The multi-slit very small angle neutron scattering instrument at the China Spallation Neutron Source.

A multi-slit very small angle neutron scattering (MS-VSANS) instrument has been finally accepted at the China Spallation Neutron Source (CSNS). It is the first spallation neutron source based VSANS instrument. MS-VSANS has a good signal-to-noise ratio and can cover a wide scattering vector magnitude range from 0.00028 to 1.4 Å-1. In its primary flight path, a combined curved multichannel beam bender and sections of rotary exchange drums are installed to minimize the background downstream of the instrument. An exchangeable multi-slit beam focusing system is integrated into the primary flight path, enabling access to a minimum scattering vector magnitude of 0.00028 Å-1. MS-VSANS has three modes, namely conventional SANS, polarizing SANS and VSANS modes. In the SANS mode, three motorized high-efficiency 3He tube detectors inside the detector tank cover scattering angles from 0.12 to 35° simultaneously. In the polarizing SANS mode, a double-V cavity provides highly polarized neutrons and a high-efficiency 3He polarization analyser allows full polarization analysis. In the VSANS mode, an innovative high-resolution gas electron multiplier detector covers scattering angles from 0.016 to 0.447°. The absolute scattering intensities of a selection of standard samples are obtained using the direct-beam technique; the effectiveness of this method is verified by testing the standard samples and comparing the results with those from a benchmark instrument. The MS-VSANS instrument is designed to be flexible and versatile and all the design goals have been achieved.

Quantitative investigation of gamma radiation in accelerator produced ISO neutron fields.

In standard monoenergetic ISO neutron fields, the neutron yield of neutron-producing reactions was measured in combination with the prompt photon yield, including photon energies up to 10 MeV, for the purpose of comparing the two yields. Separating the photons produced by the target (direct photons) from those generated by secondary neutron reactions was achieved using the time-of-flight method. Photon and neutron ambient dose equivalent values were calculated from measured spectral energy distributions. Quasi monoenergetic neutron fields are needed to systematically test the response of measuring instruments to neutron radiation. For this reason, ISO has defined a number of reference neutron radiation fields covering a wide energy range up to 19 MeV. Because neutron detectors may also be affected by photon radiation, the photon fluence in the ISO neutron fields has to be known. This work focuses on quasi monoenergetic accelerator-produced neutron fields in the energy range of 24 keV to 19 MeV.

Overview of Boron Neutron Capture Therapy in 2022.

In April 2022, the National Cancer Institute of the United States organized a 3-day seminar, dedicated to boron neutron capture therapy (BNCT). This short article summarizes a presentation from that event, which is intended to provide an overview of activities currently underway worldwide to make BNCT available for patient treatments. This overview does not claim to be exhaustive but shows a great deal of activity in all areas necessary for the complex therapy that is BNCT. A rapid increase in the number of BNCT centers can be expected over the next few years, coupled with the introduction of novel drugs for BNCT. It will be a major challenge to all stakeholders to create clinical networks that can conduct the necessary prospective clinical trials in a short time and in high quality.

Design and construction of a 9 MeV γ-ray source based on capture of moderated plutonium-beryllium neutrons in nickel.

We have designed and constructed a high-energy γ-ray source for detector characterisation and calibration. The source is a composite type based on a plutonium-beryllium neutron emitter embedded in a paraffin moderator, which is homogeneously mixed with nickel powder. The 9 MeV γ-ray source produces approximately 450 photons per second in 4π when 2.2×105 neutrons per second are emitted, corresponding to a surface flux of 9 MeV γ-rays of approximately 2.5×10-6 cm-2 per emitted neutron. Here we discuss the properties and design of this source, including the characterisation of homogeneity and high-energy γ-ray emission spectra.

[Basic Knowledge of Neutron: Physics].

Neutrons are uncharged particles and exhibit strong ability to penetrate matter. Various charged particles and gamma rays are emitted from nuclear reactions induced by neutrons passing through the matter. It is important to consider contributions of neutrons for estimating such quantities used in radiation protection as absorbed dose, equivalent dose and effective dose. In this article basic knowledge of neutron is briefly summarized concerned with physical properties of neutrons, neutron reactions, neutron sources, fluence to karma conversion coefficient, absorbed dose, equivalent dose, effective dose.

Structure evolution of nanodiamond aggregates: a SANS and USANS study.

Ultra-small-angle neutron scattering (USANS) and small-angle neutron scattering (SANS) measurements, covering length scales from micrometres to nanometres, were made to investigate the structure of nanodiamonds (NDs) and their suspensions. These nanodiamonds were produced by two different techniques, namely by the detonation method and by the laser ablation of a carbon-hydro-carbon mixture. The (U)SANS results indicated the presence of structures four orders of magnitude larger than the dimensions of a single ND particle, consisting of aggregations of ND particles. This aggregation of the ND particles was studied by employing the contrast variation technique. Two different solvents, namely H2O and dimethyl sulfoxide (and their deuterated counterparts), were used to understand the role of hydrogen in the shape and size of the aggregates. The analysis of experimental data from SANS measurements also reveals the ND particles to have an ellipsoidal structure. Using a defined shape model and the SANS contrast variation technique, it was possible to characterize the non-diamond outer shell of the particles and determine the outer layer thickness. This clarification of the structure of the NDs will allow better preparation of suspensions/samples for various applications. Understanding the structure of NDs at multiple length scales also provides crucial knowledge of particle-particle interaction and its effect on the aggregation structures.

Importance of radiobiological studies for the advancement of boron neutron capture therapy (BNCT).

Boron neutron capture therapy (BNCT) is a tumour selective particle radiotherapy, based on the administration of boron carriers incorporated preferentially by tumour cells, followed by irradiation with a thermal or epithermal neutron beam. BNCT clinical results to date show therapeutic efficacy, associated with an improvement in patient quality of life and prolonged survival. Translational research in adequate experimental models is necessary to optimise BNCT for different pathologies. This review recapitulates some examples of BNCT radiobiological studies for different pathologies and clinical scenarios, strategies to optimise boron targeting, enhance BNCT therapeutic effect and minimise radiotoxicity. It also describes the radiobiological mechanisms induced by BNCT, and the importance of the detection of biomarkers to monitor and predict the therapeutic efficacy and toxicity of BNCT alone or combined with other strategies. Besides, there is a brief comment on the introduction of accelerator-based neutron sources in BNCT. These sources would expand the clinical BNCT services to more patients, and would help to make BNCT a standard treatment modality for various types of cancer. Radiobiological BNCT studies have been of utmost importance to make progress in BNCT, being essential to design novel, safe and effective clinical BNCT protocols.

Comparison of several Li doped scintillators for measurement of neutron and γ radiation integral quantities.

Five types of scintillators for parallel detection of neutrons and γ were tested for their pulse shape discrimination ability. All detectors are based on 6Li (n,alpha)3H reaction. LiI, ZnS and polystyrene were used as scintillators. Tests were performed at different neutron and γ mixed fields (AmBe, Cf) using a small Hamamatsu photomultiplier and a Picoscope digitizer. A polystyrene-based ZnS + LiF detector was proven to be the most suitable for neutron-γ separation and similar sensitivity for both types of radiation to construct a single device to measure integral quantities neutrons and γ (fluence, ambient dose equivalent) in parallel. The ZnS + LiF detector based on plexiglass, is in principle useable also, but has low sensitivity to γ radiation. The tested Li glass or LiI crystal does not have the ability to separate neutron and γ. With amplitude discrimination it can be used as a neutron detector for a much simpler probe for a Bonner spectrometer.

Characterization of a plutonium-beryllium neutron source.

Here we present an investigation of a plutonium-beryllium neutron source available at the Horia Hulubei National Institute of Physics and Nuclear Engineering, Romania, to be used for detector characterization during the implementation of the Extreme Light Infrastructure - Nuclear Physics project. Using several different techniques and instruments, we have measured the isotopic composition for plutonium to be 75% 239Pu and 24% 240Pu, with a minor contribution from other isotopes. Furthermore, we have measured the source activity as of November 20th 2019 to be 2.220(5)×105 neutrons per second with a mean energy of 3.25(17) MeV. We have also measured both the γ-tagged and full neutron energy spectra, and discuss the origin of the observed structure in the neutron energies based on these. All these parameters are of importance both for traceability of nuclear material, radioprotection, and accurate detector characterization.

Design of an Am-Be neuron source experimental platform at Sichuan University.

Am-Be neutron sources are widely used in nuclear laboratories. It is important to design adequate shielding for Am-Be neutron sources, in order to protect experimenters from being irradiated by neutrons and their associated γ-rays. When designing shielding, it is also much desired for the source assembly to provide sufficient neutron moderation, and furnish access channels for both thermal and fast neutrons directly from the source. Such a neutron source assembly may be called a "neutron source experimental platform". This paper presents the design details of an Am-Be neutron source experimental platform at Sichuan University. The experimental platform is designed to be a Φ 100 cm × 100 cm paraffin cylinder with the Am-Be source at the center. The paraffin cylinder is inside a 0.5 cm thick stainless-steel container lined with 2 cm thick borated polyethylene. The platform has multiple access channels for thermal neutrons, and one access channel for fast neutrons. The design calculations were performed with Geant4.

Characterization of 244Cm neutron sources.

There is a need to find a replacement for the 252Cf sources currently used in Standards-based performance testing of neutron instrumentation. This is a result of its relatively short half-life, concerns about future availability and recent increases in cost. A potential replacement source, 244Cm is evaluated in this study where two commercially available sources have been acquired and their neutron emission spectra measured using a pair of spectrometers. Both instruments showed the 244Cm spontaneous neutron spectrum to be fully consistent with a MCNP6®-calculated spectrum using published Watt fission parameters for 244Cm. In addition, the emission rate of the weaker 244Cm source was established through a direct count rate comparison against a calibrated and similarly encapsulated 252Cf source. The 244Cm source emission rate was found to be in excellent agreement with value stated on the manufacturer's source certificate. It is concluded that 244Cm would be an ideal replacement for 252Cf based on its longer half-life (18.1 y) and its essentially identical neutron emission spectrum. In addition, 244Cm sources are commercially available at reasonable cost.

Boron neutron capture therapy at the crossroads - Where do we go from here?

As elegant as is the concept upon which Boron Neutron Capture Therapy (BNCT) is based, unfortunately it has not gained widespread acceptance by the physicians who are treating cancer patients on a daily basis. The question is why? Very simply put, the clinical results obtained in treating patients with high grade gliomas and recurrent tumors of the head and neck region have not been convincing enough to produce more interest in BNCT as a cancer treatment modality. There are a variety of reasons for this, one of the most important of which has been its dependency on nuclear reactors as neutron sources. With the advent of accelerator based neutron sources (ABNS), this hopefully will be addressed. If the results obtained from ongoing and soon to be initiated clinical trials can at least demonstrate equivalency to those obtained with nuclear reactors, this should address the first problem. The second problem relates to boron delivery agents, and despite the considerable efforts of chemists and biologists over the past 50 years, there are only two drugs that currently are being used clinically, sodium borocaptate (BSH) and boronophenylalanine (BPA). It is widely recognized that these two drugs are less than ideal. Perhaps new and more effective boron delivery agents will finally appear on the scene, but barring that, we will address the question of what can be done now to make BNCT a more effective cancer treatment modality.

In-house texture measurement using a compact neutron source.

In order to improve the instrumental accessibility of neutron diffraction techniques, many emerging compact neutron sources and in-house neutron diffractometers are being developed, even though the precision level of neutron diffraction experiments performed on such instruments was thought to be incomparable with that of large-scale neutron facilities. As a challenging project, the RIKEN accelerator-driven compact neutron source (RANS) was employed here to establish the technical environment for texture measurements, and the recalculated pole figures and orientation distribution functions of an interstitial-free steel sheet obtained from RANS were compared with the results from another two neutron diffractometers well established for texture measurement. These quantitative comparisons revealed that the precise neutron diffraction texture measurement at RANS has been realized successfully, and the fine region division of the neutron detector panel is invaluable for improving the stereographic resolution of texture measurements. Moreover, through selectively using the parts of the obtained neutron diffraction patterns that exhibit good statistics, the Rietveld texture analysis improves the reliability of the texture measurement to a certain extent. These technical research results may accelerate the development of other easily accessible techniques for evaluation of engineering materials using compact neutron sources, and also help to improve the data-collection efficiency for various time-resolved scattering experiments at large-scale neutron facilities.

Nedis-Serpent simulation of a neutron source assembly with complex internal heterogeneous structure.

The hybrid use of Nedis-2m and Serpent 2.1.30 codes to predict the radiation characteristics (i.e., neutron yield and energy spectrum) of an Am-Be source with a fine-grained mixture of americium dioxide (AmO2) and beryllium (Be) core was studied with a focus on the grain size influence on the simulation results. The study showed that the fine-grained structure of the source core would decrease the number of alpha particles participating in the nuclear reactions with 17,18O and 9Be nuclei, which softened the neutron energy spectrum and reduced the neutron yield. The simulations also confirmed that the source core made of the stable crystals of AmBe13 intermetallic alloy would improve the neutron yield to maximum 50% compared to the core made of AmO2. Moreover, a source with a variable neutron yield was proposed with a heterogeneous core of AmO2 rods embedded in Be. The neutron energy spectrum of heterogeneous source resembled the energy spectrum of Deuterium-Tritium (D-T) neutrons which were generated in a long magnetic trap with high-temperature plasma. The subcritical irradiation facility assembled from the nth number of heterogeneous Am-Be source can be used to study the properties of materials and the equipment operating in the epithermal and fast neutron spectra. The use of a heterogeneous Am-Be assembly, as a basic element of an irradiation installation, simplifies the handling and operation procedures because it is easily disabled by removing the Be layer, or by inserting a sheet of the appropriate size and material between the Be and Am rod.

Natural radionuclides as background sources in the Modane underground laboratory.

The Modane underground laboratory (LSM) is the deepest operating underground laboratory in Europe. It is located under the Fréjus peak in Savoie Alps in France, with average overburden of 4800 m w. e. (water equivalent), providing low-background environment for experiments in nuclear and particle physics, astrophysics and environmental physics. It is crucial to understand individual sources of background such as residual cosmic-ray flux of high-energy muons, muon-induced neutrons and contributions from radionuclides present in the environment. The identified dominant sources of background are radioactive contamination of construction materials of detectors and laboratory walls, radon contamination of the laboratory air, and neutrons produced in the laboratory. The largest neutron contribution has been identified from (α, n) reactions in low Z materials (10-7-10-4 n s-1 Bq-1) and from spontaneous fission of 238U (1.1× 10-6 n s-1 Bq-1).

Specifying the flux and dose-equivalent buildup factors for infinite slabs irradiated by radionuclide neutron sources.

In this work, flux and neutron dose-equivalent buildup parameters are calculated for six radionuclide point-like neutron sources having broad energy spectrum which irradiate infinite slab-like common shielding materials of beryllium, concrete, iron, graphite, water, and lead, employing the MCNPX code. The description of the buildup factor is made in a straightforward way which is analogous to that of gamma. The parameters are obtained for thicknesses of shield from 0.5 to 10 mean free paths (mfp). The achieved dose-equivalent buildup parameters are parameterized by polynomial expressions. Using this parameterization, one can determine these factors for the desired thickness of shield material and each neutron source.

Efficient data extraction from neutron time-of-flight spin-echo raw data.

Neutron spin-echo spectrometers with a position-sensitive detector and operating with extended time-of-flight-tagged wavelength frames are able to collect a comprehensive set of data covering a large range of wavevector and Fourier time space with only a few instrumental settings in a quasi-continuous way. Extracting all the information contained in the raw data and mapping them to a suitable physical space in the most efficient way is a challenge. This article reports algorithms employed in dedicated software, DrSpine (data reduction for spin echo), that achieves this goal and yields reliable representations of the intermediate scattering function S(Q, t) independent of the selected 'binning'.

A 13C(d,n)-based epithermal neutron source for Boron Neutron Capture Therapy.

PURPOSE: Boron Neutron Capture Therapy (BNCT) requires neutron sources suitable for in-hospital siting. Low-energy particle accelerators working in conjunction with a neutron producing reaction are the most appropriate choice for this purpose. One of the possible nuclear reactions is 13C(d,n)14N. The aim of this work is to evaluate the therapeutic capabilities of the neutron beam produced by this reaction, through a 30mA beam of deuterons of 1.45MeV. METHODS: A Beam Shaping Assembly design was computationally optimized. Depth dose profiles in a Snyder head phantom were simulated with the MCNP code for a number of BSA configurations. In order to optimize the treatment capabilities, the BSA configuration was determined as the one that allows maximizing both the tumor dose and the penetration depth while keeping doses to healthy tissues under the tolerance limits. RESULTS: Significant doses to tumor tissues were achieved up to ∼6cm in depth. Peak doses up to 57Gy-Eq can be delivered in a fractionated scheme of 2 irradiations of approximately 1h each. In a single 1h irradiation, lower but still acceptable doses to tumor are also feasible. CONCLUSIONS: Treatment capabilities obtained here are comparable to those achieved with other accelerator-based neutron sources, making of the 13C(d,n)14N reaction a realistic option for producing therapeutic neutron beams through a low-energy particle accelerator.

Simulation study of accelerator based quasi-mono-energetic epithermal neutron beams for BNCT.

Filtered neutron techniques were applied to produce quasi-mono-energetic neutron beams in the energy range of 1.5-7.5 keV at the accelerator port using the generated neutron spectrum from a Li (p, n) Be reaction. A simulation study was performed to characterize the filter components and transmitted beam lines. The feature of the filtered beams is detailed in terms of optimal thickness of the primary and additive components. A computer code named "QMNB-AS" was developed to carry out the required calculations. The filtered neutron beams had high purity and intensity with low contamination from the accompanying thermal, fast neutrons and γ-rays.

Near threshold 7Li(p,n) 7Be reaction as neutron source for BNCT.

(7)Li(p,n)(7)Be is an endothermic reaction and working near its threshold (1.88 MeV) has the advantage of neutron spectra with maximum energies of about 100 keV, considerably lower than at higher beam energies, or than using other neutron-producing reactions or as for the uranium fission spectrum, relevant for BNCT based on nuclear reactors. With this primary energy it is much easier to obtain the energies needed for treating deep seated tumors by BNCT (about 10 keV). This work studies bombarding energies up to 2.05 MeV, different beam incidence angles and the effect of the undesirable gamma production via the (7)Li(p,γp') (7)Li reaction.