High-density Polyethylene Moderator Research Articles

A compact direction-sensitive fast neutron spectrometer

A compact neutron spectrometer is described, which consists of four square prisms of plastic scintillator capable of pulse shape discrimination, eight silicon photomultipliers and a high density polyethylene moderator. The device is capable, in principle, of spectroscopic measurements up to around 60 MeV, although extendable to higher energies, and is also able to discern the direction of neutron radiation impinging on the detector. Results from simulations using Geant4, together with measurements with neutrons from an 241Am-9Be radioisotopic source, are used to characterise the detector in both spectroscopic and direction-sensitive modes. Field applications of the detector are discussed.

Feasibility of irrigation monitoring with cosmic-ray neutron sensors

Abstract. Accurate soil moisture (SM) monitoring is key in irrigation as it can greatly improve water use efficiency. Recently, cosmic-ray neutron sensors (CRNSs) have been recognized as a promising tool in SM monitoring due to their large footprint of several hectares. CRNSs also have great potential for irrigation applications, but few studies have investigated whether irrigation monitoring with CRNSs is feasible, especially for irrigated fields with a size smaller than the CRNS footprint. Therefore, the aim of this study is to use Monte Carlo simulations to investigate the feasibility of monitoring irrigation with CRNSs. This was achieved by simulating irrigation scenarios with different field dimensions (from 0.5 to 8 ha) and SM variations between 0.05 and 0.50 cm3 cm−3. Moreover, the energy-dependent response functions of eight moderators with different high-density polyethylene (HDPE) thickness or additional gadolinium thermal shielding were investigated. It was found that a considerable part of the neutrons that contribute to the CRNS footprint can originate outside an irrigated field, which is a challenge for irrigation monitoring with CRNSs. The use of thin HDPE moderators (e.g. 5 mm) generally resulted in a smaller footprint and thus stronger contributions from the irrigated area. However, a thicker 25 mm HDPE moderator with gadolinium shielding improved SM monitoring in irrigated fields due to a higher sensitivity of neutron counts with changing SM. This moderator and shielding set-up provided the highest chance of detecting irrigation events, especially when the initial SM was relatively low. However, variations in SM outside a 0.5 or 1 ha irrigated field (e.g. due to irrigation of neighbouring fields) can affect the count rate more than SM variations due to irrigation. This suggests the importance of retrieving SM data from the surrounding of a target field to obtain more meaningful information for supporting irrigation management, especially for small irrigated fields.

Designand commissioning of a neutron counter adapted to high-intensity laser matter interactions.

The advent of multi-PW laser facilities world-wide opens new opportunities for nuclear physics. With this perspective, we developed a neutron counter taking into account the specifics of a high-intensity laser environment. Using GEANT4 simulations and prototype testings, we report on the designof a modular neutron counter based on boron-10 enriched scintillators and a high-density polyethylene moderator. This detector has been calibrated using a plutonium-beryllium neutron source and commissioned during an actual neutron-producing laser experiment at the LULI2000 facility (France). An overall efficiency of 4.37(59)% has been demonstrated during calibration with a recovery time of a few hundred microseconds after laser-plasma interaction.

Using Additional Moderator to Control the Footprint of a COSMOS Rover for Soil Moisture Measurement

AbstractCosmic‐Ray Neutron Probes (CRNP) have found application in soil moisture (SM) estimation due to their conveniently large (>100 m) footprints. Here, we explore the possibility of using high‐density polyethylene (HDPE) moderator to limit the field of view, and hence, the footprint of a SM sensor formed of 12 CRNP mounted on to a mobile robotic platform (Thorvald) for better in‐field localization of moisture variation. Ultra Rapid Adaptable Neutron‐Only Simulation neutron scattering simulations are used to show that 5 cm of additional HDPE moderator (used to shield the upper surface and sides of the detector) is sufficient to (a) reduce the footprint of the detector considerably, (b) approximately double the percentage of neutrons detected from within 5 m of the detector, and (c) does not affect the shape of the curve used to convert neutron counts into SM. Simulation and rover measurements for a transect crossing between grass and concrete additionally suggest that (d) SM changes can be sensed over a length scales of tens of meters or less (roughly an order of magnitude smaller than commonly used footprint distances), and (e) the additional moderator does not reduce the detected neutron count rate (and hence increase noise) as much as might be expected given the extent of the additional moderator. The detector with additional HDPE moderator was also used to conduct measurements on a stubble field over three weeks to test the rover system in measuring spatial and temporal SM variation.

Experimental characterization of FANT, a new thermal neutron source

FANT is the acronym of Enhanced Thermal Neutron Source (Fuente Ampliada de Neutrones Térmicos, in Spanish). This is a parallelepiped box of high-density polyethylene moderator and an isotopic neutron source. The moderator has a cylindrical irradiation chamber where a rather uniform thermal neutron flux is obtained. The FANT design was previously optimized and the neutron spectra were estimated by Monte Carlo calculations with the MCNP6.1 code. To check the characteristics of the FANT thermal neutron field, measurements have been performed at the reference point inside the irradiation chamber with a Bonner sphere spectrometer holding a small 6LiI(Eu) thermal neutron detector. To unfold the neutron spectrum BUNKIUT with UTA4 response matrix and NSDann Ver 4.0 codes were used. Some issues have been found and recommendations are made about the use of large BSS inside narrow spaces, and about the capacity of NSDann code to unfold these kind of spectra. However, the results confirm that the moderation process in FANT is very effective and allows obtaining useful thermal neutron fluence rates.

Design and simulation of a three-detector instrument for accurate neutron H*(10) and energy distribution calculation

Many efforts have been made to improve the energy response of neutron rem meters to match the fluence to ambient dose equivalent conversion coefficients(DCC) accurately. In this work, a design of a portable instrument based on three thermal neutron detectors is proposed and simulated. The detectors are embedded in different depth along the cylindrical high density polyethylene moderator and their count rates are recorded simultaneously. Two different H*(10) calculation methods have been introduced. One is linear correction method based on the combination of detector responses to approximate the DCC curve provided in ICRP74. The other is spectrum unfolding method which unfold the neutron spectrum at the measuring point for further H*(10) calculation. Monte Carlo simulations have been applied to calculate the response matrices of the three detectors. And standard neutron spectra are used to evaluate the performance of the instrument. As the results indicate, linear correction method is accurate and stable for H*(10) estimation, with mean relative prediction error of 8.45% for tested spectra. Spectrum unfolding method is not stable and have large H*(10) estimation error in some cases, but it can simultaneously derive neutron energy distributions with acceptable approximate degree. The instrument is highly suitable for H*(10) measurement for one-directional parallel neutron beam as well as isotropic neutron source with determined location. And by applying an absorber shell or detector arrays, the angular influence for the instrument is effectively improved to obtain more accurate H*(10) estimation results.

Design by Monte Carlo method of a thermal neutron device using a 241Am/9Be source and high-density polyethylene moderator

A thermal neutron system intended to be used in neutron activation analysis has been designed by Monte Carlo methods. The device is based on a241Am/9Be neutron source of 111 GBq, placed inside a cylindrical cavity open inside a parallelepiped of moderator material. Three different moderator materials, water, graphite and high-density polyethylene (HDPE), were simulated to check what is the most suitable for the detection system, concluding that HDPE reach the better performance. The device achieves an increased thermal neutron flux by taking advantage of neutron moderation in the polyethylene and the neutron scattering in the irradiation chamber walls. The thermal fluence rates obtained were 904 cm−2 s−1, i.e. 8.144 cm−2 s−1 GBq−1, with a fraction of thermal neutrons at the best point of 83% of pristine fast neutrons emitted by the source. The device has been designed by Monte Carlo techniques using the MCNP6 code, and the main tasks developed were to select the moderator material and to maximize the thermal neutrons flux in the irradiation chamber.

Wearable detector device utilizing microstructured semiconductor neutron detector technology

A Wearable Detector Device (WDD) has been outfitted with Microstructured Semiconductor Neutron Detectors (MSNDs) to aid in the search and localization of special nuclear materials (SNMs). Many SNMs decay by spontaneous fission and emit free neutrons. The WDD detects these neutrons and stores interaction rate information to alert the operator to the presence of special nuclear material. The WDD is composed of 16 Modular Neutron Detectors (MNDs), each populated with a 4 × 6 array of 1-cm2 active area, 500-μm thick MSNDs. The individual MSNDs each have an intrinsic thermal-neutron detection efficiency of approximately 30%. Each MND connects to a communications dongle, and the MND and dongle were encased in a 3-in. wide by 5-in. tall by 0.6-in. thick high-density polyethylene moderator (HDPE) case. The 16 MNDs, connected to Controller Area Network dongles, communicate with a master control board, which also contains the battery bank and power conditioning electronics. The operational lifetime of the battery-powered WDD is greater than 12 h per single charge. The WDD, mounted on an ANSI 42.53 phantom, reported 8.12 ± 0.07 cps and 12.93 ± 0.07 cps for a bare and moderated 21.9-ng 252Cf source at a distance of 1 m, respectively. The background count rate was 0.446 ± 0.002 cps. The gamma-ray rejection ratio of the WDD for 137Cs measured at a dose rate of 10 mR/h was 1.8 × 10−8.

Toward achieving flexible and high sensitivity hexagonal boron nitride neutron detectors

Hexagonal boron nitride (h-BN) detectors have demonstrated the highest thermal neutron detection efficiency to date among solid-state neutron detectors at about 51%. We report here the realization of h-BN neutron detectors possessing one order of magnitude enhancement in the detection area but maintaining an equal level of detection efficiency of previous achievement. These 3 mm × 3 mm detectors were fabricated from 50 μm thick freestanding and flexible 10B enriched h-BN (h-10BN) films, grown by metal organic chemical vapor deposition followed by mechanical separation from sapphire substrates. Mobility-lifetime results suggested that holes are the majority carriers in unintentionally doped h-BN. The detectors were tested under thermal neutron irradiation from californium-252 (252Cf) moderated by a high density polyethylene moderator. A thermal neutron detection efficiency of ∼53% was achieved at a bias voltage of 200 V. Conforming to traditional solid-state detectors, the realization of h-BN epilayers with enhanced electrical transport properties is the key to enable scaling up the device sizes. More specifically, the present results revealed that achieving an electrical resistivity of greater than 1014 Ω⋅cm and a leakage current density of below 3 × 10−10 A/cm2 is needed to fabricate large area h-BN detectors and provided guidance for achieving high sensitivity solid state neutron detectors based on h-BN.

Present Status of the Microstructured Semiconductor Neutron Detector-Based Direct Helium-3 Replacement

A small form-factor, third-generation microstructured semiconductor neutron detector (MSND)-based 3He replacement (HeRep Mk III) detector has been developed to serve as a direct replacement to expensive small-diameter, high-pressure 3He gas-filled proportional neutron counters. Previously, the larger, 2-in diameter by 6-in-long HeRep Mk II utilized thirty 4-cm2 active area single-sided MSNDs to replace a similarly sized 4-atm 3He counter. The net count rate of the HeRep Mk II was 95.15% ± 9.04% of the net count rate of a similarly sized 4-atm 3He detector for bare 252Cf at a distance of 1 m. The HeRep Mk II net count rate measured 102.71% ±2.65% of the 3He detector net count rate when each were placed in 6-in diameter by 9-in-long cylindrical high-density polyethylene (HDPE) moderator casks. The HeRep Mk III was developed as a test bed for new-generation dual-sided MSNDs (DSMSNDs), as well as for reduced-power electronics. The 0.75-in diameter by 4.50-in-long HeRep Mk III was populated with twelve 1-cm2 active area DSMSNDs. The pvp-type diodes have thus far realized 53.54% ± 0.61% intrinsic thermal-neutron detection efficiency. The HeRep Mk III was tested against a 0.75-in diameter by 3.0-in-long 10-atm 3He detector with a 30-ng 252Cf source at a distance of 0.25 m. The HeRep Mk III reported 45.4% ± 0.3% and 56.7% ± 0.2% of the count rate of the 10-atm 3He detector for detectors with no exterior moderator and in a 3-in by 5.75-in cylindrical HDPE moderator cask, respectively. The HeRep Mk III gamma-ray sensitivity was measured at a dose rate of 50 mR/hr from 137Cs resulting in a net count rate of 0.01 cps, which corresponds to a gamma-ray rejection ratio of approximately 5:107.

Conceptual design of a hybrid neutron-gamma detector for study of β-delayed neutrons at the RIB facility of RIKEN

The conceptual design of the BRIKEN neutron detector at the radioactive ion beam factory (RIBF) of the RIKEN Nishina Center is reported. The BRIKEN setup is a complex system aimed at detecting heavy-ion implants, β particles, γ rays and β-delayed neutrons. The whole setup includes the Advanced Implantation Detection Array (AIDA), two HPGe Clover detectors and up to 166 3He-filled counters embedded in a high-density polyethylene moderator. The design is quite complex due to the large number and different types of 3He-tubes involved and the additional constraints introduced by the ancillary detectors for charged particles and γ rays. This article reports on a novel methodology developed for the conceptual design and optimisation of the 3He-counter array, aiming for the best possible performance in terms of neutron detection. The algorithm is based on a geometric representation of two selected detector parameters of merit, namely, the average neutron detection efficiency and the efficiency flatness as a function of a reduced number of geometric variables. The response of the neutron detector is obtained from a systematic Monte Carlo simulation implemented in GEANT4. The robustness of the algorithm allowed us to design a versatile detection system, which operated in hybrid mode includes the full neutron counter and two clover detectors for high-precision gamma spectroscopy. In addition, the system can be reconfigured into a compact mode by removing the clover detectors and re-arranging the 3He tubes in order to maximize the neutron detection performance. Both operation modes shows a rather flat and high average efficiency. In summary, we have designed a system which shows an average efficiency for hybrid mode (3He tubes + clovers) of 68.6% and 64% for neutron energies up to 1 and 5 MeV, respectively. For compact mode (only 3He tubes), the average efficiency is 75.7% and 71% for neutron energies up to 1 and 5 MeV, respectively. The performance of the BRIKEN detection system has been also quantified by means of Monte Carlo simulations with different neutron energy distributions.

Using the MCNP Taylor series perturbation feature (efficiently) for shielding problems

The Taylor series or differential operator perturbation method, implemented in MCNP and invoked using the PERT card, can be used for efficient parameter studies in shielding problems. This paper shows how only two PERT cards are needed to generate an entire parameter study, including statistical uncertainty estimates (an additional three PERT cards can be used to give exact statistical uncertainties). One realistic example problem involves a detailed helium-3 neutron detector model and its efficiency as a function of the density of its high-density polyethylene moderator. The MCNP differential operator perturbation capability is extremely accurate for this problem. A second problem involves the density of the polyethylene reflector of the BeRP ball and is an example of first-order sensitivity analysis using the PERT capability. A third problem is an analytic verification of the PERT capability.

Realization of highly efficient hexagonal boron nitride neutron detectors

We report the achievement of highly efficient 10B enriched hexagonal boron nitride (h-10BN) direct conversion neutron detectors. These detectors were realized from freestanding 4-in. diameter h-10BN wafers 43 μm in thickness obtained from epitaxy growth and subsequent mechanical separation from sapphire substrates. Both sides of the film were subjected to ohmic contact deposition to form a simple vertical “photoconductor-type” detector. Transport measurements revealed excellent vertical transport properties including high electrical resistivity (>1013 Ω cm) and mobility-lifetime (μτ) products. A much larger μτ product for holes compared to that of electrons along the c-axis of h-BN was observed, implying that holes (electrons) behave like majority (minority) carriers in undoped h-BN. Exposure to thermal neutrons from a californium-252 (252Cf) source moderated by a high density polyethylene moderator reveals that 43 μm h-10BN detectors possess 51.4% detection efficiency at a bias voltage of 400 V, which is the highest reported efficiency for any semiconductor-based neutron detector. The results point to the possibility of obtaining highly efficient, compact solid-state neutron detectors with high gamma rejection and low manufacturing and maintenance costs.

Boron neutron capture therapy design calculation of a 3H(p,n) reaction based BSA for brain cancer setup

Purpose: Boron neutron capture therapy (BNCT) is a promising technique for the treatment of malignant disease targeting organs of the human body. Monte Carlo simulations were carried out to calculate optimum design parameters of an accelerator based beam shaping assembly (BSA) for BNCT of brain cancer setup. Methods: Epithermal beam of neutrons were obtained through moderation of fast neutrons from 3 H(p,n) reaction in a high density polyethylene moderator and a graphite reflector. The dimensions of the moderator and the reflector were optimized through optimization of epithermal / fast neutron intensity ratio as a function of geometric parameters of the setup. Results: The results of our calculation showed the capability of our setup to treat the tumor within 4 cm of the head surface. The calculated peak therapeutic ratio for the setup was found to be 2.15. Conclusion: With further improvement in the polyethylene moderator design and brain phantom irradiation arrangement, the setup capabilities can be improved to reach further deep-seated tumor.

Neutron Multiplicity Measurements with 3He Alternative: Straw Neutron Detectors

Counting neutrons emitted by special nuclear material (SNM) and differentiating them from the background neutrons of various origins is the most effective passive means of detecting SNM. Unfortunately, neutron detection, counting, and partitioning in a maritime environment are complex due to the presence of high-multiplicity spallation neutrons (commonly known as “ship effect”) and to the complicated nature of the neutron scattering in that environment. A prototype neutron detector was built using 10B as the converter in a special form factor called “straws” that would address the above problems by looking into the details of multiplicity distributions of neutrons originating from a fissioning source. This paper describes the straw neutron multiplicity counter (NMC) and assesses the performance with those of a commercially available fission meter. The prototype straw neutron detector provides a large-area, efficient, lightweight, more granular (than fission meter) neutron-responsive detection surface (to facilitate imaging) to enhance the ease of application of fission meters. Presented here are the results of preliminary investigations, modeling, and engineering considerations leading to the construction of this prototype. This design is capable of multiplicity and Feynman variance measurements. This prototype may lead to a near-term solution to the crisis that has arisen from the global scarcity of 3He by offering a viable alternative to fission meters.This paper describes the work performed during a 2-year site-directed research and development (SDRD) project that incorporated straw detectors for neutron multiplicity counting. The NMC is a two-panel detector system. We used 10B (in the form of enriched boron carbide: 10B4C) for neutron detection instead of 3He. In the first year, the project worked with a panel of straw neutron detectors, investigated its characteristics, and developed a data acquisition (DAQ) system to collect neutron multiplicity information from spontaneous fission sources using a single panel consisting of 60 straws equally distributed over three rows in high-density polyethylene moderator. In the following year, we developed the field-programmable gate array and associated DAQ software. This SDRD effort successfully produced a prototype NMC with ˜33% detection efficiency compared to a commercial fission meter.

Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated keff of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor ks and external neutron source efficiency φ∗ in great details. Experiments with D-T neutrons are also underway.

Hybrid-3Hen --- New Detector for Gammas and Neutrons

Recently, a unique detector system for beta-delayed neutron spectros\-copy, Hybrid-3Hen, was constructed and used in an experiment. This detector is a modification of the neutron detector 3Hen array, which uses \mbox{$^3$He-filled} tubes at 10 atm and high density polyethylene moderator. In Hybrid-3Hen the gamma-ray detection capability was implemented by\break adding two large HPGe clover-type detectors from the CLARION/CARDS array. This modification of 3Hen enabled efficient studies of beta-delayed neutron emission by using neutron--gamma coincidences. The system is equipped with full-digital readout for easy implementation of the slow and fast coincidences between various parts of the entire detection system. This new array was used in measurements at the Holifield Radioactive Ion Beam Facility and proven to be an essential tool to measure the large beta delayed two neutron emission branching ratio from the exotic isotope $^{86}$Ga.

On Neutron Spectroscopy Using Gas Proportional Detectors Optimized by Transport Theory

We explore in this study the practical limits in designing a neutron detector array to resolve the spectra from special nuclear material (SNM) neutron sources using 3He detectors. We demonstrate that radiation transport analysis yielded a spectrum unfolding strategy based on the energy structure of the BUGLE-96 cross-section library, with 47 neutron energy groups. The initial computational model used is an isotropic planar source incident on a block of high-density polyethylene moderator. Assuming 3He is diluted throughout the moderator, the 3He(n,p) reaction rate energy group matrix in the block was computed using a completely “flat” neutron source spectrum. Analyzing the energy group matrix, there are neutrons from specific collections of energy groups (energy “bands”) that induce a maximum reaction rate in specific locations; we determined that these groups cannot be further differentiated within the energy band using 3He detectors. It was determined that optimal spectral fidelity for SNM detection and characterization is achievable using four spectral bands spanning among groups 1 through 29 (31.8 keV to 17.3 MeV). Using ideal-filter materials to remove the neutrons from different regions of the spectrum, we predicted the maximum neutron spectral resolution obtainable using this approach. To demonstrate our method, we present the optimally detected spectral differences between SNM materials (plutonium and uranium), metal and oxide, using ideal-filter materials. We have also selected a number of candidate filtering materials and, by replacing the ideal filters with real materials, we exemplified our approach with a design of a neutron detector array capable of resolving the spectra from SNM neutron sources using 3He detectors.

Design calculations of an accelerator based BSA for BNCT of brain cancer

Monte Carlo simulations were carried out to calculate optimum design parameters of an accelerator based Beam Shaping Assembly (BSA) for Boron Neutron Capture Therapy (BNCT) of Brain Cancer setup. Epithermal beam of neutrons were obtained through moderation fast neutrons from 7Li(p,n) reaction in a high density polyethylene moderator and a graphite reflector. The dimensions of the moderator and the reflector were optimized through optimization of epithermal/fast neutron intensity ratio as a function of geometric parameters of the setup. Results of our calculation showed the capability of our setup to treat the tumor within 4 cm of the head surface. The calculated Peak Therapeutic Ratio for the setup was found to be 2.15. With further improvement in the polyethylene moderator design and brain phantom irradiation-arrangement, the setup capabilities can be improved to reach further deep-seated tumor.

New source–moderator geometry to improve performance of 252Cf and 241Am–Be source-based PGNAA setups

The gamma ray yield from a 252Cf and a 241Am–Be source-based Prompt Gamma Ray Activation Analysis (PGNAA) setup has been observed to increase with enclosing their neutrons sources in a high-density polyethylene moderator. The prompt gamma rays yield from both setups depends upon the moderator length and the source position in it. For both setups, the optimum moderator length is found to be 7 cm. The optimum position of the neutron source inside moderator of the 252Cf and the 241Am–Be source-based PGNAA setups was found to be at a distance of 0.5 and 0.75 cm from the moderator-end facing the sample, respectively. Due to enclosure of the source in the moderator, about three-fold increase has been observed in the yield of prompt gamma rays from a Portland cement sample of a 252Cf and a 241Am–Be source-based PGNAA setups.