Contribution Of Neutrons Research Articles

Dry cask radiation shielding validation and estimation of cask surface dose rate with MAVRIC during long-term storage

The dry storage cask became a semi-permanent solution for spent nuclear fuel since no permanent solution has been finalized over the past three decades. Considering the difficulty of deep geological repository projects in many aspects, prolonging the service life of dry casks longer than a few decades has become an option for spent nuclear fuel storage. It must be ensured that dry cask safety objectives, including radiation shielding, criticality safety, structure health, decay heat removal, and containment be fulfilled during the prolonged service life. This paper presents numerical modeling results of the radiation dose rate distribution on a TN-32 cask surface aged up to 300 years. The numerical modeling method was validated with measurement data, and the difference between the calculated results and measurement data in the total dose rate is 7%. The detailed radiation distribution variations over cask surfaces along the axial and radial directions for 300 years of storage were obtained, and their azimuthal angular dependence was established. The maximum total dose rate over the cask surfaces is less than 10 mrem/hour (9.258 ± 0.212 mrem/hour) after 40 years, and this outcome fulfills all dose rate requirements from 10 CFR 71.47(b). After 300 years of storage, the average total dose rates on the side surface and top surface are 0.150 ± 0.007mremmrem/hour and 0.057 ± 0.003mremmrem/hour, respectively, while the maxima are 0.601 ± 0.014 mrem/hour and 0.112 ± 0.006 mrem/hour, respectively. The variations in the fuel gamma ray, structure gamma ray, neutron and secondary gamma-ray contributions to the total dose rate are presented and analyzed in this work. The neutron and gamma-ray dose rate contributions by energy group are presented and analyzed at the cask top, mid-plane, and bottom at the selected storage times.

Study of $\gamma$-ray background from cosmic muon induced neutrons

The contribution to $\gamma$-ray background from secondary neutrons, originating from cosmic muon interactions in Pb-Cu composite shield, has been measured via $(\mathrm{n},\mathrm{n}^{\prime} \gamma)$ reactions in Cu and Ge. The minimization of background plays a key role in improving the sensitivity of rare event experimental searches like neutrinoless double beta decay and neutron background is often a major concern. It is important to understand secondary neutron production from cosmic muons, especially in materials like Cu and Pb, which often form a part of the detector or shields. The direct contribution of fast neutrons generated from cosmic muon interactions to the $\gamma$-ray background via $(\mathrm{n},\mathrm{n}^{\prime} \gamma)$ reactions is investigated for the first time. Measurements are carried out in the low background HPGe detector setup, TiLES, using Pb-Cu composite shield as target material for muon interactions. Simulations have been carried out with GEANT4.10.00 and GEANT4.10.05, each with two different Physics lists. The observed intensities of $(\mathrm{n},\mathrm{n}^{\prime} \gamma)$ for Cu are significantly under-predicted by GEANT4.10.00, while GEANT4.10.05 shows good agreement with the experimental data.

Assessment of the neutron radiation field with activation foils and intermittent irradiations around a PETtrace biomedical cyclotron

This work aimed at assessing the neutron radiation field with activation foils around of the PETtrace biomedical cyclotron. Activation foils of gold, gold covered with cadmium, indium and nickel was used to evaluate the thermal, epithermal and fast neutron contributions, respectively. The irradiations of the activation foils were performed using the intermittent irradiation method to conciliate the measurements with the routine 18F production. The evaluated points showed the contributions of the thermal, epithermal and fast neutrons due to interaction with the vault room, cyclotron accelerator and main cyclotron components.

Activation of Mn, Li2O and LiF in the JSI TRIGA reactor to study potential tritium production monitors for fusion applications

Two series of measurements were performed in the JSI TRIGA research reactor in 2014 and 2017 to validate the 55Mn(n,γ)56Mn cross-sections and experimentally investigate the relationship between the 55Mn(n,γ)56Mn reaction and the rate of tritium production through the 6Li(n,t)4He reaction. Indeed, previously observed similarities between the sensitivity profiles of the neutron reaction of tritium production on lithium, 6Li(n,t)4He, and those of the 55Mn(n,γ)56Mn reaction in tritium breeder modules indicated that the latter reaction could be used as an effective monitor of tritium production, at least for short-term monitoring (the half-life of 56Mn being 2.579 h). However, experimental verification, improvements and validation of the 55Mn(n,γ)56Mn cross-sections are needed in order to meet the required accuracy. Foils of certified reference material Al–1%Mn, as well as LiF thermoluminescent detectors and Li2O samples were irradiated, both bare and under cadmium, to study the potential use of the 55Mn(n,γ)56Mn reaction for monitoring tritium production in fusion devices. Additionally, Al–0.1%Au was also irradiated for comparison, the 197Au(n,γ)198Au reaction cross-section being a standard. In order to obtain complementary information for data validation purposes, the irradiations were performed in positions within the JSI TRIGA reactor with different neutron spectra, i.e. in the central channel, the pneumatic tube and the F19 position, both in the outer ‘F’ ring of the reactor core and in the IC-40 irradiation channel located in the graphite reflector surrounding the reactor core. Bare and cadmium-covered irradiations were needed to subtract the contribution of epithermal neutrons to the 55Mn(n,γ)56Mn reaction. Calculations of the reaction rates were performed using the Monte Carlo code MCNP6.1 with a detailed model of the JSI TRIGA reactor, with the samples, the irradiation capsules and covers being modelled explicitly. The uncertainties involved in the measurements and the calculations were carefully evaluated. The principal objective was to study the energy response and correlations between the 55Mn(n,γ)56Mn reaction in irradiated Al–1%Mn and the 6Li(n,t)4He reaction in irradiated LiF and Li2O. Good consistency between the measured and calculated 55Mn(n,γ)56Mn and 197Au(n,γ)198Au reaction rates, in most cases within the uncertainty bars, was observed.

Human Transcriptomic Response to Mixed Neutron-Photon Exposures Relevant to an Improvised Nuclear Device.

In the possible event of a detonation of an improvised nuclear device (IND), the immediate radiation would consist of both photons (gamma rays) and neutrons. Since neutrons generally have a high relative biological effectiveness (RBE) for most physiological end points, it is important to understand the effect that neutrons would have on the biodosimetry methods that are being developed for medical triage purposes. We previously compared the transcriptomic response in human blood after neutron and photon irradiation. In this study, we analyzed the effect of mixed-field-neutron-photon radiation on gene expression responses in human peripheral blood, to elucidate the neutron contribution in the setting of a radiation exposure from an IND detonation. We used four combinations of mixed neutron-photon exposures, with increasing percentages of neutrons, to a cumulative dose of 3 Gy. The mixed-field exposures consisted of 0%, 5%, 15% and 25% of neutrons, where 0% corresponds to 3 Gy of pure X rays. A maximum neutron exposure, corresponding to 83% neutrons (0.75 Gy) was also used in the study. Increases were observed in both the number and expression level of genes, with increasing percentages of neutrons from 0% to 25% in the mixed-field exposures. Gene ontology analysis showed an overall predominance of TP53 signaling among upregulated genes across all exposures. Some TP53 regulated genes, such as EDA2R, GDF15 and VWCE, demonstrated increased expression with increasing neutron percentages in mixed-field exposures. Immune response, specifically natural-killer-cell mediated signaling, was the most significant biological process associated with downregulated genes. We observed significant suppression of T-cell-mediated signaling in mixed-field exposures, which was absent in the response to pure photons. In this first study investigating gene expression in human blood cells exposed to mixed neutron-photon fields similar to an actual IND explosion, we have identified a number of genes responding to the 3 Gy dose that showed increasing expression as the neutron percentage increased. Such genes may serve as better indicators of the expected biological damage than a report of total physical dose, and thus provide more relevant information for treating physicians.

Evaluation of boron-enriched plastic scintillator for thermal neutron detection

In recent years, we could observe significant progress in the design of accelerator-based neutron sources for boron neutron capture therapy (BNCT) . An important component of such systems is a neutron-sensitive detector system capable of monitoring the patient dose at significantly higher particle rates than in nuclear reactor systems. Simultaneous operation of two detectors: the first, sensitive to gammas, and the second, sensitive to gammas and neutrons, enables more accurate extraction of the neutron contribution. Scintillator-based detectors with fiber-optic readout have several important advantages. They are compact and have high speed, and additionally, the readout electronics can be placed far from the neutron source. In this research, we evaluated the main parameters of two plastic scintillators that are produced in Russia and meet this goal: SC-301 and boron-enriched SC-331. The resulting light yield of SC-331 was 8600 photons/MeV, and for SC-301 it was approximately 25% higher. Therefore, these scintillators are promising candidates for future detector systems.

Radiation shielding design of neutron source from Kartini reactor’s beam-port for SAMOP test facility

SAMOP is a subcritical assembly for 99Mo production fueled with uranyl nitrate solution which is designed to be operated by using external neutron source from a neutron generator. The most common radioisotope used in diagnosis for medical imaging in nuclear medicine is 99mTc which is daughter isotope of 99Mo. The SAMOP experimental or test facility is a facility to be used for performance testing of the SAMOP system by using neutron source from radial beam port of Kartini reactor instead of using neutron generator. This paper presents the results of the radiation shielding design and calculation for external neutron source of SAMOP experimental facility. The method used is a modeling and calculation by using MCNPX computer code. Based on the assumed source terms, the materials being used, and the geometrical arrangements, it is concluded that by using paraffin of 60 cm thickness for the beam catcher and 50 cm for the concrete of the outer shield would be sufficient to reduce the radiation dose below the maximum recommended limit. The presence of beam catcher can significantly reduce the contribution of neutrons and secondary particles to the radiation dose.

Calculation of the wide-angle neutron spectra from the 9Be(d, xn) reaction in a thick beryllium target * * Supported by the National Magnetic Confinement Fusion Science Program of China (2014GB104002), the National Natural Science Foundation of China (11705071, 11875155, 11675069, 21327801), NSAF (U1830102), the National Key Scientific Instrument and Equipment Development Project (2013YQ40861), and the Fundamental Research Funds for the

The multi-layer computing model is developed to calculate wide-angle neutron spectra, in the range from 0° to 180° with a 5° step, produced by bombarding a thick beryllium target with deuterons. The double-differential cross-sections (DDCSs) for the 9Be(d, xn) reaction are calculated using the TALYS-1.8 code. They are in agreement with the experimental data, and are much better than the PHITS-JQMD/GEM results at 15° , 30° , 45° and 60° neutron emission angles for deuteron energy of 10.0 MeV. In the TALYS-1.8 code, neutron contributions from direct reactions (break-up, stripping and knock-out reactions) are controlled by adjustable parameters, which describe the basic characteristics of typical direct reactions and control the relative intensity and the position of the ridgy hillock at the tail of DDCSs. It is found that the typical calculated wide-angle neutron spectra for different neutron emission angles and neutron angular distributions agree quite well with the experimental data for 13.5 MeV deuterons. The multi-layer computing model can reproduce the experimental data reasonably well by optimizing the adjustable parameters in the TALYS-1.8 code. Given the good agreement with the experimental data, the multi-layer computing model could provide better predictions of wide-angle neutron energy spectra, neutron angular distributions and neutron yields for the 9Be(d, xn) reaction neutron source.

Pre-saddle and pre-scission neutron emission rates in heavy-ion induced fission of 16O +208Pb and 16O +209Bi systems

Whereas there is a slight information on the pre-saddle neutron emission rate and neutron multiplicity, as well as it is impossible to separate the pre-saddle and saddle to scission neutron contributions experimentally, the theoretical studies of pre-saddle neutron emission rate and neutron multiplicity are of great importance. In the present work, the calculations of pre-saddle neutron multiplicity are performed using the analysis of fission fragment angular anisotropy data for [Formula: see text] and [Formula: see text] reaction systems. The obtained results show that the pre-saddle neutron multiplicity decreases by increasing the initial excitation energy and it has found to be characterized by a nonlinear behavior. Through the analysis of pre-saddle neutron multiplicity and pre-saddle transition time by means of the neutron clock method, the pre-saddle neutron emission rate is calculated for the first time. The findings of this study show that the pre-scission neutron emission rate is lower than the pre-saddle neutron emission rate.

Efficiency of Individual Targeting of a Three-Dimensional ITER Neutron Source in Monte Carlo Reactor Calculations

The extension of the known source angular biasing technique by individual targeting for each of the source neutrons on a given arbitrary rectangular target on the reactor first wall, indicating the area of interest in the problem, is considered in search of the possibilities to reduce the time of Monte Carlo calculation for bulky ITER models. The technique is tested in the source unit of the 3D code BLANK implementing the direct Monte Carlo method. A simple cylindrical model of ITER with a rectangular generatrix and a thickness of the side wall of the shielding material (80% stainless steel + 20% H2O) of 50 cm is calculated. The rectangular target matches the input hole of the equatorial port (2 × 2 m). The model calculations showed that the application of the technique did not result in a noticeable reduction of statistical uncertainty for the total neutron flux and its energy components in the first wall zone in the entire range of the targeting probability owing to the prevailing contributions of albedo neutrons. The error at the back surface of the model side wall was ~0.75 of the reference uncertainty (without targeting) in a targeting probability range of 0.2–0.4, which corresponds to a potential gain in computation time of up to ~2 times. These results yielded general empirical conclusions: the neutron flux estimated using this technique is unbiased, and a certain gain in the computer time, as compared to the calculation without targeting, is possible. The technique is expected to be more efficient in calculation of the neutron flux functionals in the depth of the shielding compositions, where the contribution of albedo neutrons from the first wall is insignificant, and the result is mainly determined by the intensity of the leading group of source neutrons. Since the considered technique is universal, it can be implemented as a useful option in any Monte Carlo code.

Electric dipole moment of the neutron from a flavor changing Higgs-boson

I consider neutron electric dipole moment contributions induced by flavor changing Standard Model Higgs boson couplings to quarks. Such couplings might stem from non-renormalizable SU(2)_L times U(1)_Y invariant Lagrange terms of dimension six, containing a product of three Higgs doublets. We extend previous one loop analysis to two loops. The divergent loops, due to non-renormalisabillity, are parametrized in terms of an ultraviolet cut-off Lambda . I also consider QCD corrections. Using the current experimental bound on the neutron electric dipole moment, then for cut offs from one to seven TeV, I find a constraint of order 10^{-3} for the imaginary part of the product of the Higgs flavor changing coupling for (d rightarrow b)-transition and the CKM element V_{td}. Assuming that the previous bound of the absolute value of the Higgs flavor changing coupling for (d rightarrow b)-transition obtained from B_d - bar{B_d}-mixing is saturated, the experimental bound on the neutron electric dipole moment would be reached for the bare result, if the cut off were extended up to about ca 20 TeV. However, QCD corrections suppress this result by a factor of order ten, and keep the nEDM below the experimental bound.

Physics study of liquid hydrogen moderator of CMRR reactor cold neutron source

The neutronic performance of a liquid hydrogen (LH) moderator is affected by its constituent ratio of para-hydrogen (pH) to ortho-hydrogen (oH) due to the difference of their scattering cross sections, as well as by the moderator geometric structure. Initially, the transmission and reflection properties of pure pH and oH slabs were analyzed. These indicated that the reflection method would generate a higher cold neutron flux (CNF) than the transmission method for both pH and oH, and that oH has a better reflection performance than pH. Secondly, the slab LH moderator located at the neutron channel entrance in the reactor was investigated under different conditions viz. pH percentage, slab thickness and neutron channel transparency. Results show that the slab thickness of the LH moderator in the range of 3–5 cm is appropriate for mixed pH and oH, that the optimum value of the pH proportion decreases with neutron channel transparency (optimum pH proportion for absorbent, translucent and transparency neutron channel walls are 100%, 75% and 50% respectively), and that the CNF with a transparent wall is 2–3 times higher than that with an absorbent wall. Bending the LH slab to various degrees showed that the CNF with an absorbent wall channel is determined primarily by the curvature of a concave moderator-viewed surface. However, the CNF for the transparent wall case is mainly determined by the viewed surface area due to a contribution of neutrons reflected from the heavy water. In the case of a 2 cm thick aluminum-alloy channel wall, we obtain an optimized geometry of a half-cylinder shape, balancing the area and curvature of the viewed surface. For 75% pH and 100% pH components, compared with the reference LH moderator configuration, the CNF of our optimized moderator increased by 41% and 20%, respectively.

The CANDELLE Experiment for Characterization of Neutron Sensitivity of LiF TLDs

As part of the design studies of future power and research nuclear reactors, the validation of neutron and photon calculation schemes requires the implementation of nuclear heating measurements. Such measurements are usually performed in very low-power reactors, whose core dimensions are accurately known and where irradiation conditions are entirely controlled. The use of thermoluminescent dosimeters is particularly well suited to gamma heating measurements, although the TLDs are also sensitive to neutrons depending on their composition. Hence, for gamma heating assessment in a mixed gamma–neutron field, the neutron sensitivity correction factors of TLDs has to be accurately determined to ensure a better interpretation of doses measured in a reactor with reduced uncertainties. This paper presents the CANDELLE experiment, which is dedicated to the implementation of a pure neutron field calibration of lithium-enriched TLDs, by using the Platform for Studies and Research on Nuclear Energy platform of LPSC. This calibration step is part of a promising approach that has been recently developed at INFN for medical purposes, aiming at determining the gamma and neutron contributions to the doses measured by LiF TLDs in the mixed field, from the peak height ratios of the gamma and neutron calibration glow curves of this type of TLDs. The test irradiations conducted at LPSC within the first phase of the CANDELLE project allowed retrieving the peak height of LiF TLDs in fast neutron field, hence providing an early assessment of the neutron components of doses measured in the EOLE reactor at CEA Cadarache with around 10% uncertainty at $1\sigma$ . From these preliminary results, the second phase of the project will be devoted, on the one hand, to the use of moderator screens aiming at achieving a thermal neutron field while avoiding a too strong attenuation of the flux, and on the other hand, to the determination of the neutron spectrum at different locations within the irradiation area.

Contribution of Thermal Neutrons to Soft Error Rate

Boron-10 (10B) is present in a large number of memory and logic devices below 65 nm through contamination during fabrication process, and the reaction of low-energy cosmic neutrons with high concentrations of 10B in the device contributes to the soft error rate (SER). Based on the experiments at Institut Laue–Langevin facility, the thermal neutron contribution of three devices (one static random access memory, one field-programmable gate array, and one microprocessor) is investigated for both ground level and aircraft environments. We conclude that the thermal neutron-induced single event upsets could have a similar contribution to SER as the high-energy part in atmospheric applications.

Effects of 10 MV and Flattening-Filter-Free Beams on Peripheral Dose in a Cohort of Pediatric Patients

To assess the effect of flattening-filter-free (FFF) and 10MV radiation therapy beams on the peripheral dose received by a population of pediatric patients undergoing volumetric modulated arc therapy (VMAT). Twenty-six previously delivered 6MV flattened VMAT pediatric radiation therapy treatments plans were replanned with 6MV flattened, 6MV FFF, and 10MV FFF VMAT. Monte Carlo simulation code EGSnrc was used in conjunction with a measurement-based model to obtain 3-dimensional dose distributions. Peripheral dose delivered by FFF beams was compared with that delivered by 6MV flattened beams. A statistical analysis was performed to determine whether certain clinical factors (eg, target volume, location) were associated with a change in integral relative radiation dose. Neutron dose measurements assessed the neutron contribution from the 6MV flattened and 10MV FFF x-ray beams. Both the 6MV FFF and 10MV FFF beams delivered significantly lowerperipheral radiation doses than 6MV flattened (P<.01). The dose reduction was of 3.9% (95% confidence interval [CI] 2.1-5.7) and 9.8% (95% CI, 8.0-11.6) at 5cm from the PTV and 21.9% (95% CI, 13.7-30.1) and 25.6% (95% CI, 17.6-33.6) at 30cm for 6MV FFF and 10MV FFF beams, respectively. The clinical factorsexamined did not have a significant effect on the relative magnitude of the peripheral dose reduction. The upper limit on the neutron dose was determinedto be 203μSv for the 6MV flattened and 522μSv for the 10MV FFF beam. Both FFF beams significantly (P<.01) reduced the peripheral dose. 10MV FFF was more effective at reducing peripheral dose at distances <5cm from the PTV edge. The neutron doses delivered by all beams were <1% compared with the photon doses. 10MV FFF should be used to minimize peripheral dose.

Determination of thorium fission rate based on 135I in thorium oxide cylinder bombarded by D-T fusion neutrons

Aiming at developing the activation method and provide reference for the validation of thorium fission parameters, a dedicated integral experiment was carried out in a thorium oxide cylinder bombard by D-T fusion neutrons. Thorium fission rate in the axial direction of the cylinder was determined by detecting the 1260.409 keV gamma emitted from fission product 135I, with experimental uncertainties of 6.2%. MCNP calculations employing ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0, CENDL-3.1 library data were in good agreement with measured thorium fission rates within experimental uncertainties. The agreement of measured reaction rates of monitors with the calculations on the whole demonstrated the reliability of this experiment in some degree, and could provide reference for the analyses of thorium fission rates. Neutron spectra and the contribution of neutrons in each sub-energy range to thorium fission rate were also analyzed. The activation method to determine thorium fission rate was developed and the data obtained in this work could provide reference for the validation of thorium fission parameters.

CHARACTERIZATION OF THE EPITHERMAL NEUTRON FIELD PRODUCED BY p+7Li REACTION IN A TANDEM ACCELERATOR USING A BONNER SPHERE SPECTROMETER.

The proton beam produced in the Nuclear Physics line of the tandem accelerator of the Centro Nacional de Aceleradores was used to generate a neutron field. In particular, 1.912 MeV protons were used to produce well-known epithermal neutrons through the p+7Li → n+7Be reaction. The aim of the work was to characterize this field while testing the performance of a Bonner sphere spectrometer in the epithermal range. Measurements were performed in four locations at different angle (0°, 30°, 60° and 90°) from beam incidence direction in order to study the angular dependence of the field. Both a parametric and numerical unfolding methods were tested to process the counts of the central detectors and obtain the energy distribution of the neutron fluence. In addition, a Monte Carlo simulation was carried out to complete the study and provide a guess spectrum for numerical unfolding. It was found that the fluence rate and mean energy decrease as the angle from beam direction increases. Total fluence was 2.75, 1.36, 0.366 and 0.216 cm-2 per charge collected in the target at 0°, 30°, 60° and 90°, respectively. Mean energy of the field ranges from 46 to 17 keV at 0° and 60°, respectively. In all cases, given that the irradiation room is so large, the contribution of thermal neutrons to the field is small. Regarding the unfolding, the total fluences estimated by all methods were in agreement within the uncertainties.

Three–Dimensional parton structure of light nuclei

Two promising directions beyond inclusive deep inelastic scattering experiments, aimed at unveiling the three dimensional structure of the bound nucleon, are reviewed, considering in particular the 3He nuclear target. The 3D structure in coordinate space can be accessed through deep exclusive processes, whose non-perturbative part is encoded in generalized parton distributions. In this way, the distribution of partons in the transverse plane can be obtained. As an example of a deep exclusive process, coherent deeply virtual Compton scattering off 3He nuclei, important to access the neutron generalized parton distributions (GPDs), will be discussed. In Impulse Approximation (IA), the sum of the two leading twist, quark helicity conserving GPDs of 3He, H and E, at low momentum transfer, turns out to be dominated by the neutron contribution. Besides, a technique, able to take into account the nuclear effects included in the Impulse Approximation analysis, has been developed. The spin dependent GPD of 3He is also found to be largely dominated, at low momentum transfer, by the neutron contribution. The knowledge of the GPDs H,E and of 3He is relevant for the planning of coherent DVCS off 3He measurements. Semi-inclusive deep inelastic scattering processes access the momentum space 3D structure parameterized through transverse momentum dependent parton distributions. A distorted spin-dependent spectral function has been recently introduced for 3He, in a non-relativistic framework, to take care of the final state interaction between the observed pion and the remnant in semi-inclusive deep inelastic electron scattering off transversely polarized 3He. The calculation of the Sivers and Collins single spin asymmetries for 3He, and a straightforward procedure to effectively take into account nuclear dynamics and final state interactions, will be reviewed. The Light-front dynamics generalization of the analysis is also addressed.

Fast neutron sensitivity of neutron detectors based on Boron-10 converter layers

In the last few years many detector technologies for thermal neutron detection have been developed in order to face the shortage of 3He, which is now much less available and more expensive. Moreover the 3He-based detectors can not fulfil the requirements in performance, e.g. the spatial resolution and the counting rate capability needed for the new instruments. The Boron-10-based gaseous detectors have been proposed as a suitable choice. This and other alternative technologies are being developed at ESS. Higher intensities mean higher signals but higher background as well. The signal-to-background ratio is an important feature to study, in particular the γ-ray and the fast neutron contributions. This paper investigates, for the first time, the fast neutrons sensitivity of 10B-based thermal neutron detector. It presents the study of the detector response as a function of energy threshold and the underlying physical mechanisms. The latter are explained with the help of theoretical considerations and simulations.

Fast neutron detection at near-core location of a research reactor with a SiC detector

The measurable charged-particle produced from the fast neutron interactions with the Si and C nucleuses can make a wide bandgap silicon carbide (SiC) sensor intrinsically sensitive to neutrons. The 4H-SiC Schottky detectors have been fabricated and tested at up to 500 °C, presenting only a slightly degraded energy resolution. The response spectrum of the SiC detectors were also obtained by exposing the detectors to external neutron beam irradiation and at a near-core location where gamma-ray field is intense. The fast neutron flux of these two locations are ∼4.8×104cm−2⋅s−1 and ∼2.2×107cm−2⋅s−1, respectively. At the external beam location, a Si detector was irradiated side-by-side with SiC detector to disjoin the neutron response from Si atoms. The contribution of gamma ray, neutron scattering, and charged-particles producing reactions in the SiC was discussed. The fast neutron detection efficiencies were determined to be 6.43×10−4 for the external fast neutron beam irradiation and 6.13×10−6 for the near-core fast neutron irradiation.