Mixed Neutron Research Articles

Hardening Approach to Use CMOS Image Sensors for Fusion by Inertial Confinement Diagnostics

A hardening method is proposed to enable the use of CMOS image sensors for Fusion by Inertial Confinement Diagnostics. The mitigation technique improves their radiation tolerance using a reset mode implemented in the device. The results obtained evidence a reduction of more than 70% in the number of transient white pixels induced in the pixel array by the mixed neutron and γ-ray pulsed radiation environment.

Intercomparison of Monte Carlo radiation transport codes to model TEPC response in low-energy neutron and gamma-ray fields

Tissue-equivalent proportional counters (TEPC) can potentially be used as a portable and personal dosemeter in mixed neutron and gamma-ray fields, but what hinders this use is their typically large physical size. To formulate compact TEPC designs, the use of a Monte Carlo transport code is necessary to predict the performance of compact designs in these fields. To perform this modelling, three candidate codes were assessed: MCNPX 2.7.E, FLUKA 2011.2 and PHITS 2.24. In each code, benchmark simulations were performed involving the irradiation of a 5-in. TEPC with monoenergetic neutron fields and a 4-in. wall-less TEPC with monoenergetic gamma-ray fields. The frequency and dose mean lineal energies and dose distributions calculated from each code were compared with experimentally determined data. For the neutron benchmark simulations, PHITS produces data closest to the experimental values and for the gamma-ray benchmark simulations, FLUKA yields data closest to the experimentally determined quantities.

Determination of Electrochemical Corrosion Potential along the JMTR In-Pile Loop - II: Validation of ECP Electrodes and the Extrapolation of Measured ECP Data

A water chemistry research program using the in-pile loop in the Japan Materials Testing Reactor (JMTR) has been launched to develop data that are applicable for model verification as well as model benchmarking. In advance of the in-pile loop experiment performed in the JMTR, reliability of in-pile electrochemical corrosion potential (ECP) measurement and applicability of the theoretical models were investigated, based on experimental data previously obtained in the in-pile loop of the LVR-15 experimental reactor at the Research Center Rez in the Czech Republic. The responses of different types of reference electrodes used for the ECP measurements were compared with each other to confirm their reliability under several different irradiation conditions corresponding to the core peripherals of boiling water reactors (BWRs). The corrosive conditions along the in-pile loop were first calculated using combined models of water radiolysis and ECP, and the calculated results were then compared with the ECP measurement data to validate the model.As a result, it was confirmed that the reference electrodes performed reliably under mixed neutron and gamma-ray irradiation conditions with minor calibration of each electrode prior to application in the loop, and that the combined models of water radiolysis and ECP can be applied for the evaluation of the corrosive conditions of the in-pile loop and BWR cores and their peripherals.

Measurements of Thermal Neutron Response in Cherenkov Glasses Designed for MeV Photon Detection

In order to obtain a meaningful gamma ray measurement in mixed neutron and high-energy photon environments, a quantitative understanding of neutron-induced backgrounds is crucial. Neutrons that can produce Cherenkov photons indirectly through (n, γ) activation reactions are found in radiation backgrounds in both passive and active-source search environments. The objectives of this research are to establish the properties and fabrication techniques for materials suitable for the production of photon-sensitive Cherenkov detectors (including low neutron sensitivity), or, for certain samples, to understand how the response to neutrons can be electronically discriminated from the signal caused by the interaction of MeV photons. Over a hundred glass samples have been fabricated, and radiation characterization results from selected samples are reported. The thermal neutron responses of these glasses were determined using neutron activation followed by time dependent radiation measurement and gamma ray spectroscopy. The results showed that the probability of thermal neutron capture, the thermal neutron activation time, the half-life of the activation products, the abundance of the target materials, and the magnitude of the energy of gamma and beta radiations that result from activation play significant roles in understanding the observed count rate due to Cherenkov radiation. Implications for Cherenkov detectors to be used for MeV photon detection are discussed.

Fast non-overlapping Schwarz domain decomposition methods for solving the neutron diffusion equation

Studying numerically the steady state of a nuclear core reactor is expensive, in terms of memory storage and computational time. In order to address both requirements, one can use a domain decomposition method, implemented on a parallel computer. We present here such a method for the mixed neutron diffusion equations, discretized with Raviart–Thomas–Nédélec finite elements. This method is based on the Schwarz iterative algorithm with Robin interface conditions to handle communications. We analyse this method from the continuous point of view to the discrete point of view, and we give some numerical results in a realistic highly heterogeneous 3D configuration. Computations are carried out with the MINOS solver of the APOLLO3®1APOLLO3 is a registered trademark in France.1 neutronics code.

The effect of neutron and mixed gamma and neutron irradiation on the solar properties of borosilicate glass

Neutron and mixed gamma and neutron irradiation, at different absorbed doses, of borosilicate glass with four different chemical compositions was conducted to investigate the effects on the solar properties of the glass. Irradiation was performed in the tangential beam tube and central thimble of a nuclear reactor. The effect of thermal and epithermal neutrons on such solar properties as secondary heat-transfer factor, solar factor (the total solar energy transmittance), and shading coefficient of the borosilicate glass were investigated to determine the effect on the solar properties of borosilicate glass, because of its neutron absorbance property.

Finding a suitable shield for mixed neutron and photon fields based on an Am–Be source

This study aimed to investigate a shielding design against neutron and photon rays from neutron irradiators based on Am–Be sources, using the Monte Carlo simulation. Different shielding materials were studied, including borated polyethylene, DaGa concrete, and epoxy resin with colemanite. The Monte Carlo N-particle code (MCNP) was used to design shielding. A new type of neutron and photon shielding material based on 40 % galena, 55 % polyethylene, and 5 % boric acid is proposed. The results show that the total effective dose of radiation is significantly reduced by the optimum radius of this shielding system.

Necrosis is not induced by gadolinium neutron capture in glioblastoma multiforme cells

Purpose: A comparative study of the effects of different radiation modalities on cell death was performed.Materials and methods: Radiation modalities included γ-rays, fast neutrons, a mixed energy neutron beam called the modified enhanced thermal neutron beam and the mixed beam including Auger electron irradiation by gadolinium neutron capture. U87 (human brain tumor cells) cell survival curve data were modeled to predict how cells died. Transmission electron microscopy (TEM) images were assembled into a morphology of cell death (MCD) database and used to determine the fraction of necrotic or autophagic cells.Results: Linear energy transfer (LET) differences for the different radiation modalities were revealed by modeling. All radiation modalities induced autophagy but only fast neutrons induced significant levels of necrosis. No necrosis, above control levels, was found in cells irradiated with mixed beam irradiation including Auger electrons. The number of autophagosomes increased with increasing time after exposure to all radiation modalities indicating progression of autophagy but only cells irradiated with the mixed beam plus Auger electrons exhibited extreme autophagy.Conclusions: Mixed neutron beam irradiation plus Auger electron irradiation from gadolinium neutron capture is a moderately high LET modality that kills U87 cells without the induction of necrosis and with progression of autophagy to an extreme state.

Characterization of extended range Bonner Sphere Spectrometers in the CERF high-energy broad neutron field at CERN

The accurate determination of the ambient dose equivalent in the mixed neutron–photon fields encountered around high-energy particle accelerators still represents a challenging task. The main complexity arises from the extreme variability of the neutron energy, which spans over 10 orders of magnitude or more. Operational survey instruments, which response function attempts to mimic the fluence-to-ambient dose equivalent conversion coefficient up to GeV neutrons, are available on the market, but their response is not fully reliable over the entire energy range. Extended range rem counters (ERRC) do not require the exact knowledge of the energy distribution of the neutron field and the calibration can be done with a source spectrum. If the actual neutron field has an energy distribution different from the calibration spectrum, the measurement is affected by an added uncertainty related to the partial overlap of the fluence-to-ambient dose equivalent conversion curve and the response function. For this reason their operational use should always be preceded by an “in-field” calibration, i.e. a calibration made against a reference instrument exposed in the same field where the survey-meter will be employed. In practice the extended-range Bonner Sphere Spectrometer (ERBSS) is the only device which can serve as reference instrument in these fields, because of its wide energy range and the possibility to assess the neutron fluence and the ambient dose equivalent (H⁎(10)) values with the appropriate accuracy. Nevertheless, the experience gained by a number of experimental groups suggests that mandatory conditions for obtaining accurate results in workplaces are: (1) the use of a well-established response matrix, thus implying validation campaigns in reference monochromatic neutrons fields, (2) the expert and critical use of suitable unfolding codes, and (3) the performance test of the whole system (experimental set-up, elaboration and unfolding procedures) in a well controlled workplace field. The CERF (CERN-EU high-energy reference field) facility is a unique example of such a field, where a number of experimental campaigns and Monte Carlo simulations have been performed over the past years.With the aim of performing this kind of workplace performance test, four different ERBSS with different degrees of validation, operated by three groups (CERN, INFN-LNF and Politecnico of Milano), were exposed in two fixed positions at CERF. Using different unfolding codes (MAXED, GRAVEL, FRUIT and FRUIT SGM), the experimental data were analyzed to provide the neutron spectra and the related dosimetric quantities. The results allow assessing the overall performance of each ERBSS and of the unfolding codes, as well as comparing the performance of three ERRCs when used in a neutron field with energy distribution different from the calibration spectrum.

Method to Calibrate Fission Chambers in Campbelling Mode

Fission chambers are neutron detectors which are widely used to instrument experimental reactors such as material testing reactors or zero power reactors. In the presence of a high level mixed gamma and neutron flux, fission chambers can be operated in Campbelling mode (also known as “fluctuation mode” or “mean square voltage mode”) to provide reliable and precise neutron related measurements. Fission chamber calibration in Campbelling mode (in terms of neutron flux) is usually done empirically using a calibrated reference detector. A major drawback of this method is that calibration measurements have to be performed in a neutron environment very similar to the one in which the calibrated detector will be used afterwards. What is proposed here is a different approach based on characterizing the fission chamber response in terms of fission rate. This way, the detector calibration coefficient is independent from the neutron spectrum and can be determined prior to the experiment. The fissile deposit response to the neutron spectrum can then be assessed independently by other means (experimental or numerical). In this paper, the response of CEA-made miniature fission chambers in Campbelling mode is studied. A theoretical model of the signal is used to calculate the calibration coefficient. The model's input parameters come from statistical distribution of individual pulses. Supporting measurements were made in the CEA Cadarache zero power reactor MINERVE and results are compared to an empirical Campbelling mode calibration. The tested fission chamber calibration coefficient is roughfly 2 <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$10^{-26}$</tex></formula> <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">${\rm A}^{2}$</tex></formula> /Hz/(c/s). Both numerical and empirical methods give consistent results, however a deviation of about 15% was observed.

Study of resistive micromegas detectors in a mixed neutron and photon radiation environment

The Muon ATLAS Micromegas Activity (MAMMA) focuses on the development and testing of large-area muon detectors based on the bulk-Micromegas technology. These detectors are candidates for the upgrade of the ATLAS Muon System in view of the luminosity upgrade of Large Hadron Collider at CERN (sLHC). They will combine trigger and precision measurement capability in a single device. A novel protection scheme using resistive strips above the readout electrode has been developed. The response and sparking properties of resistive Micromegas detectors were successfully tested in a mixed (neutron and gamma) high radiation environment supplied by the Tandem accelerator at the N.C.S.R. Demokritos in Athens. Monte-Carlo studies have been employed to study the effect of 5.5 MeV neutrons impinging on Micromegas detectors. The response of the Micromegas detectors on the photons originating from the inevitable neutron inelastic scattering on the surrounding materials of the experimental facility was also studied.

Experimental study and FLUKA simulations of a prototype micromegas chamber in a mixed neutron and photon radiation field

Detectors based on the micromegas principle have already been used in several atomic, nuclear and particle physics experiments. They have also been proposed as one of the options to upgrade the ATLAS muon spectrometer in the very forward/backward region. To meet this end, it is imperative to study their performance in a mixed (neutron and gamma) radiation field. The general-purpose Monte-Carlo code FLUKA has been employed in the present work in order to study the effect of 5.5MeV neutrons impinging on a prototype micromegas detector developed for sLHC. The response of the detector to the photons originating from the inevitable neutron inelastic scattering on the surrounding materials of the experimental facility was also studied, through comparisons with experimental data.

Dose evaluation in criticality accidents using response of panasonic TL personal dosemeters (UD-809/UD-802)

This study gives the results of dosimetry measurements carried out in the Silène reactor at Valduc (France) with neutron and photon personal thermoluminescence dosemeters (TLDs) in mixed neutron and gamma radiation fields, in the frame of the international accident dosimetry intercomparison programme in 2002. The intercomparison consisted of a series of three irradiation scenarios. The scenarios took place at the Valduc site (France) by using the Silène experimental reactor. For neutron and photon dosimetry, Panasonic model UD-809 and UD-802 personal TLDs were used together.

Rhodium self‐powered neutron detector as a suitable on‐line thermal neutron flux monitor in BNCT treatments

A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed-field thermal neutron sensitivities derived from measurements performed at the RA-6 were compared and no significant differences were found. Global RA-6-based thermal neutron sensitivity showed agreement with pure thermal neutron sensitivity measurements performed in the RA-3 spectrum. Additionally, the detector response proved nearly unchanged by differences in neutron spectra from real (RA-6 BNCT beam) and ideal (considered for calibration calculations at RA-3) neutron source descriptions. The results confirm that the special design of the Rh SPND can be considered as having a pure thermal response for neutron spectra with epithermal-to-thermal flux ratios up to 12%. In addition, the linear response of the detector to thermal flux allows the use of a mixed-field thermal neutron sensitivity of 1.95 ± 0.05 × 10(-21) A n(-1)[middle dot]cm² [middle dot]s. This sensitivity can be used in spectra with up to 21% epithermal-to-thermal flux ratio without significant error due to epithermal neutron and gamma induced effects. The values of the measured fluxes in clinical applications had discrepancies with calculated results that were in the range of -25% to +30%, which shows the importance of a local on-line independent measurement as part of a treatment planning quality control system. The usefulness of the CNEA Rh SPND for the on-line local measurement of thermal neutron flux on BNCT patients has been demonstrated based on an appropriate neutron spectra calibration and clinical applications.

Calculation and measurement of neutron flux in the VVER-1000 mock-up on the LR-0 research reactor

The neutron fluence load on a reactor pressure vessel is an important physical quantity affecting material degradation and consequently reliable assessment of pressure vessel integrity and lifetime prolongation beyond designed conditions. This degradation is influenced by mixed neutron–photon source from the core periphery as well as by material parameters of the reactor pressure vessel. Computational procedures and experimental determination of neutron fluxes in the VVER-1000 mock-up internal structures as well as the results achieved are described in this paper. The calculation were performed with the MCNPX code with different nuclear data libraries. Nuclear data were processed using NJOY code. The neutron spectra measurements were performed with a two-parameter stilbene spectrometer.

Characterization of radiation fields and dose assessment from fuels manufacturing for advanced fuel cycles

The purpose of this study was to examine radiological hazards introduced to workers from the fabrication of fuels with minor actinides (MA) and reprocessed uranium (RepU) and determine the feasibility of using shielded gloveboxes instead of remotely controlled operations in hot cells. Of particular concern is the increase in photon source term from the daughter products in RepU and the mixed neutron and photon fields introduced by MA recycle. In the interest of keeping the glovebox worker's radiation dose as low as reasonable achievable (ALARA), dose rates were calculated for various typical and bounding fuel compositions for light water reactors (LWR) and fast reactor (FR) mixed oxide (MOX) fuels with and without MA. The impact of varying the separation efficiency of americium (Am) and curium (Cm) was examined because current separation processes in reprocessing and recycling do not allow for the complete separation of Am from Cm. The additional Cm will cause a significant increase in the neutron source term. The sensitivity of the fuels to aging time was also examined by decaying the recycled feedstocks from 6 months to 3 years to simulate the effect of delays in reuse of recycled materials. The highest photon and neutron sources were used to calculate the additional shielding requirements that would be needed for fuel fabrication. Through insight gained from this study, it can be concluded that a standard glovebox with one quarter inch stainless steel walls can be used to fabricate fuels with Am with little to no additional shielding. The introduction of small amounts of Cm in the fuels will require the fuel fabrication to be preformed remotely in hot cells. Thus, stressing the importance of developing methods to increase the separation efficiency of Am from Cm.

Dose determination using alanine detectors in a mixed neutron and gamma field for boron neutron capture therapy of liver malignancies

Boron Neutron Capture Therapy for liver malignancies is being investigated at the University of Mainz. One important aim is the set-up of a reliable dosimetry system. Alanine dosimeters have previously been applied for dosimetry of mixed radiation fields in antiproton therapy, and may be suitable for measurements in mixed neutron and gamma fields. Material and methods. Two experiments have been carried out in the thermal column of the TRIGA Mark II reactor at the University of Mainz. Alanine dosimeters have been irradiated in a phantom and in liver tissue. Results. For the interpretation and prediction of the dose for each pellet, beside the results of the measurements, calculations with the Monte Carlo code FLUKA are presented here. For the phantom, as well as for the liver tissue, the measured and calculated dose and flux values are in good agreement. Discussion. Alanine dosimeters, in combination with flux measurements and Monte Carlo calculations with FLUKA, suggest that it is possible to establish a system for monitoring the dose in a mixed neutron and gamma field for BNCT and other applications in radiotherapy.

ENTROPY ENTRAINMENT AND DISSIPATION IN FINITE TEMPERATURE SUPERFLUIDS

Neutron stars are expected to contain several distinct superfluid components, ranging from the neutron superfluid which coexists with the elastic crust to the mixed neutron superfluid/proton superconductor in the outer core and more exotic phases like superfluid hyperons and colour-flavour-locked superconducting quarks in the deep core. These different phases may have significant effect on the dynamics of the system. Building on a general variational framework for multifluid dynamics, we consider the behaviour of superfluid systems at finite temperatures (as required to understand various dissipation channels). As a demonstration of the validity of the underlying principles, such as treating the excitations in the system as a massless "entropy" fluid, we show that the model is formally equivalent to the traditional two-fluid approach for superfluid helium. In particular, we demonstrate how the entropy entrainment is related to the "normal fluid density". We also show how the superfluid constraint of irrotationality reduces the number of dissipation coefficients in the system. The analysis provides insight into the more general problem where vortices are present in the superfluid, and we discuss how the so-called mutual friction force can be accounted for. The end product is a formalism for finite temperature effects in a single condensate that can be applied to both low temperature laboratory systems and the various superfluid phases in a neutron star. This provides a key step towards the modelling of more realistic neutron star dynamics, and the understanding of a range of phenomena from pulsar glitches to magnetar seismology and the gravitational-wave-driven r-mode instability.

Development of a thick gas electron multiplier for microdosimetry

A new tissue-equivalent proportional counter based on a thick gas electron multiplier (THGEM) was developed and tested for microdosimetry. A systematic test was conducted at the McMaster Accelerator Laboratory to investigate the overall performance of the prototype detector. A mixed neutron–gamma-ray radiation field was generated using the 7Li(p,n) reaction. The detector was operated at low voltage initially to test the stability and then the relative multiplication gain was measured as a function of the operating high voltage. A drift potential of 100 V and a THGEM bias of 727 V generated a multiplication gain sufficient for the detection of both neutron and gamma-ray radiation. A consistent microdosimetric pattern was observed between the THGEM detector and standard TEPC for microdosimetry.

Proof of principle of a high-spatial-resolution, resonant-response γ-ray detector for Gamma Resonance Absorption in 14N

The development of a mm-spatial-resolution, resonant-response detector based on a micrometric glass capillary array filled with liquid scintillator is described. This detector was developed for Gamma Resonance Absorption (GRA) in 14N. GRA is an automatic-decision radiographic screening technique that combines high radiation penetration (the probe is a 9.17 MeV γ-ray) with very good sensitivity and specificity to nitrogenous explosives. Detailed simulation of the detector response to electrons and protons generated by the 9.17 MeV γ-rays was followed by a proof-of-principle experiment, using a mixed γ-ray and neutron source. Towards this, a prototype capillary detector was assembled, including the associated filling and readout systems. Simulations and experimental results indeed show that proton tracks are distinguishable from electron tracks at relevant energies, based on a criterion that combines track length and light intensity per unit length.