Measurements of neutron distribution in neutrons–γ-rays mixed field using imaging plate for neutron capture therapy

The imaging plate (IP) technique is tried to be used as a handy method to measure the spatial neutron distribution via the ¹⁵⁷Gd(n,γ)¹⁵⁸Gd reaction for neutron capture therapy (NCT). For this purpose, IP is set in a water phantom and irradiated in a mixed field of neutrons and γ-rays. The Hiroshima University Radiobiological Research Accelerator is utilized for this experiment. The neutrons are moderated with 20-cm-thick D₂O to obtain suitable neutron field for NCT. The signal for IP doped with Gd as a neutron-response enhancer is subtracted with its contribution by γ-rays, which was estimated using IP without Gd. The γ-ray response of Gd-doped IP to non-Gd IP is set at 1.34, the value measured for ⁶⁰Co γ-rays, in estimating the γ-ray contribution to Gd-doped IP signal. Then measured distribution of the ¹⁵⁷Gd(n,γ)¹⁵⁸Gd reaction rate agrees within 10% with the calculated value based on the method that has already been validated for its reproducibility of Au activation. However, the evaluated distribution of the ¹⁵⁷Gd(n,γ)¹⁵⁸Gd reaction rate is so sensitive to γ-ray energy, e.g. the discrepancy of the ¹⁵⁷Gd(n,γ)¹⁵⁸Gd reaction rate between measurement and calculation becomes 30% for the photon energy change from 33 keV to 1.253 MeV.     

Excitation functions of proton-induced reactions on natFe and enriched 57Fe with particular reference to the production of 57Co

Excitation functions of the reactions ^natFe(p,xn)^55,56,57,58Co, ^natFe(p,x)⁵¹Cr, ^natFe(p,x)⁵⁴Mn, ⁵⁷Fe(p,n)⁵⁷Co and ⁵⁷Fe(p,α)⁵⁴Mn were measured from their respective thresholds up to 18.5 MeV, with particular emphasis on data for the production of the radionuclide ⁵⁷Co (T_1/2=271.8 d). The conventional stacked-foil technique was used, and the samples for irradiation were prepared by an electroplating or sedimentation process. The measured excitation curves were compared with the data available in the literature as well as with results of nuclear model calculations. From the experimental data, the theoretical yields of the investigated radionuclides were calculated as a function of the proton energy. Over the energy range E_p=15→5 MeV the calculated yield of ⁵⁷Co from the ⁵⁷Fe(p,n)⁵⁷Co process amounts to 1.2 MBq/μA h and from the ^natFe(p,xn)⁵⁷Co reaction to 0.025 MBq/μA h. The radionuclidic impurity levels are discussed. Use of highly enriched ⁵⁷Fe as target material would lead to formation of high-purity ⁵⁷Co.     

Concrete enclosure for shielding a neutron source

In the aim to design a shielding for a 0.185 TBq ²³⁹PuBe isotopic neutron source several Monte Carlo calculations were carried out using MCNP5 code. First, a point-like source was modeled in vacuum and the neutron spectrum and ambient dose equivalent were calculated at several distances ranging from 5 cm up to 150 cm, these calculations were repeated modeling a real source, including air, and a 1×1×1 m³ enclosure with 5, 15, 20, 25, 30, 50 and 80 cm-thick Portland type concrete walls. At all the points located inside the enclosure neutron spectra from 10⁻⁸ up to 0.5 MeV were the same regardless the distance from the source showing the room-return effect in the enclosure, for energies larger than 0.5 MeV neutron spectra are diminished as the distance increases. Outside the enclosure it was noticed that neutron spectra becomes “softer” as the concrete thickness increases due to reduction of mean neutron energy. With the ambient dose values the attenuation curve in terms of concrete thickness was calculated.     

Dose profile modeling of Idaho National Laboratory's active neutron interrogation laboratory

A new laboratory has been commissioned at Idaho National Laboratory for performing active neutron interrogation research and development. The facility is designed to provide radiation shielding for deuterium–tritium (DT) fusion (14.1 MeV) neutron generators (2×10⁸ n/s), deuterium–deuterium (DD) fusion (2.5 MeV) neutron generators (1×10⁷ n/s), and ²⁵²Cf spontaneous fission neutron sources (6.96×10⁷ n/s, 30 μg). Shielding at the laboratory is comprised of modular concrete shield blocks 0.76 m thick with tongue-in-groove features to prevent radiation streaming, arranged into one small and one large test vault. The larger vault is designed to allow operation of the DT generator and has walls 3.8 m tall, an entrance maze, and a fully integrated electrical interlock system; the smaller test vault is designed for ²⁵²Cf and DD neutron sources and has walls 1.9 m tall and a simple entrance maze. Both analytical calculations and numerical simulations were used in the design process for the building to assess the performance of the shielding walls and to ensure external dose rates are within required facility limits. Dose rate contour plots have been generated for the facility to visualize the effectiveness of the shield walls and entrance mazes and to illustrate the spatial profile of the radiation dose field above the facility and the effects of skyshine around the vaults.     

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method (¹⁹⁷Au (n, γ) ¹⁹⁸Au and ⁵⁹Co (n, γ) ⁶⁰Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.     

High-power electron beam tests of a liquid-lithium target and characterization study of 7Li(p,n) near-threshold neutrons for accelerator-based boron neutron capture therapy

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center (SNRC). The target is intended to demonstrate liquid-lithium target capabilities to constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals. The lithium target will produce neutrons through the ⁷Li(p,n)⁷Be reaction and it will overcome the major problem of removing the thermal power >5 kW generated by high-intensity proton beams, necessary for sufficient therapeutic neutron flux.In preliminary experiments liquid lithium was flown through the target loop and generated a stable jet on the concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power densities of more than 4 kW/cm² and volumetric power density around 2 MW/cm³ at a lithium flow of ~4 m/s, while maintaining stable temperature and vacuum conditions. These power densities correspond to a narrow (σ=~2 mm) 1.91 MeV, 3 mA proton beam. A high-intensity proton beam irradiation (1.91–2.5 MeV, 2 mA) is being commissioned at the SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator.In order to determine the conditions of LiLiT proton irradiation for BNCT and to tailor the neutron energy spectrum, a characterization of near threshold (~1.91 MeV) ⁷Li(p,n) neutrons is in progress based on Monte-Carlo (MCNP and Geant4) simulation and on low-intensity experiments with solid LiF targets. In-phantom dosimetry measurements are performed using special designed dosimeters based on CR-39 track detectors.     

A TPD and AR based comparison of accelerator neutron irradiation fields between 7Li and W targets for BNCT

The characteristics of moderator assembly dimension was investigated for the usage of ⁷Li(p,n) neutrons by 2.3–2.8 MeV protons and W(p,n) neutrons by 50 MeV protons. The indexes were the treatable protocol depth (TPD) and advantage depth (AD). Consequently, a configuration for W target with the Fe filter, Fluental moderator, Pb reflector showed the TPD of 5.8 cm and AD of 9.3 cm. Comparable indexes were found for the Li target in a geometry with the MgF₂ moderator and Teflon reflector.     

Evaluation of thermal neutron irradiation field using a cyclotron-based neutron source for alpha autoradiography

It is important to measure the microdistribution of ¹⁰B in a cell to predict the cell-killing effect of new boron compounds in the field of boron neutron capture therapy. Alpha autoradiography has generally been used to detect the microdistribution of ¹⁰B in a cell. Although it has been performed using a reactor-based neutron source, the realization of an accelerator-based thermal neutron irradiation field is anticipated because of its easy installation at any location and stable operation. Therefore, we propose a method using a cyclotron-based epithermal neutron source in combination with a water phantom to produce a thermal neutron irradiation field for alpha autoradiography. This system can supply a uniform thermal neutron field with an intensity of 1.7×10⁹ (cm⁻² s⁻¹) and an area of 40 mm in diameter. In this paper, we give an overview of our proposed system and describe a demonstration test using a mouse liver sample injected with 500 mg/kg of boronophenyl-alanine.     

Preparation of clinical-scale (99)Mo/(99m)Tc column generator using neutron activated low specific activity (99)Mo and nanocrystalline γ-Al(2)O(3) as column matrix

IntroductionPreparation of clinical-scale ⁹⁹Mo/^99mTc generator using (n,γ) activated low specific activity ⁹⁹Mo and nanocrystalline γ-Al₂O₃ as a high capacity sorbent matrix is attempted.MethodsNanocrystalline γ-Al₂O₃ was synthesized by ‘solid state mechanochemical’ reaction of aluminum nitrate with ammonium bicarbonate. Experimental parameters were optimized to effectively separate ^99mTc from ⁹⁹Mo using this sorbent as the column matrix. The performance features of a 13 GBq (350 mCi) ⁹⁹Mo/^99mTc generator using this sorbent and ⁹⁹Mo produced by (n,γ) route having specific activity 12.9–18.5 GBq/g were evaluated for 10 days.ResultsThe sorbent possessed the requisite selectivity for ⁹⁹Mo and demonstrated a maximum sorption capacity of 200 ± 5 mg Mo/g, which is ~ 10 times higher than that of ordinary acidic alumina. The overall yield of ^99mTc was > 80%, with radionuclidic purity > 99.99% and radiochemical purity > 99%. The yield of ^99mTc varied from 7.8 to 2.1 GBq in the eluate for the six days of operation of the generator. The radioactive concentration of ^99mTc eluted was adequate for the formulation of radiopharmaceuticals. The performance of the generator remained consistent over an extended period of 10 days. The eluted ^99mTc was suitable for the formulation of ^99mTc-DMSA and ^99mTc-EC resulting in high radiolabeling yields (> 98%).ConclusionThe effectiveness of γ-Al₂O₃ as a new generation sorbent in the development of clinically useful ⁹⁹Mo/^99mTc generator using low specific activity ⁹⁹Mo and yielding ^99mTc with adequate radioactive concentration and high purity suitable for formulation of radiopharmaceuticals is demonstrated.     

A Monte Carlo study on neutron and electron contamination of an unflattened 18-MV photon beam

Recent studies on flattening filter (FF) free beams have shown increased dose rate and less out-of-field dose for unflattened photon beams. On the other hand, changes in contamination electrons and neutron spectra produced through photon (E>10 MV) interactions with linac components have not been completely studied for FF free beams. The objective of this study was to investigate the effect of removing FF on contamination electron and neutron spectra for an 18-MV photon beam using Monte Carlo (MC) method. The 18-MV photon beam of Elekta SL-25 linac was simulated using MCNPX MC code. The photon, electron and neutron spectra at a distance of 100 cm from target and on the central axis of beam were scored for 10×10 and 30×30 cm² fields. Our results showed increase in contamination electron fluence (normalized to photon fluence) up to 1.6 times for FF free beam, which causes more skin dose for patients. Neuron fluence reduction of 54% was observed for unflattened beams. Our study confirmed the previous measurement results, which showed neutron dose reduction for unflattened beams. This feature can lead to less neutron dose for patients treated with unflattened high-energy photon beams.     

Design and construction of a thermal neutron beam for BNCT at Tehran Research Reactor

An irradiation facility has been designed and constructed at Tehran Research Reactor (TRR) for the treatment of shallow tumors using Boron Neutron Capture Therapy (BNCT). TRR has a thermal column which is about 3 m in length with a wide square cross section of 1.2×1.2 m². This facility is filled with removable graphite blocks. The aim of this work is to perform the necessary modifications in the thermal column structure to meet thermal BNCT beam criteria recommended by International Atomic Energy Agency. The main modifications consist of rearranging graphite blocks and reducing the gamma dose rate at the beam exit. Activation foils and TLD700 dosimeter have been used to measure in-air characteristics of the neutron beam. According to the measurements, a thermal flux is 5.6×10⁸ (n cm⁻² s⁻¹), a cadmium ratio is 186 for gold foils and a gamma dose rate is 0.57 Gy h⁻¹.     

Effects of neutron irradiation on optical and chemical properties of CR-39: Potential application in neutron dosimetry

Effects of high-dose neutron irradiation on chemical and optical properties of CR-39 were studied using FTIR (Fourier Transform Infrared) and UV–vis (Ultraviolet–Visible) spectroscopy. The primary goal was to find a correlation between the neutron dose and the corresponding changes in the optical and chemical properties of CR-39 resulted from the neutron irradiation. The neutrons were produced by bombarding a thick Be target with 22-MeV protons. In the FTIR spectra, prominent absorbance peaks were observed at 1735 cm⁻¹ (CO stretching), 1230 cm⁻¹(C–O–C stretching), and 783 cm⁻¹(C–H bending), the intensities of which decreased with increasing neutron dose. The optical absorbance in the visible range increased linearly with the neutron dose. Empirical relations were established to estimate neutron doses from these optical properties. This technique is particularly useful in measuring high doses, where track analysis with an optical microscope is difficult because of track overlapping.     

Simultaneous production of high specific activity 64Cu and 61Co with 11.4 MeV protons on enriched 64Ni nuclei

The ⁶⁴Cu and ⁶¹Co radionuclides were produced simultaneously by irradiation of enriched ⁶⁴Ni on a low energy proton-only cyclotron. Nickel targets were prepared by electrodeposition of enriched ⁶⁴Ni (>95%) on Au backing at thicknesses of 25–225 mg/cm² with efficiencies >99%. Irradiations up to 30 μA for 8 h were performed with 11.4 MeV protons using a water-cooled target mounting. Radiochemical separation of ⁶⁴Cu and ⁶¹Co from ⁶⁴Ni was performed by chromatography of the chlorocomplexes in a single step using an anion exchange resin column with a yield >95%. Using this method, the Ni target material was recovered and re-plated for subsequent production runs with an overall efficiency >96%. The excitation function for the ⁶⁴Ni(p,n)⁶⁴Cu reaction was measured and compared with published values. Experimental thick target saturation yields of 159 mCi/μA for ⁶⁴Cu and 715 μCi/μA for ⁶¹Co were achieved. Typical specific activities of ⁶⁴Cu were found to be 18.8±3.3 Ci/μmol.     

Experimental studies and nuclear model calculations on proton induced reactions on manganese up to 45 MeV with reference to production of 55Fe, 54Mn and 51Cr

Excitation functions of the reactions ⁵⁵Mn(p,n)⁵⁵Fe, ⁵⁵Mn(p,x)⁵⁴Mn and ⁵⁵Mn(p,x)⁵¹Cr were measured from their respective thresholds up to 18 MeV in the first case and up to 45 MeV in the latter two cases, using the conventional stacked-foil technique. The radioactivity of ⁵⁵Fe was determined via high resolution X-ray spectrometry and of other radionuclides via high resolution γ-ray spectrometry. Nuclear model calculations were performed using the codes ALICE-IPPE, EMPIRE and TALYS. In some cases, good agreement was found between the experimental and theoretical data while in others considerable deviations were observed. From the experimental data the expected integral yields of the three investigated radionuclides were calculated.     

Beam shaping assembly optimization for 7Li(p,n)7Be accelerator based BNCT

Within the framework of accelerator-based BNCT, a project to develop a folded Tandem-ElectroStatic-Quadrupole accelerator is under way at the Atomic Energy Commission of Argentina. The proposed accelerator is conceived to deliver a proton beam of 30 mA at about 2.5 MeV. In this work we explore a Beam Shaping Assembly (BSA) design based on the ⁷Li(p,n)⁷Be neutron production reaction to obtain neutron beams to treat deep seated tumors.     

Neutron radiography using neutron imaging plate

The aims of this research are to study properties of a neutron imaging plate (NIP) and to test it for use in nondestructive testing (NDT) of materials. The experiments were carried out by using a BAS-ND 2040 Fuji NIP and a neutron beam from the Thai Research Reactor TRR-1/M1. The neutron intensity and Cd ratio at the specimen position were approximately 9×10⁵ ns/cm² s and 100 respectively. It was found that the photostimulated luminescence (PSL) readout of the imaging plate was directly proportional to the exposure time and approximately 40 times faster than the conventional NR using Gd converter screen/X-ray film technique. The sensitivities of the imaging plate to slow neutron and to Ir-192 gamma-rays were found to be approximately 4.2×10^?3 PSL/mm² per neutron and 6.7×10^?5 PSL/mm² per gamma-ray photon respectively. Finally, some specimens containing light elements were selected to be radiographed with neutrons using the NIP and the Gd converter screen/X-ray film technique. The image quality obtained from the two recording media was found to be comparable.     

Cross section measurement for the 95Mo(n, α)92Zr reaction at 4.0, 5.0 and 6.0 MeV

Measurements of cross sections of the ⁹⁵Mo(n, α)⁹²Zr reaction at E_n=4.0, 5.0 and 6.0 MeV were carried out at the 4.5 MV Van de Graaff of Peking University, China. A twin gridded ionization chamber and two large-area ⁹⁵Mo samples were adopted. Fast neutrons were produced through the D(d, n)³He reaction by using a deuterium gas target. A small ²³⁸U fission chamber was employed for absolute neutron flux determination. Present data are compared with existing evaluations and measurement.     

Distribution of spatial photoneutrons inside a 70 kg water phantom via neutron activation analysis

This study evaluated spatial Φ_th inside a 70 kg water phantom using the NAA method. Fifty indium foils were placed inside the water phantom and exposed under 15 MV LINAC for 2.5 min to yield the 10 Gy X-ray dose. The Φ_th value at the isocenter of the water was 1.03×10⁶ n cm^?2/Gy-X, and the maximum quantity of Φ_th appeared at the water surface along the z-axis, 3.99×10⁶ n cm^?2/Gy-X. The thermal neutron dose at isocenter of the water phantom occupied approximately 0.151% of the whole photo and neutron dose.     

Neutron activation cross-sections for Ytterbium isotopes at (14.6±0.3) MeV

The cross-sections of ^natYb (n,x)^172,173 Tm, ¹⁷⁴Yb(n,p) ¹⁷⁴ Tm, ¹⁷⁴Yb (n,α) ¹⁷¹Er, ¹⁷⁶Yb(n,p) ¹⁷⁶ Tm, ¹⁷⁶Yb(n,α)¹⁷³ Er, and ¹⁷⁶ Yb(n,n′)^176mYb have been measured at 14.6±0.3 MeV neutron energy, among them two cross-sections ^natYb (n,x)^172,173Tm are reported for the first time. These experimental cross-sections are compared with experimental data found in the literature, with evaluated nuclear data in JENDL-4.0 and TENDL-2010 libraries and with theoretically calculated values based on nuclear reaction modular codes EMPIRE-3.0 and TALYS-1.2.     

Measurements of isomeric cross sections for the (n,α) reaction on the 142Nd isotope at approximately 14 MeV neutrons

In this study, the activation cross sections were measured for ¹⁴²Nd(n,α)^139mCe reaction at four neutron energies between 13.57 and 14.83 MeV, which were produced by a neutron generator through ³H(²H,n)⁴He reaction. The production of short-lived activity and the spectra accumulation were performed by the cyclic activation technique. Induced gamma-ray activities were measured using a high resolution gamma ray spectrometer equipped with a high-purity Germanium (HpGe) detector. In the cross section measurements, corrections were made regarding the effects of the gamma-ray attenuation, the dead-time, the fluctuation of the neutron flux, and low energy neutrons. The measured cross sections were compared with the published literature and the results of the model calculation (TALYS 1.4).