Determination of spatial and energy distributions of neutrons in experiments on transmutation of radioactive waste using relativistic protons

The experiments on transmutation of ¹²⁹I and ²³⁷Np using uranium-lead targets surrounded by a paraffin moderator were performed at the Joint Institute for Nuclear Research (JINR, Russia). The targets were irradiated by 1.5 GeV and 7.4 GeV protons at the Synchrophasotron of JINR. In the frame of present work spatial and energy distributions of neutrons on the surface of the paraffin moderator were measured using SSNTD technique. It is shown that measured values of spetral indices do not pretend on the energy of incident protons but depend on the target composition. The presence of the uranium insertion softens neutron spectra.     

Neutron yields from massive lead and uranium targets irradiated with relativistic protons

Long-lived isotopes can be transmuted into stable or short-lived elements either by neutron captures or neutron induced fission. The need of a large excess of neutrons has led to the use of accelerator driven sources (ADS). A series of experiments were carried out at the Synchrophasotron/Nuclotron of the Joint Institute for Nuclear Research (JINR) Dubna, using protons of 1.0 GeV. Solid Lead and Uranium targets surrounded by paraffin moderator were irradiated. On the outer surface of the moderator a number of Solid State Track Detectors were placed to monitor neutron spatial distribution. The results showed that the maximum neutron production was reached within the range of one to two proton mean free paths in the target. Then decreasing neutron production follows the proton beam attenuation along the target. Moreover, the results showed both targets neutron production evolution along the target, to be the same. However, neutron flux per incident proton is depended on the target mass, which was found to be higher for the heavier target.     

加速器驱动次临界系统散裂靶参数优化计算

采用国际开源程序包Geant4,构建高能质子束轰击加速器驱动次临界系统(ADS)散裂靶的物理模型,模拟计算质子轰击液态金属铅、铅-铋合金和汞靶的泄漏中子谱分布,以及计算不同能量质子对应的铅靶泄漏中子产额和轴向积分分布,获得1 Ge V质子对应的铅圆柱靶优化参数,考虑入射质子的利用率和整个堆芯的体积质量,优化靶半径范围为16～24 cm,靶高为100 cm,相关研究结果可为(ADS)散裂靶的物理和工程设计提供理论依据。     

Determination of slow neutron flux and spatial distribution using LR-115 2B detectors and radiochemical sensors and comparison of the results with Monte Carlo simulations

The fluence and spatial distribution of slow neutrons on the surface of a cylindrical paraffin moderator, surrounding a Pu/Be neutron source was measured using LR-115 2B detectors via ¹⁰B(n,)⁷Li reaction and ¹³⁹La radiochemical sensors. The obtained results were compared with Monte Carlo simulation using MCNP-4B2 code.     

Designing of the 14 MeV neutron moderator for BNCT

In boron neutron capture therapy (BNCT), the ratio of the fast neutron flux to the neutron flux in the tumor (RFNT) must be less than 3%. If a D-T neutron generator is used in BNCT, the 14 MeV neutron moderator must be optimized to reduce the RFNT. Based on the neutron moderation theory and the simulation results, tungsten, lead and diamond were used to moderate the 14 MeV neutrons. Satisfying RFNT of less than 3%, the maximum neutron flux in the tumor was achieved with a three-layer moderator comprised of a 3 cm thick tungsten layer, a 14 cm thick lead layer and a 21 cm thick diamond layer.     

HL-2A中子相机屏蔽系统模拟

利用蒙特卡罗模拟程序,建立了HL-2A中子相机蒙特卡罗粒子输运(MCNP)物理模型,对D-D聚变中子和γ射线的屏蔽进行了模拟计算。对石蜡碳酸锂混合物、聚乙烯、铅和316L不锈钢4种常用中子慢化吸,收剂组成的屏蔽层材料的屏蔽效果进行了对比。计算结果表明,石蜡碳酸锂混合物和铅组合是中子相机的最佳屏蔽层材料,其中石蜡碳酸锂混合物用于慢化吸收中子,铅用于屏蔽中子和γ射线。此外,利用MCNP模拟计算得到了屏蔽中子和γ射线所需的屏蔽厚度,以及准直管的中子散射率。     

HL-2A�����������ϵͳģ��

The physical model of the neutron camera Monte-Carlo partical transport (MCNP) for HL-2A was established by using Monte-Carlo simulation code. The shielding of D-D fusion neutrons and gamma rays was simulated. The shielding effects were compared for four common shielding materials, including mixture of paraffin lithium carbonate, polyethylene, lead, 316L stainless steel. Calculation results show that mixture of paraffin lithium carbonate and lead are the best shielding materials for neutron camera, among them the mixture of paraffin lithium carbonate is used for slowing-down and absorpting neutrons, while lead is used for blocking neutrons and gamma rays. In addition, both the required thickness of shielding material for neutron and gamma ray and the neutron scattering rate of collimator tube have been obtained by using MCNP simulation.     

The spatial distribution of thermal and epithermal neutrons in a graphite moderated spallation neutron field

The Gamma-3 assembly is located at the Joint Institute for Nuclear Research, Dubna, Russia. It consists of a cylindrical lead target (ø = 8 cm, L = 58.8 cm) surrounded by reactor grade graphite (110 × 110 × 60 cm). The target was irradiated with a beam of 1.6 GeV deuterons from the Nuclotron accelerator and CR-39 track detectors coupled to LR-115 2B film were used to measure the slow neutron distribution on the surface of the graphite. The detection efficiency of the CR-39 in the CR-39/LR-115 2B system was measured using a custom made calibration setup and found to be (1.12 ± 0.05) × 10⁻³ and (6.1 ± 1.2) × 10⁻⁴ tracks per neutron, for thermal and epithermal neutrons respectively, under the etching and counting procedures described in this work. The irradiation of the Gamma-3 was also simulated using MCNPX 2.7 Monte Carlo code and good agreement between the experimental and calculated track densities was found. This serves as a good validation for the computational models used to simulate spallation neutron production, transport and moderation.     

Neutron light output function and resolution investigation of the deuterated organic liquid scintillator EJ-315

In this paper we investigate the light response to fast neutrons and estimate the pulse height resolution of a deuterated liquid organic scintillator, EJ-315, considering the detector's non-linear light response to gamma-rays. Initially, collision data and a neutron beam trigger are recorded in coincidence mode, and incident neutron energy is calculated with a time-of-flight technique. Fast neutrons are further discriminated from gamma-ray background based on the scintillation material decay patterns using a pulse shape discrimination algorithm. A light response matrix composed of multiple neutron energy and their corresponding light outputs is derived. The pulse height resolution property of the EJ-315 is characterized utilizing the derivatives of the pulse height distributions with corrections of the measurements setup uncertainties. Additionally, the EJ-315's pulse height resolution is also characterized by comparing the smoothed derivatives of quasi-monoenergetic neutron pulse height distributions, given by the Peierls-formula-based analytic model, to match the measurement data. Results show rather consistent 10–13% pulse resolution for mono-energetic neutrons with kinetic energy above 2 MeV. The resolution decreases slightly with an increase in neutron energy indicating the improved resolution performance of EJ-315 in the higher energy events.     

Comparative Study of Depth Dose Distribution of Fast Neutrons Using Solid State Nuclear Track Detectors

A comparative study was made for measuring neutron depth dose distribution in a water phantom by means of different solid state nuclear track detectors. The used detectors were Makrofol, CR-39 and LR-115 with threshold energies of 1 and 0.1 MeV, respectively. Two exposure positions were used, the source in contact with the surface and at 30 cm from the phantom surface. The fluence distribution of ²⁵²Cf as a function of distance in air and in the water phantom were investigated. It was found that the neutron build up factor depends on the irradiation geometry, its values were 1.7 and 1.4 for both exposure positions. These results are in good agreement with the calculated results given in literature. The neutron attenuation and relaxation length had been also determined.     

Intercomparative studies of brachytherapy techniques combined with CF Boron neutron capture therapy

In this work the depth dose distribution for γ-rays, slow and fast neutrons, nuclear recoils and -particles inside a tumor phantom filled with 100 ppm ¹⁰B in distilled water was measured. The depth dose distributions were measured and represented on three-dimensional plots when a ²⁵²Cf neutron source was situated in interstitial, surface and intracavitary brachytherapy techniques. Comparison of the three brachytherapy techniques when combined with BNCT revealed that the damage induced in the tumor by the -particles as nuclear recoils in the interstitial geometry is much more than the other two techniques.     

Measurement of neutron-induced activation cross-sections using spallation source at JINR and neutronic validation of the Dubna code

A beam of 1 GeV proton coming from Dubna Nuclotron colliding with a lead target surrounded by 6 cm paraffin produces spallation neutrons. A Th-foil was kept on lead target (neutron spallation source) in a direct stream of neutrons for activation and other samples of ¹⁹⁷Au, ²⁰⁹Bi, ⁵⁹Co, ¹¹⁵In and ¹⁸¹Ta were irradiated by moderated beam of neutrons passing through 6 cm paraffin moderator. The gamma spectra of irradiated samples were analyzed using gamma spectrometry and DEIMOS software to measure the neutron cross-section. For this purpose neutron fluence at the positions of samples is also estimated using PREPRO software. The results of cross-sections for reactions ²³²Th(n, γ), ²³²Th(n, 2n), ¹⁹⁷Au(n, γ), ¹⁹⁷Au(n, α), ¹⁹⁷Au(n, xn), ⁵⁹Co(n, α), ⁵⁹Co(n, xn), ¹⁸¹Ta(n, γ) and ¹⁸¹Ta(n, xn) are given in this paper. Neutronics validation of the Dubna Cascade Code is also done using cross-section data by other experiments.      

Velocity distribution of desorbing CO2 in CO oxidation on Pd(110) under steady-state conditions

The velocity and angular distributions of desorbing CO₂ in CO oxidation on Pd(110) were studied under the steady-state conditions with cross-correlation time-of-flight techniques. Sharp changes were found in both distributions at the switching point of the rate-determining step. The velocity curves involved fast and slow components. The fast one, which was suddenly suppressed at the switching point, showed a translational temperature from 2300 to 4000 K. The slow component was thermalized to the surface temperature. A rapid site switching for CO₂ formation is proposed, in which the fast component is formed on flat parts of the surface and the slow one is formed on structural defects.     

Distribution of fast and slow states by photoelectric measurements on real InSb and Si surfaces

At T = 79 K, X rays change the surface properties of n and p type InSb single cryslals as well as the properties of Si-MOS structures. In order to study the influence of surface parameters the distributions of fast and slow surface states were measured at various stages of irradiation. The distribution of fast states on real InSb surfaces has been investigated by techniques similiar to those employed in measurements of current voltage characteristics. By using a combination of integral ac field effect with differential ac field effect the continuous spectrum of fast states was studied not only within the depletion region but also under strong accumulation and inversion. Analysis of photodischarging spectroscopy gives information on the energetic location of slow surface states on real Si surfaces.     

Neutron response study of two CR-39 personal dosemeters with air and Nylon converters

A neutron personal dosemeter using CR-39 as a detector and hydrogenated materials as proton converters for fast neutron detection plus an air layer for thermal neutron detection is being developed in our laboratory. To increase the CR-39 response to thermal neutrons, the air converter was substituted with Nylon in some dosemeters. Several dosemeters with these two configurations were mounted on a water-filled phantom and exposed under different incidence angles (0, 30 and 60) to: (i) Three ISO neutron sources (²⁴¹Am–Be, bare ²⁵²Cf and moderated ²⁵²Cf with Cd shielding), and (ii) two realistic neutron sources (SIGMA and moderated ²⁵²Cf) at the IRSN (Cadarache) facilities. The irradiated detectors were electrochemically etched and evaluated in order to determine their dose equivalent response in terms of H_p(10,α). The results obtained are compared to those obtained from Monte Carlo simulations using the MCNPX code.     

Measurements of neutron yields and spatial distributions in U/Pb, Pb and Hg thick targets bombarded by 0.5 and 1.0 GeV protons

Measurements have been carried out on neutron yields and spatial distributions in U/Pb, Pb and Hg thick targets and the surrounding paraffin moderators bombarded by 0.533 and 1.0 GeV protons. CR-39 detectors were deployed on the surfaces of targets and moderators to record the neutrons produced in the targets. The measurements show that:

Full-size image

Precision measurement of the quasifree pn-->dvarphi reaction close to threshold

The quasifree pn-->dvarphi reaction has been studied at the Cooler Synchrotron COSY-Jülich, using the internal proton beam incident on a deuterium cluster-jet target and detecting a fast deuteron in coincidence with the K+K- decay of the varphi meson. The energy dependence of the total and differential cross sections are extracted for excess energies up to 80 MeV by determining the Fermi momentum of the target neutron on an event-by-event basis. Though these cross sections are consistent with s-wave production, the kaon angular distributions show the presence of p waves at quite a low energy. Production on the neutron is found to be stronger than on the proton but not by as much as for the eta meson.     

The effects of shielding around the source on neutron energy spectra

The knowledge of neutron energy spectra contributes to unambiguous identification of neutron sources in the fields of nuclear safeguards and nuclear non-proliferation. Since a real scenario situation includes the presence of shielding around the source, we have investigated the influence of the potential shielding surrounding the source on the shape of energy spectra for a few neutron sources. We have applied the maximum-likelihood, expectation–maximisation (MLEM) method with one-step-late (OSL) algorithm for neutron spectra unfolding. The pulse height distributions used in the unfolding procedures were simulated with the high accuracy by using the MCNP-PoliMi code based on the Monte Carlo method. A possibility to identify the shielded neutron sources by using the unfolding method was examined with two continuous-in-energy sources, such as ²⁵²Cf and ²⁴¹Am–Be in source-shielding configurations with lead (Pb) and polyethylene (PE) blocks. The results of calculations have shown that the identification of ²⁵²Cf and ²⁴¹Am–Be sources with 2.5 cm of Pb and PE shield can be achieved successfully by using the MLEM method with the OSL algorithm. However, the unfolded results for ²⁵²Cf and ²⁴¹Am–Be sources with 10 cm of PE shield significantly deviate from the reference spectra and the sources cannot be correctly identified on the basis of their unfolded energy spectra.     

The measurements of thermal neutron flux distribution in a paraffin phantom

The term ‘thermal flux’ implies a Maxwellian distribution of velocity and energy corresponding to the most probable velocity of 2200 m s^???1 at 293.4 K. In order to measure the thermal neutron flux density, the foil activation method was used. Thermal neutron flux determination in paraffin phantom by counting the emitted rays of indium foils with two different detectors (Geiger–Muller counter and NaI(Tl)) was the aim of this project. The relative differences of the outcome of the experiments were between 2.5% and 5%. The final results were compared with MCNP4C outputs and the best agreement was generated using NaI(Tl) by a minimum discrepancy of about 0.6% for the foil placed 8.5 cm from the neutron source.     

Measurement of dose enhancement by neutron capture of boron in Cf interstitial brachytherapy

The effect of boron concentration in water on the gamma, fast and slow neutrons and alpha particles components at the central, forward and backward surfaces inside tumor phantom of 4.2 cm diameter and 4.4cm height, during brachytherapy by neutrons from ²⁵²Cf were investigated. The source was at the centre of a cubic shaped water phantom of 30 cm side. The study was carried for different concentrations of boron from H₃BO₃, Li₂B₄O₇ and H₃¹⁰BO₃. The effect of source to tumor distance on the different components of radiation was also measured. The results indicated that the use of ¹⁰B compounds enhances the damage and is recommended for successful boron neutron capture therapy (BNCT).