cross section. [ABSTRACT FROM AUTHOR]

This paper deals with the improvements of the linear artificial neural network unfolding approach aimed at accurately determining the incident neutron spectrum. The effects of the transfer functions and pre-processing of the simulated pulse height distributions from liquid scintillation detectors on the artificial neural networks performance have been studied. A better energy resolution and higher reliability of the linear artificial neural network technique have been achieved after implementation of the results of this study. The optimized structure of the network was used to unfold both monoenergetic and continuous neutron energy spectra, such as the spectra of 252Cf and 241Am-Be sources, traditionally used in the nuclear safeguards experiments. We have demonstrated that the artificial neural network energy resolution of 0.1 MeV is comparable with the one obtained by the reference maximum likelihood expectation-maximization method which was implemented by using the one step late algorithm. Although the maximum likelihood algorithm provides the unfolded results of higher accuracy, especially for continuous neutron sources, the artificial neural network approach with the improved performances is more suitable for fast and robust determination of the neutron spectra with sufficient accuracy.