γ射线对含氢正比计数管中子能谱测量的影响

通过蒙特卡罗模拟和实验测量相结合的方法,获得从50keV~10MeV区间γ射线在直径为30mm内充1.013 25MPa氢气的球形含氢正比计数管上的能量响应,结果显示,γ射线在该计数管中的能量沉积主要集中在100keV附近及以下。Am-Be中子源和137 Cs源的实验测量结果显示,强137 Cs源的γ射线会严重影响含氢正比计数管对Am-Be中子源100keV以下能谱的测量,这表明,裂变材料介质内的强γ射线同样会影响到介质内100keV以下中子能谱的测量。根据计数管对反冲质子和电子电离信号的收集特性,采用上升时间法甄别掉本底γ射线是可行的。     

Influence of Bonner sphere response functions above 20 MeV on unfolded neutron spectra and doses

Monte Carlo (MC) codes for neutron transport calculations such as MCNP, MCNPX, FLUKA, PHITS, and GEANT4, crucially rely on cross sections that describe the interaction of neutrons with nuclei. For neutron energies below 20 MeV, evaluated cross sections are available that are validated against experimental data. In contrast, for high energies (above 20 MeV) experimental data are scarce and, for this reason, every neutron transport code is based on theoretical nuclear models to describe interactions of neutrons with nuclei in matter. Here we report on the calculation of a complete set of response functions for a Bonner spheres spectrometer (BSS), by means of GEANT4 using the Bertini and Binary Intranuclear Cascade (INC) models for energies above 20 MeV. The recent results were compared with those calculated by MCNP/LAHET and MCNP/HADRON MC codes. It turns out that, whatever code used, the response functions were rather similar for neutron energies below 20 MeV, for all 16 detector/moderator combinations of the considered BSS system. For higher energies, however, differences of more than a factor of 2 were observed, depending on neutron energy, detector/moderator combination, MC code, and nuclear model used. These differences are discussed in terms of neutron fluence rates measured at the environmental research station (UFS), “Schneefernerhaus”, (Zugspitze mountain, Germany, 2650 m a.s.l.) for energies below 0.4 eV (thermal neutrons), between 0.4 eV and 100 keV (epithermal neutrons), between 100 keV and 20 MeV (evaporation neutrons), and above 20 MeV (cascade neutrons). In terms of total neutron fluence rates, relative differences of up to 4% were obtained when compared to the standard MCNP/LAHET results, while in terms of total ambient dose equivalent, relative differences of up to 8% were obtained. Both the GEANT4 Binary INC and Bertini INC gave somewhat larger fluence and dose rates in the epithermal region. Most relevant for dose, however, those response functions calculated with the GEANT4 Bertini INC model provided about 18% less neutrons in the cascade region, which led to a roughly 13% smaller contribution of these neutrons to ambient dose equivalent. It is concluded that doses from secondary neutrons from cosmic radiation as deduced from BSS measurements are uncertain by about 10%, simply because of some differences in nuclear models used by various neutron transport codes.     

Design of a graphite-moderated 241Am–Li neutron field to simulate reactor spectra

A neutron calibration field using ²⁴¹Am–Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the ²⁴¹Am–Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.     

Experimental validation of a Tokamak neutron spectrometer based on Bonner spheres

In order to realize real-time fusion neutron spectrum diagnosis for the HL 2A Tokamak, a Bonner Sphere Spectrometer(BSS) array has been developed, consisting of eight polyethylene spheres(PS) with embedded3 He proportional counters.To validate its spectrometric capability, spectrum measurement of an241Am-Be neutron source was carried out and is described.The Monte Carlo code Geant4 was used to calculate the response functions,taking this interference into consideration.Finally, the neutron spectrum was unfolded in the energy range from10-9MeV to 20 MeV.The unfolded spectrum has remarkable consistency with the ISO 8529-1 standard241Am-Be neutron spectrum which is a preliminary demonstration that this BSS is reliable and practical.     

Design and validation of a single sphere multi-detector neutron spectrometer based on LiF: Mg,Cu,P thermoluminescent dosemeters

This communication describes a new neutron spectrometer consisting of pairs of ⁷Li and ⁶Li based thermoluminescent dosemeters (MCP-6, MCP-7) located at selected positions within a single moderating polyethylene sphere. The spatial arrangement of the dosemeters has been designed using the MCNPX Monte Carlo code to calculate the response matrix in order to obtain a nearly isotropic response for neutrons in the energy range up to 20 MeV. A partial validation of the calculated response matrix has been performed with the calibrated ²⁴¹Am–Be neutron source at the INFN–LNF Laboratory, using the shadow cone technique.     

The influence of field size and off-axis distance on photoneutron spectra of the 18 MV Siemens Oncor linear accelerator beam

At present, high energy electron linear accelerators (LINACs) producing photons with energies higher than 10 MeV have a wide use in radiotherapy (RT). However, in these beams fast neutrons could be generated, which results in undesired contamination of the therapeutic beams. These neutrons affect the shielding requirements in RT rooms and also increase the out-of-field radiation dose to patients. The neutron flux becomes even more important when high numbers of monitor units are used, as in the intensity modulated radiotherapy. Herein, to evaluate the exposure of patients and medical personnel, it is important to determine the full radiation field correctly. A model of the dual photon beam medical LINAC, Siemens ONCOR, used at the University Hospital Centre of Osijek was built using the MCNP611 code. We tuned the model according to measured photon percentage depth dose curves and profiles. Only 18 MV photon beams were modeled. The dependence of neutron dose equivalent and energy spectrum on field size and off-axis distance in the patient plane was analyzed. The neutron source strength (Q) defined as a number of neutrons coming from the head of the treatment unit per x-ray dose (Gy) delivered at the isocenter was calculated and found to be 1.12 × 10¹² neutrons per photon Gy at isocenter. The simulation showed that the neutron flux increases with increasing field size but field size has almost no effect on the shape of neutron dose profiles. The calculated neutron dose equivalent of different field sizes was between 1 and 3 mSv per photon Gy at isocenter. The mean energy changed from 0.21 MeV to 0.63 MeV with collimator opening from 0 × 0 cm² to 40 × 40 cm². At the 50 cm off-axis the change was less pronounced. According to the results, it is reasonable to conclude that the neutron dose equivalent to the patient is proportional to the photon beam-on time as suggested before. Since the beam-on time is much higher when advanced radiotherapy techniques are used to fulfill high conformity demands, this makes the neutron flux determination even more important. We also showed that the neutron energy in the patient plane significantly changes with field size. This can introduce significant uncertainty in dosimetry of neutrons due to strong dependence of the neutron detector response on the neutron energy in the interval 0.1–5 MeV.     

Investigation of fast neutron interaction with235U

The angular distribution of neutrons emitted by elastic, inelastic and fission processes on²³⁵U were measured at the incident neutron energies of 1.5, 1.9, 2.3, 4.0, 4.5, 5.0 and 5.5 MeV using nanosecond time-of-flight technique. The differential elastic scattering cross sections and their angular distributions at all the seven energies are presented. The total elastic scattering cross sections, angle and energy integrated cross sections for the inelastically scattered neutrons in energy bands of 200 keV, fission cross sections and the angular distributions of fission neutrons were extracted at 1.5, 1.9 and 2.3 MeV incident neutron energies. The energy distributions of the prompt fission neutrons and of the inelastically scattered neutrons are given at the incoming neutron energies of 1.5, 1.9 and 2.3 MeV; and the average fission neutron energies and the inelastic neutron evaporation temperatures were also evaluated at these energies.     

Measurement of the neutron spectrum of a Pu-C source with a liquid scintillator

The neutron response function for a BC501A liquid scintillator (LS) has been measured using a series of monoenergetic neutrons produced by the p-T reaction. The proton energies were chosen such as to produce neutrons in the energy range of 1 to 20 MeV. The principles of the technique of unfolding a neutron energy spectrum by using the measured neutron response function and the measured Pulse Height (PH) spectrum is briefly described. The PH spectrum of neutrons from the Pu-C source, which will be used for the calibration of the reactor antineutrino detectors for the Daya Bay neutrino experiment, was measured and analyzed to get the neutron energy spectrum. Simultaneously the neutron energy spectrum of an Am-Be source was measured and compared with other measurements as a check of the result for the Pu-C source. Finally, an error analysis and a discussion of the results are given.     

Unfolding the fast neutron spectra of a BC501A liquid scintillation detector using GRAVEL method

Accurate knowledge of the neutron energy spectra is useful in basic research and applications. The overall procedure of measuring and unfolding the fast neutron energy spectra with BC501A liquid scintillation detector is described. The recoil proton spectrum of 241Am-Be neutrons was obtained experimentally. With the NRESP7 code, the response matrix of detector was simulated. Combining the recoil proton spectrum and response matrix, the unfolding of neutron spectra was performed by GRAVEL iterative algorithm. A MatLab program based on the GRAVEL method was developed. The continuous neutron spectrum of ²⁴¹Am-Be source and monoenergetic neutron spectrum of D-T source have been unfolded successfully and are in good agreement with their standard reference spectra. The unfolded ²⁴¹Am-Be spectrum are more accurate than the spectra unfolded by artificial neural networks in recent years.     

Neutron production with mixture of deuterium and krypton in Sahand Filippov type plasma focus facility

This Letter reports the order of magnitude enhancement in neutron yield from Sahand plasma focus device with krypton seeded deuterium operation. The highest average neutron yield of 2.2×¹⁰⁹ neutrons per shot was achieved at 1.00 Torr deuterium with 3% krypton which is higher than the best average neutron yield of 3.18×¹⁰⁸ neutrons per shot for pure deuterium operation. Estimation of average neutron energy showed that the maximum and minimum average energies are 2.98±0.6 MeV at 16 kV in 0.25 Torr deuterium with 3% Kr and 2.07±0.2 MeV at 18 kV operation in 0.5 Torr deuterium with 3% Kr, respectively. The anisotropy of neutron emission from Sahand DPF showed that the neutrons are produced mainly by beam-target mechanisms.     

On the use of averaging stopping power of recoiled nuclei for determining maximum neutron energy of 241Am–Be by superheated drop detectors

The average stopping power of the recoiled nuclei generated by neutron elastic interactions with the Freon-12 drops in a superheated drop detector has been used to determine the maximum neutron energy of the ²⁴¹Am–Be source. In an elastic interaction of neutrons with the Freon-12 liquid, the nuclei of ¹²C, ¹⁹F and ³⁵Cl with different values of stopping power are scattered. The stopping power of these scattered nuclei corresponding to the energy transferred to them through the head-on collision was extracted from the SRIM code. The stopping power values were weighted by considering the neutron–nucleus elastic scattering cross section and the number of each nucleus in the Freon-12 molecule and the average stopping power was calculated from known neutron energy.The maximum energy of the ²⁴¹Am–Be neutron source was estimated as 10.9 ± 3.0 MeV. The consistency between the determined energy and the other reported values confirms the validity of using the average stopping power in the superheated drop detectors. The average stopping power was also used to determine the threshold neutron energy as a function of external applied pressure at different temperatures. Knowing the threshold neutron energy as function of applied pressure, can be used in pressure scanning method for neutron spectrometry by superheated drop detectors.     

A preliminary study of the performance of a novel design of multi-element tissue equivalent proportional counter for neutron monitoring

In this study we investigate the dependence of the sensitivity of a TEPC upon its surface area and demonstrate that a compact multi-element tissue equivalent proportional counter (METEPC) has a counting sensitivity comparable to a commercially available 12.7 cm (5 inch) diameter spherical TEPC. The METEPC incorporates 61 cylindrical counting volumes of internal diameter of 0.5 cm and height 5 cm, machined in a single block of tissue equivalent plastic. It is the simplest design available in the multi-element geometrical configuration and is approximately nine times smaller in volume than that of a conventional 12.7 cm spherical TEPC. The neutron sensitivity of commercially available TEPCs and the METEPC simulating a 2 μm tissue site size was examined experimentally using the McMaster University 1.25 MV double stage Tandetron accelerator, which produces low energy neutrons via the ⁷Li(p, n)⁷Be reaction. The mean energy of the neutron beams produced ranged from 34 keV to 354 keV. The results presented in this study suggests that the compact METEPC is able to produce measurements in low dose rate radiation environments with the same precision in a given length of time as could be obtained with a 12.7 cm diameter spherical TEPC.     

Dosimetry and spectrometry at accelerator based neutron source for boron neutron capture therapy

An innovative accelerator-based neutron source for boron neutron capture therapy has started operation at the Budker Institute of Nuclear Physics, Novosibirsk. This facility is based on a compact vacuum insulation tandem accelerator designed to produce proton current up to 10 mA. Epithermal neutrons are proposed to be generated by 1.915 MeV protons bombarding a lithium target using ⁷Li(p,n)⁷Be threshold reaction.In the article, techniques to detect neutron and gamma-rays at the facility are described. Gamma radiation is measured with NaI and BGO gamma-spectrometers. The total yield of neutrons is determined by measuring the 477 keV γ-quanta from beryllium decay. For the rough analysis of the generated neutron spectrum we used bubble detectors. As the epithermal neutrons are of interest for neutron capture therapy the NaI detector is used as activation detector. We plan to use a time-of-flight technique for neutron spectra measurement. To realize this technique a new solution of short time neutron generation is proposed.     

Evaluation of target photon dose mixed in mono-energetic neutron fields using 7Li(p,n)7Be reaction

Target photons mixed in the 144, 250 and 565 keV mono-energetic neutron calibration fields were measured using a cylindrical NaI(Tl) detector with 7.62 cm both in diameter and in length. The ambient dose equivalent H*(10) of the photons was evaluated by applying the “G(E) function” to the measured pulse height spectrum. Neutrons induce photons by nuclear reactions in the NaI(Tl) detector and affect the pulse height spectrum. In order to eliminate the influence of these neutron events, the time-of-flight technique was applied with operating the accelerator in the pulse mode. The ratios by the ambient dose equivalent H*(10) of the photons to the 144, 250 and 565 keV neutrons were evaluated to be 3.3%, 4.7% and 0.9%, respectively. Although high energy photons ranging from 6 to 7 MeV are emitted by the ¹⁹F(p,αγ)¹⁶O reactions, the dose of the target photons is low enough to calibrate neutron dosemeters except for ones with high sensitivity to the photons.     

Experimental and theoretical study of long counters on the departure of “point” assumption and scattering background influence

A new long counter has been constructed and calibrated with the inherent efficiency of 11.5 cm² in the 2.5 MeV neutron field, and a relative flat energy response of 20% disparity within 0.5 MeV–15 MeV. Using this detector and another commercial DePangher counter, the departure of “point” assumption at near distance and the scattering background influence are investigated by performing three groups of experiments in low, middle and high scattering room. Simulation studies are also performed to supplement the experimental data.Two conclusions are deduced. First, the direct efficiency η is very sensitive when the source is very near, and the departure of point assumption is very evident at this situation. So, a revising method is tried to put forward, employing the on-site calibration with a known neutron source and theoretical calculated coefficient D = (η/ε)/(η_cf252/ε_cf252). And second, scattering backing counts are found mainly from the nearest wall or scatterers. However, the relatively uniform isotropic scattered neutrons contribute insensitive to the source-detector distance, which means the background tally can be eliminated by some simple ways.     

Study of n-γ discrimination for 0.4-1 MeV neutrons using the zero-crossing method with a BC501A liquid scintillation detector

An experimental system aimed at n-γ discrimination using the zero-crossing method with a φ3"× 2" BC501A liquid scintillation detector was established and tested with an Am-Be neutron source. Two-dimensional plots of energy versus zero-crossing time were obtained. The quality of n-γ discrimination was checked by the figure-of-merit (FOM), the neutron peak-to-valley ratio, and the proportion of leaked neutrons over all neutron events. The performance of n-γ discrimination in terms of FOM was compared with previous work done by other groups. The n-γ discrimination in four different energy regions with an interval of 0.1 MeV between 0.3 MeV and 0.7 MeV was studied, and the results indicate that the n-γ discrimination threshold can go down to 0.4 MeV.     

The role of the 14N(n,p)14C reaction in neutron irradiation of soft tissues

The cell-killing potential of the ¹⁴N(n,p)¹⁴C reaction was considered with regard to neutron absorption in human nuclear DNA and respiratory phosphates for: (A) 10¹² thermal neutrons in 1 kg of soft tissue, (B) a mono-energetic beam of 2 MeV neutrons incident in 1 kg of soft tissue such that the total collision kerma was 10 J/kg, and (C) an evenly distributed 0–66 MeV neutron beam, also incident in 1 kg such that the total collision kerma was 20 J/kg. For case (A) 0.0017 ¹⁴N(n,p)¹⁴C reactions could be expected per DNA double strand, case (B) 0.053, and case (C) 0.0039. The probabilities that a proton emitted outside the nucleus would cross nuclear DNA were estimated from ¹⁴N tissue content for adult skeletal muscle, liver, and kidney tissues, for (1) nuclear DNA being concentrated in a sphere of 1.8 μm diameter, and (2) nuclear DNA being evenly distributed in a spherical nucleus 5 μm in diameter. It was concluded that even in a nitrogen-rich tissue exposed to a collision kerma of 20 J/kg by a 0–66 MeV fast neutron beam, the ¹⁴N(n,p)¹⁴C reaction directly kills at most 10 cells in every 1000, 4 of these by DNA nitrogen absorption and 6 by the ¹⁴N(n,p)¹⁴C protons emitted elsewhere in the cell. However, the dose due to the ¹⁴N(n,p)¹⁴C reaction should be measured where exposure to thermal neutron fluxes is significant. For therapeutic neutron doses the number of respiratory phosphate molecules in which the ¹⁴N(n,p)¹⁴C reaction occurs is insignificant, and doses from ¹⁴C-decay after neutron therapy are also negligible.     

241Am-Be中子参考辐射的蒙特卡罗模拟研究

研究用于校准场所中子剂量监测仪表的241Am-Be中子参考辐射场计量特性。采用蒙特卡罗方法模拟了空气自由中子参考辐射(FRNR),GB/T 14055规定的最小尺寸中子参考辐射(SRNR)和实际中子参考辐射(ARNR)中不同检验点处中子周围剂量当量率、散射中子占比和能谱分布特征。研究结果表明,空气对FRNR中的剂量率和能谱分布影响小,近似为理想中子参考辐射;采用5%含硼聚乙烯作屏蔽的最小尺寸SRNR可减少热中子,降低散射中子占比,影锥法不适用于小尺寸中子参考辐射中对散射中子的修正;ARNR中的散射中子更少、占比更低,影锥法所得散射中子占比与理论值基本一致。     

2.5-MeV neutron source controlled by high-intensity pulsed laser generating plasma

A laptop neutron source suited for the most demanding field or laboratory applications is presented. It is based on laser ablation of CD₂ primary targets, plasma acceleration of the D⁺ ions, and their irradiation of secondary CD₂ targets. The deuterium–deuterium (D-D) fusion reaction is induced in the secondary target, according to the values of fusion cross-section versus deuteron energy, which show a significant probability also at relatively low ion energies. The experiments were completed in the PALS laboratory, Prague, detecting monoenergetic neutrons at 2.45 MeV with an emission flux of about 10⁹ neutrons per laser shot. Other experiments demonstrating the possibility to induce D-D events were performed at IPPLM, Warsaw, and at INFN-LNS, Catania, where the deuterons were accelerated at about 4 MeV and 50 keV, respectively. In the last case, a low laser intensity and a post-ion acceleration system were employed. A special interaction chamber, under vacuum, is proposed to develop a new source of monochromatic neutrons or thermalized distribution of neutrons     

GEANT4 Monte Carlo simulation response of parallel plate avalanche counter for fast neutrons detection

In the present study, GEANT4 based Monte Carlo codes have been employed to evaluate parallel plate avalanche counter (PPAC) response for fast neutron detection. In order to detect fast neutrons, a thin polyethylene layer is coated on the surface of the electrode of the PPAC. Neutrons entering the converter produce protons which enter the counter and are detected. Fast neutrons in the energy range of 4.0 MeV–20.0 MeV have been transported onto the PPAC surface using GEANT4 MC code. The performance of the PPAC counter has been evaluated by means of simulation by employing QGSP_BERT_HP and QGSP_BIC_HP physics lists. The detection efficiencies of polyethylene-coated PPAC are 1.69 × 10^?2 and 1.86 × 10^?2 using converter thickness of 1 mm and 2 mm, respectively. The obtained results are discussed in detail.